WorldWideScience

Sample records for design basis accidents

  1. Assessment of WWER fuel condition in design basis accident

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.; Sokolov, N.; Andreeva-Andrievskaya, L.; Vlasov, Yu.; Nechaeva, O.; Salatov, A.

    1994-01-01

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR's. 12 figs., 11 refs

  2. Assessment of WWER fuel condition in design basis accident

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu; Sokolov, N; Andreeva-Andrievskaya, L; Vlasov, Yu; Nechaeva, O; Salatov, A [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR`s. 12 figs., 11 refs.

  3. Canister storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

    KOPELIC, S.D.

    1999-01-01

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  4. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    PIEPHO, M.G.

    1999-10-20

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  5. Canister storage building design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  6. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    CROWE, R.D.

    1999-01-01

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  7. Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    PIEPHO, M.G.

    1999-01-01

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR

  8. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    International Nuclear Information System (INIS)

    CROWE, R.D.; PIEPHO, M.G.

    2000-01-01

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  9. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    1999-09-09

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  10. Study on risk factors of PWR accidents beyond design basis

    International Nuclear Information System (INIS)

    Ahn, Seung Hoon; Nah, W. J.; Bang, Y. S.; Oh, D. Y.; Oh, S. H.

    2005-01-01

    Development of the regulatory guidelines for Beyond Design Basis Accidents (BDBA) with high risk requires a detailed investigation of major factors contributing to the event risk. In this study, each event was classified by the level of risk, based on the probabilistic safety assessment results, so that BDBA with high risk could be selected, with consideration of foreign and domestic regulations, and operating experiences. The regulatory requirements and technical backgrounds for the selected accidents were investigated, and effective regulatory approaches for risk reduction of the accidents. The following conclusions were drawn from this study: - Selected high risk BDBA is station blackout, anticipated without scram, total loss of feedwater. - Major contributors to the risk of selected events were investigated, and appropriate assessment of them was recommended for development of the regulatory guidelines

  11. Cold Vacuum Drying facility design basis accident analysis documentation

    International Nuclear Information System (INIS)

    CROWE, R.D.

    2000-01-01

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls

  12. Cold Vacuum Drying facility design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    CROWE, R.D.

    2000-08-08

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

  13. Transient and accident analyses topical design basis documents

    International Nuclear Information System (INIS)

    Chi, Larry; Eckert, Eugene; Grim, Brit

    2004-01-01

    The designers and operators of nuclear power plants have extensively documented system functions, licensing performance, and operating procedures for all conditions. This paper presents a complementary, systematic approach for the documentation of all requirements that are based on the analysis of operational transients, abnormal transients, accidents, and other events which are included in the design and licensing basis for the plant. Up to now, application of the approach has focused on required mitigation actions (automatic or manual). All mitigation actions are directly identified with all applicable reactor events, as well as the plant-unique systems that work together to perform each function. The approach is also applicable to all operational functions. The approach makes extensive use of data base methods, thereby providing effective ways to interrogate the information for the varied users of this information. Examples of use include: evaluations of system design changes and equipment modifications, safety evaluations of any plant change (e.g., USNRC 10CFR50.59 review), plant operations (e.g., manual actions during unplanned events), system interactions, classification of safety-related equipment, environmental qualification of equipment, and mitigation requirements for different reactor operating states. This approach has been applied in customized ways to several boiling water reactor (BWR) units, based on the desires and needs of the specific utility. (author)

  14. Accident beyond the design basis management with the coolant loss at the NPP with WWER

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Klyuchnikov, A.A.; Kolykhanov, V.N.

    2010-01-01

    The analysis of status and experience of development on modelling and accident beyond the design basis management, including the severe accidents, at the nuclear power plants is carried out. The methodical providing of manuals on the accident beyond the design basis management with the coolant loss on the basis of simulated critical system configurations providing the necessary safety function performance on reactor unit is proposed. The project of symptom-oriented manuals on accident beyond the design basis management with the coolant loss on the serial power unit with WWER-1000 on the basis of developed methodical providing and well known results of deepened safety analysis is presented.

  15. Some conditions affecting the definition of design basis accidents relating to sodium/water reactions

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1984-01-01

    The possible damaging effects of large sodium/water reactions on the steam generator, IHX and secondary circuit are considered. The conditions to be considered in defining the design basis accidents for these components are discussed, together with some of the assumptions that may be associated with design assessments of the scale of the accidents. (author)

  16. A comparison of the consequences of the design basis accident of the Greek Research Reactor with those of a serious realistic accident

    International Nuclear Information System (INIS)

    Kollas, J.G.; Anoussis, J.N.

    1985-12-01

    An analysis of the radiological consequences of the design basis and the coolant flow blockage accidents of the Greek Research Reactor is presented. The results indicate that the consequences of the coolant flow blockage accident are practically trivial being 1-2 orders of magnitude lower than the corresponding consequences of the design basis accident. (author)

  17. Assessment Of Source Term And Radiological Consequences For Design Basis Accident And Beyond Design Basis Accident Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong; Tran Tri Vien

    2011-01-01

    The paper presents results of the assessment of source terms and radiological consequences for the Design Basis Accident (DBA) and Beyond Design Basis Accident (BDBA) of the Dalat Nuclear Research Reactor. The dropping of one fuel assembly during fuel handling operation leading to the failure of fuel cladding and the release of fission products into the environment was selected as a DBA for the analysis. For the BDBA, the introduction of a step positive reactivity due to the falling of a heavy block from the rotating bridge crane in the reactor hall onto a part of the platform where are disposed the control rod drives is postulated. The result of the radiological consequence analyses shows that doses to members of the public are below annual dose limit for both DBA and BDBA events. However, doses from exposure to operating staff and experimenters working inside the reactor hall are predicted to be very high in case of BDBA and therefore the protective actions should be taken when the accident occurs. (author)

  18. Beyond-design-basis accident management in the RF regulation documents

    International Nuclear Information System (INIS)

    Bukrinskij, A.M.

    2010-01-01

    The article observes the issues of the management of beyond-design-basis accidents (BDBA) in the existing regulations in Russia. The ideology of the approach to the definition of the BDBA list to formulate the management guidelines has been proposed [ru

  19. A cliff edge evaluation for CANDU-6 beyond design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.M.; Kho, D.W., E-mail: wolsong@khnp.co.kr [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Yi, S.D.; Kang, S.H.; Kim, S.R. [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

    2015-07-01

    The condition of nuclear power plant in the event of station black out (SBO) accompanying large-scale natural disaster exceeding design basis accident (DBA) was evaluated. Additional scenarios were added to the evaluation to review capability of the plant to endure different conditions with different actions. The analysis resulted that the key action required from the operator was to ensure the opening of main steam safety valves (MSSVs) in the secondary side and of motor-operated valves for high pressure injection of Emergency Core Cooling System (HPECCS) to mitigate accidents or extend the cliff edge. (author)

  20. Tests of qualification of national components of nuclear power plants under design basis accident

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1990-01-01

    With the purpose of qualifying national components of nuclear power plants, whose working must be maintained during and after an accident, the Thermohydraulic Division of CDTN have done tests to check the equipment stability, under Design Basis Accident conditions. Until this moment, the following components were tested: electrical junction boxes (connectors); coating systems for wall, inside cover and steel containment; hydraulics components of personnel and equipment airlock. This work describes the test instalation, the tests performed and its results. The components tested, in a general way, fulfil the specified requirements. (author) [pt

  1. Screening and analysis of beyond design basis accident of 49-2 SPR

    International Nuclear Information System (INIS)

    Zhang Yadong; Guo Yue; Wu Yuanyuan; Zou Yao

    2015-01-01

    The beyond design basis accident was analyzed to ensure safe operation of 49-2 Swimming Pool Reactor (SPR) after design life. Because it's difficult to use PSA method, the unconditional assumed severe accidents were adopted to obtain a conservative result. The main conclusions were obtained by analyzing anticipated transients without scram in station blackout (SBO ATWS), horizontal channel rupture, core uncovering after shutdown and emergency response capacity. The results show that the core is safe in SBO ATWS, and the fuel elements will not melt as long as the core are not exposed in 2.5 h in loss of coolant accident caused by horizontal channel rupture and other factors. The passive siphon breaker function and various ways of emergency core makeup can ensure that the core is not exposed. (authors)

  2. Beyond Design Basis Severe Accident Management as an Element of DiD Concept Strengthening

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsov, M., E-mail: kuznetsov_mv@vosafety.ru [FSUE VO “Safety”, Moscow (Russian Federation)

    2014-10-15

    The 4{sup th} Level of DiD is ensured by management of beyond design basis accidents which is achieved by implementation of the Beyond Design Basis Accidents Management Guidance (BDBAMG) and, if necessary, by additional technical devices and organizational measures at NPP Unit. BDBAMG is located between Levels 3 and 5 in DiD and is related to them. It is connected with Level 3 by means of conditions generated at this Level and according to which BDBAM should be initiated (Level 4). It is associated with Level 5 by conditions which necessitate implementation of Emergency planning. Both types of conditions should be identified in BDBAMG. BDBAs including the phase of severe damage of fuel and protective barriers (severe accidents) in accordance with Russian regulatory framework are a subset of all BDBAs set. In this connection, such accident scenarios meet the representativeness criterion for further analysis and development of Guidance for their management. BDBAMG availability, as it provides robustness of DiD as a whole, is an obligatory condition for obtaining a NPP operational license. In the process of BDBAMG development and implementation a feedback with technical and organizational measures, comprising Level 1 and, to a less extent, Level 2, comes up. BDBAMG verification is an important final stage of its development. Addressing severe accidents, it is a challenging issue for a full scope simulator and may require its software modernization to make it responsive to severe accident phenomena. The existing BDBAMGs should be updated due to NPP Unit modernizations and in conjunction with the latest knowledge on severe accident phenomenology and lessons learnt from known events (e.g. NPP Fukushima). Thus, improvements incorporated in BDBAMG, enhance the strength of DiD. (author)

  3. Guidance on the implementation of modifications to mitigate beyond design basis accidents

    International Nuclear Information System (INIS)

    Harris, S.; Marczak, J.; O'Neill, M.

    2014-01-01

    Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. Operating experience with use of the guidance is also discussed. (author)

  4. Guidance on the implementation of modifications to mitigate beyond design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    Harris, S.; Marczak, J.; O' Neill, M. [Ontario Power Generation, Pickering, ON (Canada)

    2014-07-01

    Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. Operating experience with use of the guidance is also discussed. (author)

  5. Guidance on the Implementation of Modifications to Mitigate Beyond Design Basis Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Dermarkar, F.; Marczak, J.; O’Neill, M., E-mail: fred.dermarkar@opg.com [Ontario Power Generation, Pickering, Ontario (Canada)

    2014-10-15

    Following the events at Fukushima, Canadian Nuclear Power Plants (NPP) procured equipment and initiated modifications to improve response capability for Beyond Design Basis Accidents (BDBA). These changes were not typical of other design modifications to the nuclear power plants and reinforced the need for additional guidance for modifications to address BDBA. This paper describes the guidance that was developed to guide the design, procurement, installation, operation, and maintenance of equipment and modifications to mitigate BDBAs. The guidance developed prescribes a graded approach based on a categorization of the nature of the modification. Four categories of modifications are introduced, with the distinction being the degree of interface with existing design basis systems, structures and components (SSCs). This has resulted in a cost-effective means of implementing additional capability to mitigate BDBA conditions, and yet ensure the design basis capability of SSCs is maintained. (author)

  6. Considerations on Fail Safe Design for Design Basis Accident (DBA) vs. Design Extension Condition (DEC): Lesson Learnt from the Fukushima Accident

    International Nuclear Information System (INIS)

    Ha, Jun Su; Kim, Sungyeop

    2014-01-01

    The fail safety design is referred to as an inherently safe design concept where the failure of an SSC (System, Structure or Component) leads directly to a safe condition. Usually the fail safe design has been devised based on the design basis accident (DBAs), because the nuclear safety has been assured by securing the capability to safely cope with DBAs. Currently regards have been paid to the DEC (Design Extension Condition) as an extended design consideration. Hence additional attention should be paid to the concept of the fail safe design in order to consider the DEC, accordingly. In this study, a case chosen from the Fukushima accident is studied to discuss the issue associated with the fail safe design in terms of DBA and DEC standpoints. For the fail safe design to be based both on the DBA and the DEC, a Mode Changeable Fail Safe Design (MCFSD) is proposed in this study. Additional discussions on what is needed for the MCFSD to be applied in the nuclear safety are addressed as well. One of the lessons learnt from the Fukushima accident should include considerations on the fail-safe design in a changing regulatory framework. Currently the design extension condition (DEC) including severe accidents should be considered during designing and licensing NPPs. Hence concepts on the fail safe design need to be changed to be based on not only the DBA but also the DEC. In this study, a case on a fail-safe design chosen from the Fukushima accident is studied to discuss the issue associated with the fail safe design in terms of DBA and DEC conditions. For the fail safe design to be based both on the DBA and the DEC, a Mode Changeable Fail Safe Design (MCFSD) is proposed in this study. Additional discussions on what is needed for the MCFSD to be applied in the nuclear safety are addressed as well

  7. Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2015-11-01

    The demands on nuclear fuel have recently been increasing, and include transient regimes, higher discharge burnup and longer fuel cycles. This has resulted in an increase of loads on fuel and core internals. In order to satisfy these demands while ensuring compliance with safety criteria, new national and international programmes have been launched and advanced modelling codes are being developed. The Fukushima Daiichi accident has particularly demonstrated the need for adequate analysis of all aspects of fuel performance to prevent a failure and also to predict fuel behaviour were an accident to occur.This publication presents the Proceedings of the Technical Meeting on Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents, which was hosted by the Nuclear Power Institute of China (NPIC) in Chengdu, China, following the recommendation made in 2013 at the IAEA Technical Working Group on Fuel Performance and Technology. This recommendation was in agreement with IAEA mid-term initiatives, linked to the post-Fukushima IAEA Nuclear Safety Action Plan, as well as the forthcoming Coordinated Research Project (CRP) on Fuel Modelling in Accident Conditions. At the technical meeting in Chengdu, major areas and physical phenomena, as well as types of code and experiment to be studied and used in the CRP, were discussed. The technical meeting provided a forum for international experts to review the state of the art of code development for modelling fuel performance of nuclear fuel for water cooled reactors with regard to steady state and transient conditions, and for design basis and early phases of severe accidents, including experimental support for code validation. A round table discussion focused on the needs and perspectives on fuel modelling in accident conditions. This meeting was the ninth in a series of IAEA meetings, which reflects Member States’ continuing interest in nuclear fuel issues. The previous meetings were held in 1980 (jointly with

  8. Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents

    International Nuclear Information System (INIS)

    Houghton, W.J.

    1980-06-01

    A probabilistic risk assessment (PRA) approach is proposed to be used to scrutinize selection of accident sequences. A technique is described in this Licensing Topical Report to identify candidates for Design Basis Accidents (DBAs) utilizing the risk assessment results. As a part of this technique, it is proposed that events with frequencies below a specified limit would not be candidates. The use of the methodology described is supplementary to the traditional, deterministic approach and may result, in some cases, in the selection of multiple failure sequences as DBAs; it may also provide a basis for not considering some traditionally postulated events as being DBAs. A process is then described for selecting a list of DBAs based on the candidates from PRA as supplementary to knowledge and judgments from past licensing practice. These DBAs would be the events considered in Chapter 15 of Safety Analysis Reports of high-temperature gas-cooled reactors

  9. iROCS: Integrated accident management framework for coping with beyond-design-basis external events

    International Nuclear Information System (INIS)

    Kim, Jaewhan; Park, Soo-Yong; Ahn, Kwang-Il; Yang, Joon-Eon

    2016-01-01

    Highlights: • An integrated mitigating strategy to cope with extreme external events, iROCS, is proposed. • The strategy aims to preserve the integrity of the reactor vessel as well as core cooling. • A case study for an extreme damage state is performed to assess the effectiveness and feasibility of candidate mitigation strategies under an extreme event. - Abstract: The Fukushima Daiichi accident induced by the Great East Japan earthquake and tsunami on March 11, 2011, poses a new challenge to the nuclear society, especially from an accident management viewpoint. This paper presents a new accident management framework called an integrated, RObust Coping Strategy (iROCS) to cope with beyond-design-basis external events (BDBEEs). The iROCS approach is characterized by classification of various plant damage conditions (PDCs) that might be impacted by BDBEEs and corresponding integrated coping strategies for each of PDCs, aiming to maintain and restore core cooling (i.e., to prevent core damage) and to maintain the integrity of the reactor pressure vessel if it is judged that core damage may not be preventable in view of plant conditions. From a case study for an extreme damage condition, it showed that candidate accident management strategies should be evaluated from the viewpoint of effectiveness and feasibility against accident scenarios and extreme damage conditions of the site, especially when employing mobile or portable equipment under BDBEEs within the limited time available to achieve desired goals such as prevention of core damage as well as a reactor vessel failure.

  10. Maximal design basis accident of fusion neutron source DEMO-TIN

    Energy Technology Data Exchange (ETDEWEB)

    Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

  11. Overview of Mobile Equipment Used in Case of Beyond Design Basis Accident at NPP Krsko

    International Nuclear Information System (INIS)

    Lukacevic, H.; Kopinc, D.; Ivanjko, M.

    2016-01-01

    Terrorist attack in USA in the September 11, 2001 and accident at the Fukushima - Daiichi Nuclear Power Station in the March 11, 2011 highlight the importance of mitigating strategies in responding to Beyond Design Basis Accident (BDBA), while ensuring cooling of reactor core, containment and spent fuel pool. Nuclear Power Plant Krsko (NEK) has acquired additional mobile equipment and made necessary modifications on existing systems for the connection of this equipment (fast couplers). Usage of mobile equipment is not only limited to design basis accident (DBA), but, also to prevent and mitigate the consequences in case of BDBA, when other plant systems are not available. NEK also decided to take steps for upgrade of safety measures and prepared Safety Upgrade Program (SUP), which is consistent with the nuclear industry response to the Fukushima accident and is implementing main projects and modifications related to SUP. NEK mobile equipment is not required to operate under normal reactor plant operation except for periodic surveillance testing and is incorporated into the normal training process. Equipment is dislocated from the reactor building and most of the equipment is located in the new building, able to withstand extreme natural events, including earthquakes and tornadoes. The usage of all mobile equipment is prescribed as an additional option in NEK operating procedures in following cases and enables following options: filling various tanks, filling the steam generators, filling the containment, additional compressed air source, spent fuel pool refilling and spraying, alternative power supply. This document provides an overview of NEK mobile equipment, which consists of various mobile fire protection pumps, air compressors, protective equipment, fire trucks, diesel generators. Sufficient fuel supply for the equipment is provided on site for a minimum three days of operation. (author).

  12. Critical analysis of accident scenario and consequences modelling applied to light-water reactor power plants for accident categories beyond the design basis accident (DBA)

    International Nuclear Information System (INIS)

    Brofferio, C.; Cagnetti, P.; Ferrara, V.; Manilia, E.; Pietrangeli, G.; Sennis, C.

    1985-01-01

    A critical analysis and sensitivity study of the modelling of accident scenarios and environmental consequences are presented, for light-water reactor accident categories beyond the standard design-basis-accident category. The first chapter, on ''source term'' deals with the release of fission products from a damaged core inventory and their migration within the primary circuit and the reactor containment. Particular attention is given to the influence of engineering safeguards intervention and of the chemical forms of the released fission products. The second chapter deals with their release to the atmosphere, transport and wet or dry deposition, outlining relevant partial effects and confronting short-duration or prolonged releases. The third chapter presents a variability analysis, for environmental contamination levels, for two extreme hypothetical scenarios, evidencing the importance of plume rise. A numerical plume rise model is outlined

  13. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  14. Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

    International Nuclear Information System (INIS)

    Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

    2013-01-01

    The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

  15. Criteria for calculating the efficiency of HEPA filters during and after design basis accidents

    International Nuclear Information System (INIS)

    Bergman, W.; First, M.W.; Anderson, W.L.; Gilbert, H.; Jacox, J.W.

    1994-12-01

    We have reviewed the literature on the performance of high efficiency particulate air (HEPA) filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be structurally damaged and have a residual efficiency of 0%. Despite the many studies on HEPA filter performance under adverse conditions, there are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen when there was insufficient data

  16. An overview of the UK regulatory expectation for design basis accident analysis

    International Nuclear Information System (INIS)

    Trimble, Andy

    2013-01-01

    The UK Health and Safety Executive published its most recent regulatory expectations in the 2006 version of it's safety assessment principles (SAPs). This built on experience regulating the full range of facilities for which it is responsible. Thus the principles underpinning all regulatory safety case assessment are the same but the implementation differs depending on the application. This paper will describe the published design basis accident analysis (DBAA) logic in context with other technical aspects of the regulatory expectation for safety cases. It will further illustrate the DBAA methodology with practical examples from actual experience on reprocessing plant gained over the last 15 years or so. Among the examples will be the relevance of conventional safety fault initiators to nuclear safety assessment. It will further demonstrate the derivation of facility limits and conditions necessary for nuclear safety. (authors)

  17. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    International Nuclear Information System (INIS)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W.

    2016-01-01

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  18. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-05-15

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  19. Regulatory requirements on accident management and emergency preparedness - concept of nuclear and radiation safety during beyond-design-basis accidents

    International Nuclear Information System (INIS)

    Yanke, R.

    2002-01-01

    Actual practice the and proposals for further activities in the field of Accident Management (AM) in the member countries of the Co-operation Forum of WWER regulators and in Western countries have been assessed. Further the results of the last working group on AM , the overview of interactions of severe accident research and the regulatory positions in various countries, IAEA reports, practice in Switzerland and Finland, were taken into consideration. From this information, the working group derived recommendations on Accident Management. The general proposals correspond to the present state of the art on AM. They do not describe the whole spectra of recommendations on AM for NPPs with WWER reactors. A basis for the implementation of an AM program is given, which could be extended in a follow-up working group. The developments and research concerning AM have to be continued. The positions of various countries with regard to the 'Interactions of severe accident research and the regulatory positions' are given. On the basis of the working group proposals, the WWER regulators could set regulatory requirements and support further developments of AM strategies, making use of the benefits of common features of NPPs with WWER reactors. Concerted actions in the field of AM between the WWER regulators would bundle the development of a unified concept of recommendations and speed up the implementation of AM measures in order to minimise the risks involved in nuclear power generation

  20. Probabilistic risk assessment for the Los Alamos Meson Physics Facility worst-case design-basis accident

    International Nuclear Information System (INIS)

    Sharirli, M.; Butner, J.M.; Rand, J.L.; Macek, R.J.; McKinney, S.J.; Roush, M.L.

    1992-01-01

    This paper presents results from a Los Alamos National Laboratory Engineering and Safety Analysis Group assessment of the worse-case design-basis accident associated with the Clinton P. Anderson Meson Physics Facility (LAMPF)/Weapons Neutron Research (WNR) Facility. The primary goal of the analysis was to quantify the accident sequences that result in personnel radiation exposure in the WNR Experimental Hall following the worst-case design-basis accident, a complete spill of the LAMPF accelerator 1L beam. This study also provides information regarding the roles of hardware systems and operators in these sequences, and insights regarding the areas where improvements can increase facility-operation safety. Results also include confidence ranges to incorporate combined effects of uncertainties in probability estimates and importance measures to determine how variations in individual events affect the frequencies in accident sequences

  1. Development of dose calculation program (DBADOSE) incorporating alternative source term due to design basis accident

    International Nuclear Information System (INIS)

    Bae, Young Jig; Nam, Ki Mun; Lee, Yu Jong; Chung, Chan Young

    2003-01-01

    Source terms presented in TID-14844 and Regulatory Guide 1.4 have been used for radiological analysis of design basis accidents for licensing existing pressurized water reactor (PWR). However, more realistic and physically-based source term based on results of study and experiments for about 30 years after the publication of TID-14844 was developed and presented in NUREG-1465 published by U.S NRC in 1995. In addition, ICRP has revised dose concepts and criteria through the publication of ICRP-9, 26, 60 and recommended effective dose concepts rather than critical organ concept since the publication of ICRP-26. Accordingly, multipurpose computer program called DBADOSE incorporating alternative source terms in NUREG-1465 and effective dose concepts in ICRP-60 was developed. Comparison of results of DBADOSE with those of POSTDBA and STARDOSE was performed and verified and no significant difference and inaccuracy were found. DBADOSE will be used to evaluate accidental doses for licensing application according to the domestic laws that are expected to be revised in the near future

  2. Including severe accidents in the design basis of nuclear power plants: An organizational factors perspective after the Fukushima accident

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Frutuoso e Melo, P.F.

    2015-01-01

    Highlights: • The Fukushima accident was man-made and not caused by natural phenomena. • Vulnerabilities were known by regulator and licensee but measures were not taken. • There was lack of independence and transparency of the regulatory body. • Laws and regulations have not been updated to international standards. • Organizational failures have played an important role in the Fukushima accident. - Abstract: The Fukushima accident was clearly an accident made by humans and not caused by natural phenomena as was initially thought. Vulnerabilities were known by both regulators and operator but they postponed measures. The emergency plan was not effective in protecting the public, because the involved parties were not sufficiently prepared to make the right decisions. The shortcomings and faults mentioned above resulted from the lack of independence and transparency of the regulatory body. Even laws and regulations, and technical standards, have not been upgraded to international standards. Regulators have not defined requirements and left for the operator to decide what would be more appropriate. In this aspect, there was clearly a lack of independence between these bodies and operator’s lobby power. The above situation raised the question of urgent updating of institutions, in particular those responsible for nuclear safety. The above evidences show that several nuclear safety principles were not followed. This paper intends to highlight some existing safety criteria that were developed from the operational experience of the severe accidents that occurred at TMI and Chernobyl that should be incorporated in the design of new nuclear power plants and to provide appropriate design changes (backfittings) for reactors that belong to the previous generation prior to the occurrence of these accidents, through the study of design vulnerabilities. Furthermore, the main criteria that define an effective regulatory agency are also discussed. Although these

  3. Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

    2015-08-15

    Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

  4. PCDP [Prototypical Spent Fuel Consolidation Equipment Demonstration Project] design basis accident report 9315-P-103, Rev. A

    International Nuclear Information System (INIS)

    1987-12-01

    The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) has identified a requirement to integrate the spent fuel rod consolidation design activities of each of several proposed geological repository facilities and the Monitored Retrievable Storage (MRS) facility, and to develop efficient and cost-effective equipment for the consolidation process. The equipment to be developed for the rod consolidation system will be required to operate in a dry environment at rates which can be appropriately scaled to approximate the waste management system acceptance rates, irrespective of repository geologic characteristics or the existence of an MRS facility in the waste management system. The purpose of this report is to identify and analyze the range of facility credible events and accident occurrences (from minor to the design basis accidents) and their causes and consequences. For each situation, the considerations to prevent or mitigate the event or accident is addressed

  5. Results of the reliability investigations for the design basis accident 'Rupture of a cold primary coolant system'

    International Nuclear Information System (INIS)

    Hoertner, H.; Nieckau, E.; Spindler, H.

    1976-12-01

    This report gives a comprehensive presentation of the detailed reliability investigation carried out for the engineered safety features installed to cope with the design basis accident 'Large LOCA' of a German nuclear power plant with pressurized water reactor. The investigation is based on the engineered safety features of the Biblis Nuclear Power Plant, Unit A. The reliability investigation is carried out by means of a fault tree analysis. The influence of common-mode failures is assessed. (orig.) [de

  6. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Gyori, C.; Schubert, A.; Laar, J. van de; Hozer, Z.; Spykman, G.

    2008-01-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events

  7. Guidelines for calculation of atmospheric dispersion and radiological consequences of design basis reactor accidents - Severe accident calculation guidelines, EPR

    International Nuclear Information System (INIS)

    Martens, R.; Schmitz, B.M.; Horn, M.

    1999-01-01

    The activities carried out within the (reduced) project period (1. Sept. until 31. Dec. 1998) for coordinated harmonization between France and Germany, of guidelines for calculation of the radiological consequences of a severe reactor accident, are summarized. (orig./CB) [de

  8. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  9. Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

    International Nuclear Information System (INIS)

    Chvetsov, I.; Kouznetsov, I.; Volkov, A.

    2000-01-01

    GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

  10. Innovative safety features of VVER for ensuring high degree of autonomy during beyond design basis accidents

    International Nuclear Information System (INIS)

    Kumar, Abhay; Mohan, Joe; Kumar, Devesh; Chaudhry, S.M.; Rao, Srinivasa; Gupta, S.K.

    2010-01-01

    The effectiveness of Passive Heat Removal System (PHRS) in during a station black-out (SBO) accident was assessed by using SCDAP/Relap5. The analysis gave evidence that (i) the Passive Heat Removal System (PHRS) is capable of limiting the consequences of station black out (SBO) and acts as an effective engineered safety system, and (ii) the PHRS intervention prevents core degradation and excessive core heat-up. (P.A.)

  11. Calculation of stricken to mortality and incidence cancers due to beyond design basis accidents of the Esfahan Fuel Production Factory

    International Nuclear Information System (INIS)

    Heydari Azar, A.; Shahshahani, M.; Roshanzamir, M.; Sabouhi, R.

    2008-01-01

    In this investigation the amount of absorbed doses by the different pathways of Cloud shine, Ground shine, deposition of radioactive materials on skin and cloths, ingestion, inhalation and the consequences of radioactive material releases due to Beyond Design Basis Accidents such as fire, sintering furnace explosion, criticality and earthquake in Esfahan Fuel Production factory by the residents are evaluated. The calculations related to atomic cloud distribution, estimation of delivered dose and decay chains are performed by PCCOSYMA dose. These computations are based on radioactive source terms, distribution height of radioactive materials. actions for reducing the absorbed dose, human body physiological characteristics, metrological condition and population distribution. Finally, the number of peoples who are stricken to mortality and morbidity cancers and risk values are calculated for 1 year and 50 years

  12. Structural analysis of the CAREM-25 nuclear power plant subjected to the design basis accident and seismic loads

    International Nuclear Information System (INIS)

    Ambrosini, Daniel; Codina, Ramón H.; Curadelli, Oscar; Martínez, Carlos A.

    2017-01-01

    Highlights: • Structural analysis of CAREM-25 NPP is presented. • Full 3D numerical model was developed. • Transient thermal and static structural analyses were performed. • Modeling guidelines for numerical structural analysis of NPP are recommended. • Envelope condition of DBA dominates the structural behavior. - Abstract: In this paper, a numerical study about the structural response of the Argentine nuclear power plant CAREM-25 subjected to the design basis accident (DBA) and seismic loads is presented. Taking into account the hardware capabilities available, a full 3D finite element model was adopted. A significant part of the building was modeled using more than 2 M solid elements. In order to take into account the foundation flexibility, linear springs were used. The springs and the model were calibrated against a greater model used to study the soil-structure interaction. The structure was subjected to the DBA and seismic loads as combinations defined by ASME international code. First, a transient thermal analysis was performed with the conditions defined by DBA and evaluating the time history of the temperature of the model, each 1 h until 36 h. The final results of this stage were considered as initial conditions of a static structural analysis including the pressure defined by DBA. Finally, an equivalent static analysis was performed to analyze the seismic response considering the design basis spectra for the site. The different loads were combined and the abnormal/extreme environmental combination was the most unfavorable for the structure, defining the design.

  13. Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: Conclusions from an experimental program in a 4-loop test facility (PKL)

    International Nuclear Information System (INIS)

    Umminger, K.J.; Kastner, W.; Mandl, R.M.; Weber, P.

    1993-01-01

    Within the scope of German reactor safety research, extensive experiments covering the behavior of nuclear power plants under accident conditions have been carried out in the PKL test facility which simulates a 4-loop, 1,300 MWe KWU-designed PWR. While the investigations dealing with design-basis accidents and with the efficiency of the emergency core cooling systems have been largely completed, the main interest nowadays concentrates on the investigation of beyond-design-basis accidents to demonstrate the safety margins of nuclear power plants and to investigate the contribution of the built-in safety features for a further reduction of the residual risk. The thermal hydraulic behavior of a PWR under these extreme accident conditions was experimentally investigated within the PKL III B test program. This paper presents the fundamental findings with some of the most important results being discussed in detail. Future plans are also outlined

  14. Japan Catastrophic Earthquake and Tsunami in Fukushima Daiichi NPP; Is it Beyond Design Basis Accident or a Design Deficiency and Operator Unawareness?

    International Nuclear Information System (INIS)

    Gaafar, M.A.; Refeat, R.M.; EL-Kady, A.A.

    2012-01-01

    On March 11, 2011 a catastrophic earthquake and tsunami struck the north east coast of Japan. This catastrophe damaged fully or partially the six units of the Fukushima Daiichi Nuclear Power Plant.Questions were raised following the aftermath, whether it is beyond design basis accident caused by severe natural event or a failure by the Japanese authorities to plan to deal with such accident. There are many indications that the Utility of Fukushima Daiichi NPP, Tokyo Electric Power Company (TEPCO), did not pay enough attention to numerous facts about the incompatibility of the site and several design defects in the plant units. In fact there are three other NPP sites nearby Fukushima Daiichi Plant (about 30 to 60 Km far from Fukushima Daiichi NPP), with different site characteristics, which survived the same catastrophic earthquake and tsunami, but they were automatically turned into a safe shutdown state. These plants sites are Fukushima Daini Plant (4 units), Onagawa Plant (3 units) and Tokai Daini (II) Plant (one unit). In this paper, the aftermath Fukushima Daiichi plant integrity is pointed out. Some facts about the site and design concerns which could have implications on the accident are discussed. The response of Japan Authority is outlined and some remarks about their actions are underlined. The impacts of this disaster on the Nuclear Power Program worldwide are also discussed.

  15. Criteria for calculating the efficiency of deep-pleated HEPA filters with aluminum separators during and after design basis accidents

    International Nuclear Information System (INIS)

    Bergman, W.; First, M.W.; Anderson, W.L.

    1995-01-01

    We have reviewed the literature on the performance of HEPA filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). This study is only applicable to the standard deep-pleated HEPA filter with aluminum separators as specified in ASME N509[1]. Other HEPA filter designs such as the mini-pleat and separatorless filters are not included in this study. The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be damaged and have a residual efficiency of 0%. There are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen. The estimation of the efficiency of the HEPA filters under DBA conditions involves three steps: (1) The filter pressure drop and environmental parameters are determined during and after the DBA, (2) Comparing the filter pressure drop to a set of threshold values above which the filter is damaged. There is a different threshold value for each combination of environmental parameters, and (3) Determining the filter efficiency. If the filter pressure drop is greater than the threshold value, the filter is damaged and is assigned 0% efficiency. If the pressure drop is less, then the filter is not damaged and the efficiency is determined from literature values of the efficiency at the environmental conditions

  16. Criteria for calculating the efficiency of deep-pleated HEPA filters with aluminum separators during and after design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    Bergman, W.; First, M.W.; Anderson, W.L. [Lawrence Livermore National Laboratory, CA (United States)] [and others

    1995-02-01

    We have reviewed the literature on the performance of HEPA filters under normal and abnormal conditions to establish criteria for calculating the efficiency of HEPA filters in a DOE nonreactor nuclear facility during and after a Design Basis Accident (DBA). This study is only applicable to the standard deep-pleated HEPA filter with aluminum separators as specified in ASME N509[1]. Other HEPA filter designs such as the mini-pleat and separatorless filters are not included in this study. The literature review included the performance of new filters and parameters that may cause deterioration in the filter performance such as filter age, radiation, corrosive chemicals, seismic and rough handling, high temperature, moisture, particle clogging, high air flow and pressure pulses. The deterioration of the filter efficiency depends on the exposure parameters; in severe exposure conditions the filter will be damaged and have a residual efficiency of 0%. There are large gaps and limitations in the data that introduce significant error in the estimates of HEPA filter efficiencies under DBA conditions. Because of this limitation, conservative values of filter efficiency were chosen. The estimation of the efficiency of the HEPA filters under DBA conditions involves three steps: (1) The filter pressure drop and environmental parameters are determined during and after the DBA, (2) Comparing the filter pressure drop to a set of threshold values above which the filter is damaged. There is a different threshold value for each combination of environmental parameters, and (3) Determining the filter efficiency. If the filter pressure drop is greater than the threshold value, the filter is damaged and is assigned 0% efficiency. If the pressure drop is less, then the filter is not damaged and the efficiency is determined from literature values of the efficiency at the environmental conditions.

  17. Analysis of the metallic containment integrity of Angra 2/3 reactor under the effects of the design basis accident

    International Nuclear Information System (INIS)

    Costa, J.R.

    1981-06-01

    The application of Condru 4 computer code, developed to determine the maximum values of pressure and temperature that occur inside the metallic containment building of PWR nuclear power plants, in case of a hypothetic accident - LOCA - considered as a Design Basic Accident - DBA. The hypothesis, input and results for the simulation of a loss of coolant in the hot leg of the Angra-2/3 reactors, considered as the most critical case for that Kind of project, are presented. The analysis was made with input provided by the manufacturer. (Author) [pt

  18. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  19. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    International Nuclear Information System (INIS)

    Lazaro, A.; Schikorr, M.; Mikityuk, K.; Ammirabile, L.; Bandini, G.; Darmet, G.; Schmitt, D.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D.; Stempniewicz, M.

    2014-01-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

  20. Beyond design basis external flooding. Generic design assessment and lessons learned from the Fukushima Dai-ichi accident

    International Nuclear Information System (INIS)

    MacLeod, Tanya; Smith, Leslie; Allmark, Tim; Ford, Peter

    2017-01-01

    New reactors intended for construction in GB undergo the Office for Nuclear Regulation's (ONR's) Generic Design Assessment (GDA). GDA is a pre-licensing process that provides requesting parties with the opportunity to demonstrate at an early stage that the design is capable of meeting the legal requirements of Great Britain. During GDA, the intended reactor site may not yet be known. Therefore, requesting parties usually define a 'Generic Site' with characteristics typical for Great Britain. These characteristics should, as far as possible, bound the characteristics of known potential sites so that reactors of the proposed type could potentially be built at various suitable locations. This paper critically reviews ONR's approach to ensuring that external flooding is appropriately addressed at the GDA stage and covers: An overview of ONR's approach to post-Fukushima assessment. Changes to ONR's SAPs (Safety Assessment Principles) related to External Flooding. Two examples of post-Fukushima GDA approaches to External Flooding. Uncertainty and the provision of adequate safety measures. The paper concludes that the identification of potential vulnerabilities in the design to external flooding combined with a consideration of post-Fukushima resilience enhancements has led to increased regulatory confidence in the robustness of new reactor designs in GB against external flooding. (author)

  1. An assessment of the radiological consequences of the Greek Research Reactor's design basis accident with the use of low enriched uranium

    International Nuclear Information System (INIS)

    Kollas, J.G.

    1985-09-01

    An analysis of the radiological consequences of the design basis accident in the low enriched uranium fueled 5 MW Greek Research Reactor is presented. For the source term thirty-five isotopes are taken into consideration and conservative figures of fission product release are adopted. To estimate the reactor's consequences for Athens population a CRAC2 consequence model version is used. The results indicate that limiting dose and effects are respectively the thyroid dose and the thyroid effects induced in the 3,081,000 inhabitants of Athens region. (author)

  2. Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

    International Nuclear Information System (INIS)

    Heard, F.J.

    1997-01-01

    The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

  3. Report of the working group 'Regulatory requirements on AM - Concept of nuclear and radiation safety during beyond-design-basis accidents'

    International Nuclear Information System (INIS)

    Bobaly, P.

    2001-01-01

    The developed working group report contains the following main paragraphs: legal basis and basis for regulatory requirements for on-site and off-site Accident Management (AM), regulatory requirements or recommendations for on-site AM and for emergency preparedness, background information concerning the implementation and review of an AM program as a basis for an AM guideline. Overview about AM/SAM implementation in member countries of the SAMINE project; measure and candidates for high level actions based upon US SAMG; interactions of severe accident research and the regulatory positions, relationship between different components of an accident management programme are also given

  4. Methods for formulation of design basis accidents within a risk-informed approach to safety regulation of new nuclear power plants

    International Nuclear Information System (INIS)

    Beer, B.C.; Apostolakis, G.E.; Golay, M.W.

    2000-01-01

    Within a project sponsored by the U.S. Department of Energy (DOE) an investigation is being conducted into creating a risk-informed safety regulatory framework and design process based upon the use of probabilistic risk assessment (PRA). In conjunction with efforts to formulate an overall regulatory framework (i.e., reported in PSAM 5 by F. Duran, A. Camp, G. Apostolakis and M. Golay, 'A Framework for Regulatory Requirements and Industry Standards for New Nuclear Power Plants'), this paper addresses the potential role(s) of Design Basis Accidents (DBAs) within this new framework. Currently that role, if any, is unclear. In previous nuclear safety regulatory treatments, DBAs have been of great practical value for both designers and regulators. However, they have suffered from being inconsistently formulated, and lacking fundamental justification. Any DBA set is likely to be formulated uniquely for a specific reactor concept. The staff of any nuclear power plant (NPP) in the U.S. routinely calculates the likelihood of core damage, the likelihood of radioactive release and the likelihood of adverse health effects due to radioactive release. As the accuracy of such estimates improves industry-wide, safety regulators consider weighing these calculated risks more heavily than strict adherence to the prescriptive conservatisms of existing regulations, hence risk-informed regulation. DBAs, despite their prescriptive nature, can remain useful tools for regulators and designers in a risk-informed regulatory framework, providing that they can be formulated in a fashion consistent with the risk profiles of a plant. DBAs also offer the opportunity to take into account factors of uncertainty not captured in a PRA, which are typically addressed via defense-in-depth features and subjective judgements. Designers seeking only to create a plant having a calculated risk below a certain value, while minimizing cost, may find themselves in an inefficient trial-and-error process as they

  5. Medical basis for radiation accident preparedness

    International Nuclear Information System (INIS)

    Huebner, K.F.; Fry, S.A.

    1980-01-01

    The International Conference on The Medical Basis for Radiation Accident Preparedness was organized by the staff of the Radiation Emergency Assistance Center/Training Site (REAC/TS) of the Medical and Health Sciences Division of Oak Ridge Associated Universities (ORAU). The philosophical importance of relating, through investigation and education, the intellectual resources of higher education to the important social problems associated with energy, health, and the environment was the foundation of the meeting. The symposium, held under the auspices of the US Department of Energy, was the ninth since 1960 of a series of international conferences addressing the various aspects of radiation accidents. The approach of this most recent conference differed somewhat from that of those preceding it, in that it sought an international review of the gamut of the medical aspects of radiation injury, not only for the experts in the field, but also for other physicians and scientists who, in view of current events, have had the need to know thrust upon them. Individual entries were made for the separate papers

  6. Design basis 2

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, G.; Soerensen, P. [Risoe National Lab., Roskilde (Denmark)

    1996-09-01

    Design Basis Program 2 (DBP2) is comprehensive fully coupled code which has the capability to operate in the time domain as well as in the frequency domain. The code was developed during the period 1991-93 and succeed Design Basis 1, which is a one-blade model presuming stiff tower, transmission system and hub. The package is designed for use on a personal computer and offers a user-friendly environment based on menu-driven editing and control facilities, and with graphics used extensively for the data presentation. Moreover in-data as well as results are dumped on files in Ascii-format. The input data is organized in a in-data base with a structure that easily allows for arbitrary combinations of defined structural components and load cases. (au)

  7. Evaluation on Cooling Performance of Containment Fan Cooler during Design Basis Accident with Loss of Offsite Power for Kori 3 and 4 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Bok; Lee, Sang Won [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Park, Young Chan [Atomic Creative Technology Co., LTD., Daejeon (Korea, Republic of)

    2007-10-15

    The purpose of this study is to evaluate cooling performance of containment fan cooler units and to review a technical background related to Generic Letter 96-06. In case that design basis accident (DBA) and loss of offsite power (LOOP) occurs, component cooling water (CCW) pumps cannot provide the cooling water source to fan cooler units while fan coolers coast down. Fan cooler units and CCW pumps are restarted by emergency diesel generator (EDG) operation and it takes about 30 seconds. In this scenario, before the EDG restarts and CCW flowrate is restored, heated air in the containment passes through coil of fan cooler units without cooling water source. In this situation, the boiling of water in the fan cooler units may occur. Restarting of CCW pumps may bring about condensation by injected cooling water and water hammer may occur. This thermal-hydraulic effect is sensitive to system configuration, i.e system pressure, containment pressure/temperature, EDG restarting time, etc. In this study, the evaluation of containment fan cooler units was performed for Kori 3 and 4 nuclear power plant.

  8. Effects of non-latching blast valves on the source term and consequences of the design-basis accidents in the Device Assembly Facility (DAF)

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1993-08-01

    The analysis of the Design-Basis Accidents (DBA) involving high explosives (HE) and Plutonium (Pu) in the assembly cell of the Device Assembly Facility (DAF), which was completed earlier, assumed latching blast valves in the ventilation system of the assembly cell. Latching valves effectively sealed a release path through the ventilation duct system. However, the blast valves in the assembly cell, as constructed are actually non-latching valves, and would reopen when the gas pressure drops to 0.5 psi above one atmosphere. Because the reopening of the blast valves provides an additional release path to the environment, and affects the material transport from the assembly cell to other DAF buildings, the DOE/NV DAF management has decided to support an additional analysis of the DAF's DBA to account for the effects of non-latching valves. Three cases were considered in the DAF's DBA, depending on the amount of HE and Pu involved, as follows: Case 1 -- 423 number-sign HE, 16 kg Pu; Case 2 -- 150 number-sign HE 10 kg Pu; Case 3 -- 55 number-sign HE 5 kg Pu. The results of the analysis with non-latching valves are summarized

  9. System thermalhydraulics for design basis accident analysis and simulation: Status of tools and methods and direction for future R&D

    Energy Technology Data Exchange (ETDEWEB)

    Bestion, D., E-mail: dominique.bestion@cea.fr

    2017-02-15

    Highlights: • A state of the art on system code application is presented. • Requirements for demonstration of code up-scaling capabilities are proposed. • The role of multi-scale analysis in safety analysis is explained. • Uncertainty quantifications methodologies for system codes and CFD codes are compared and discussed. - Abstract: System thermalhydraulic investigations of Design Basis Accident require several tools and methods including the Process Identification and Ranking Table, the scaling, experiment analysis, modelling, code development, code Validation and Verification, and Uncertainty Quantification. This paper intends to give an overview of these methods and tools showing what the state of the art is, and presenting some recent advances. Recommendations are made with future direction for R&D including the need of new advanced experiments and instrumentation, and the future role of CFD and multi-scale analyses. For many people it is not clear what current system codes are, and what they can be. Then the main characteristics of these codes are recalled and propositions are made to clarify the code capabilities and limitations and to improve the knowledge of the conditions for a correct application of the codes for safety in a Best Estimate Plus Uncertainty approach. Also, the on-going developments of 3-field models and Transport of Interfacial Area are summarized and associated experimental needs are identified. The growing role of 3D modelling of reactor core and Pressure Vessel requires additional experimental data for a proper validation. CFD in open medium also contributes to investigations when 3D geometrical aspects play an important role. Recent activities performed in the OECD-NEA Working Group for Analysis and Management of Accidents is summarized and recent applications of two-phase CFD to boiling flows and two-phase PTS scenarios are reported. The role of a multi-scale approach for safety issues is illustrated with the LOCA transients

  10. Simulations of the design basis accident at conditions of power increase and the o transient of MSIV at overpressure conditions of the Laguna Verde Power Station

    International Nuclear Information System (INIS)

    Araiza M, E.; Nunez C, A.

    2001-01-01

    This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the containment is not challenged during a loss of coolant accident at uprate power conditions.The analysis of the main steam isolation valve transients at overpressure conditions, and the analysis of the particular cases of the failure of one to six safety relief valves to open, show that the vessel peak pressures are below the design pressure and have no significant effect on vessel integrity. (Author)

  11. Consideration of severe accidents in design of advanced WWER reactors

    International Nuclear Information System (INIS)

    Fedorov, V.G.; Rogov, M.F.; Podshibyakin, A.K.; Fil, N.S.; Volkov, B.E.; Semishkin, V.P.

    1998-01-01

    Severe accident related requirements formulated in General Regulations for Nuclear Power Plant Safety (OPB-88), in Nuclear Safety Regulations for Nuclear Power Stations' Reactor Plants (PBYa RU AS-89) and in other NPP nuclear and radiation guides of the Russian Gosatomnadzor are analyzed. In accordance with these guides analyses of beyond design basis accidents should be performed in the reactor plant design. Categorization of beyond design basis accidents leading to severe accidents should be made on occurrence probability and severity of consequences. Engineered features and measures intended for severe accident management should be provided in reactor plant design. Requirements for severe accident analyses and for development of measures for severe accident management are determined. Design philosophy and proposed engineered measures for mitigation of severe accidents and decrease of radiation releases are demonstrated using examples of large, WWER-1000 (V-392), and medium size WWER-640 (V-407) reactor plant designs. Mitigation of severe accidents and decrease of radiation releases are supposed to be conducted on basis of consistent realization of the defense in depth concept relating to application of a system of barriers on the path of spreading of ionizing radiation and radioactive materials to the environment and a set of engineered measures protecting these barriers and retaining their effectiveness. Status of fulfilled by OKB Gidropress and other Russian organizations experimental and analytical investigations of severe accident phenomena supporting design decisions and severe accident management procedures is described. Status of the works on retention of core melt inside the WWER-640 reactor vessel is also characterized

  12. Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

    International Nuclear Information System (INIS)

    Lázaro, A.; Ammirabile, L.; Bandini, G.; Darmet, G.; Massara, S.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Mikityuk, K.; Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Stempniewicz, M.

    2014-01-01

    Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes

  13. Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Lázaro, A., E-mail: aurelio.lazaro-chueca@ec.europa.eu [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); UPV—Universidad Politecnica de Valencia, Cami de vera s/n-46002, Valencia (Spain); Ammirabile, L. [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Massara, S. [EDF, 1 avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2-28006 Madrid (Spain); Mikityuk, K. [PSI—Paul Scherrer Institut, 5232 Villigen Switzerland (Switzerland); Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R. [KIT—Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen Germany (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, PO Box 9034 6800 ES, Arnhem (Netherlands)

    2014-01-15

    Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

  14. Accident considerations in LMFBR design

    International Nuclear Information System (INIS)

    Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

    1975-12-01

    LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

  15. Interior design conceptual basis

    CERN Document Server

    Sully, Anthony

    2015-01-01

    Maximizing reader insights into interior design as a conceptual way of thinking, which is about ideas and how they are formulated. The major themes of this book are the seven concepts of planning, circulation, 3D, construction, materials, colour and lighting, which covers the entire spectrum of a designer’s activity. Analysing design concepts from the view of the range of possibilities that the designer can examine and eventually decide by choice and conclusive belief the appropriate course of action to take in forming that particular concept, the formation and implementation of these concepts is taken in this book to aid the designer in his/her professional task of completing a design proposal to the client. The purpose of this book is to prepare designers to focus on each concept independently as much as possible, whilst acknowledging relative connections without unwarranted influences unfairly dictating a conceptual bias, and is about that part of the design process called conceptual analysis. It is assu...

  16. Severe Accident Test Station Design Document

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Keiser, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  17. Severe Accident Test Station Design Document

    International Nuclear Information System (INIS)

    Snead, Mary A.; Yan, Yong; Howell, Michael; Keiser, James R.; Terrani, Kurt A.

    2015-01-01

    The purpose of the ORNL severe accident test station (SATS) is to provide a platform for evaluation of advanced fuels under projected beyond design basis accident (BDBA) conditions. The SATS delivers the capability to map the behavior of advanced fuels concepts under accident scenarios across various temperature and pressure profiles, steam and steam-hydrogen gas mixtures, and thermal shock. The overall facility will include parallel capabilities for examination of fuels and irradiated materials (in-cell) and non-irradiated materials (out-of-cell) at BDBA conditions as well as design basis accident (DBA) or loss of coolant accident (LOCA) conditions. Also, a supporting analytical infrastructure to provide the data-needs for the fuel-modeling components of the Fuel Cycle Research and Development (FCRD) program will be put in place in a parallel manner. This design report contains the information for the first, second and third phases of design and construction of the SATS. The first phase consisted of the design and construction of an out-of-cell BDBA module intended for examination of non-irradiated materials. The second phase of this work was to construct the BDBA in-cell module to test irradiated fuels and materials as well as the module for DBA (i.e. LOCA) testing out-of-cell, The third phase was to build the in-cell DBA module. The details of the design constraints and requirements for the in-cell facility have been closely captured during the deployment of the out-of-cell SATS modules to ensure effective future implementation of the in-cell modules.

  18. Design basis II: Design for events

    International Nuclear Information System (INIS)

    Frisch, W.

    1982-01-01

    In a lecture of this title, it could be expected that all events which are a basis for system and component design are described. According to the title of the Course 'Instrumentation and Control of Nuclear Power Plants' emphasis is put on events originating within the plant (no consideration of external events such as air plane crash or earth-quake). The lecture is divided into the two parts 'Transients' and 'Loss of coolant accidents (LOCAs)'. Due to the complex interaction between systems and components during transients, the first part is the main part of the lecture, while the second part (LOCAs) is only a very brief description of emergency core cooling system functions and the typical course of a large and small LOCA event. The first part on anticipated transients with intact primary coolant system boundary (non-LOCA-transients) covers several aspects of the analysis, such as classification, brief system description, transient description, analysis of anticipated transients without scram (ATWS) and analytical methods. Due to the time restriction necessary within the course, only a small section of the entire area can be presented in this paper. (orig.)

  19. Experience in construction of new safe confinement over ChNPP-4 damages in beyond design-basis accident: view in 30 year

    International Nuclear Information System (INIS)

    Nosovskij, A.V.

    2016-01-01

    Many organizations and institutions participated in the elimination of ChNPP-4 accident. However, the main efforts on conservation of the damaged unit and ChNPP-3 commissioning were performed in 1986-1987 by experts of the enterprises and organizations of the Ministry of Medium Machine Building of the Soviet Union, who have been sent to the staff of the specially created Construction Administration US-605. The paper presents the activity of US-605 experts, describes administrative and technical measures of radiation safety during construction of the Shelter in difficult radiation conditions. Such efforts enabled accumulation of significant and unique experience in mitigation of severe accident consequences, which shall be used to prevent any nuclear accidents and eliminate their consequences

  20. An analysis of uncertainty and of dependence on season of year of ingestion population dose arising from design basis accidents in advanced thermal reactors

    International Nuclear Information System (INIS)

    Nair, S.; Ponting, A.C.

    1985-03-01

    The results of a detailed study of ingestion collective dose from five limiting PWR design basis releases are presented, the PWR being chosen as being typical of an advanced thermal reactor for which source terms are readily available. The ingestion collective dose was calculated for a range of wind direction/weather scenarios for releases from a typical U.K. rural and a U.K. semi-urban site and scenarios identified where the ingestion pathway was of potential significance. The dependence of the ingestion collective dose for these cases on the season of year when the release occurs was investigated. An analysis was carried out of the uncertainty in the ''worst case'' ingestion calculations arising from uncertainties in foodchain input parameters. An efficient but comprehensive set of dynamic foodchain computer models was produced and the literature surveyed to produce probability distribution functions (PDF's) for all relevant independent input data items. These were used to produce output PDF's for food contamination levels and for ingestion collective dose from the five releases. Finally, the study has highlighted several areas central to ingestion collective dose assessments where the available data are inadequate. This led to the formulation of a set of future research requirements which will need to be met both to obtain a better fundamental understanding of foodchain transfer and to reduce uncertainties in ingestion collective dose assessments. (author)

  1. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    In 1985 an incident at Toledo Edison's Davis Besse plant caused the U.S. Nuclear Regulatory Commission (NRC) to re-evaluate the technical information that the utilities had readily available to support the design of their plants. The Design Basis programs, currently on going in most U.S. utilities, have been the nuclear industry's response to the needs identified by this re-evaluation. In order to understand the Design Basis programs which have been implemented by the U.S. nuclear utilities, it is necessary to understand the problem as it was perceived by the nuclear industry (the utilities, the original NSSS designers and the regulators) after the Davis-Besse incident, the subsequent programs undertaken by the industry under the leadership of INPO and NUMARC, the NRC's actions, and the overall evolution of the industry's vision in relation to this problem. This paper presents the history of the design basis efforts from the first recognition of the problem by the NRC after the Davis-Besse incident, describes the actions taken by the NRC, INPO, NUMARC, the U.S. utilities and the NSSS designers, and brings the problem statement up-to-date in relation to the vision presently held by the U.S. nuclear industry. It then presents a technical discussion to develop a detailed definition of design basis information to support the problem statement. The information originally supplied by the NSSS designers during the plant design and construction is discussed as well as its relationship to the previously defined design basis information. This section of the paper concludes by defining the additional information needed by nuclear utilities to satisfy the requirements developed from the problem statement. Having developed a definition of the additional information (i.e., information not originally supplied during design and construction) required to solve the design basis problem as it is presently perceived by the U.S. nuclear industry, the paper then discusses design basis

  2. Factors correlated with traffic accidents as a basis for evaluating Advanced Driver Assistance Systems.

    Science.gov (United States)

    Staubach, Maria

    2009-09-01

    This study aims to identify factors which influence and cause errors in traffic accidents and to use these as a basis for information to guide the application and design of driver assistance systems. A total of 474 accidents were examined in depth for this study by means of a psychological survey, data from accident reports, and technical reconstruction information. An error analysis was subsequently carried out, taking into account the driver, environment, and vehicle sub-systems. Results showed that all accidents were influenced by errors as a consequence of distraction and reduced activity. For crossroad accidents, there were further errors resulting from sight obstruction, masked stimuli, focus errors, and law infringements. Lane departure crashes were additionally caused by errors as a result of masked stimuli, law infringements, expectation errors as well as objective and action slips, while same direction accidents occurred additionally because of focus errors, expectation errors, and objective and action slips. Most accidents were influenced by multiple factors. There is a safety potential for Advanced Driver Assistance Systems (ADAS), which support the driver in information assimilation and help to avoid distraction and reduced activity. The design of the ADAS is dependent on the specific influencing factors of the accident type.

  3. Severe accidents risk assessment as a basis for emergency preparedness

    International Nuclear Information System (INIS)

    Sinka, D.; Mikulicic, V.

    2000-01-01

    The paper demonstrates, by example of the Republic of Croatia, the possibilities of implementing risk assessment as basis for nuclear accident emergency preparedness development. Individual risks of severe accidents for citizens of the biggest Croatian population centers, as well as collective risk for entire population have been assessed using the PRONEL method. The assessment covered 90 power reactors located at a distance up to 1.000 km. The conducted assessment shows the risks for various regions of the Republic of Croatia, and comparison between them. If risk would be taken as basic criterion in nuclear emergency planning, the results of assessment would directly indicate the necessary preparation level for each region. Furthermore, the assessment of risks from individual power plants and power plant types indicates to which facilities the greatest attention should be paid in nuclear accidents preparedness development. Risks from groups of power plants formed in accordance with their respective distance from exposure location shows what kind of tools for determining consequences and protective actions during a nuclear accident should be made available. (author)

  4. System Design and the Safety Basis

    International Nuclear Information System (INIS)

    Ellingson, Darrel

    2008-01-01

    The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination and decommissioning (D and D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities

  5. Technical Details on Beyond Design Basis Event Pilot Evaluations

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2013-01-01

    The primary focus of the BDBE pilot project was the review of BDBE analysis and mitigation features at four DOE nuclear facilities representing a range of DOE sites, nuclear facility types/activities, and responsible program offices. The pilots looked at (1) how beyond design basis accidents were evaluated and documented in the facility Documented Safety Analysis, (2) potential BDBE vulnerabilities and margins to failure of facility safety features as obtained from general area and specific system walkdowns and design documents reviews, and (3) preparations made in facility and site emergency management programs to respond to severe accidents. It also evaluated whether draft BDBE guidance on safety analysis and emergency management could be used to improve the analysis of and preparations for mitigating severe and beyond design basis accidents. The details of these activities are organized in this report as described below.

  6. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    Il'kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I.; Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K.; Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A.; Haire, Jonathan M.; Forsberg, C.W.

    2004-01-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  7. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  8. International Peer Reviews of Design Basis

    International Nuclear Information System (INIS)

    Hughes, Peter

    2013-01-01

    International peer reviews: Design and safety assessment review service: - Review of design requirements; - Review in support of licensing; - Review in support of severe accident management; - Review in support of modifications; - Review in relation to periodic safety, or life extension; - Reviews take place at any time in NPP lifecycle from concept, through design and operations

  9. Severe accident mitigation through containment design

    International Nuclear Information System (INIS)

    Bergeron, K.D.

    1990-01-01

    Recent US Department of Energy plans to construct a Heavy Water Reactor for the production of defense nuclear materials have created a unique opportunity to explore ways to mitigate severe accident concerns in the design stage. Drawing on an extensive background in USNRC-sponsored severe accident work, Sandia National Laboratories has been exploring a number of Heavy Water New Production Reactor (HW-NPR) containment design strategies that might mitigate the consequences of a core-melt accident without greatly impacting construction cost or reactor operations. Severe accident specialists have undertaken these assessments with the intent of providing the plant designers with some of the phenomenological advantages and disadvantages of various mitigation strategies. This paper will highlight some of the more interesting concepts and summarize the results obtained. 9 refs., 2 tabs

  10. Severe accident mitigation through containment design

    International Nuclear Information System (INIS)

    Bergeron, K.D.

    1990-01-01

    Recent U.S. Department of Energy plans to construct a Heavy Water Reactor for the production of defense nuclear materials have created a unique opportunity to explore ways to mitigate severe accident concerns in the design stage. Drawing on an extensive background in US-NRC-sponsored severe accident work, Sandia National Laboratories has been exploring a number of Heavy Water New Production Reactor (HW-NPR) containment design strategies that might mitigate the consequences of a core-melt accident without greatly impacting construction cost or reactor operations. Severe accident specialists have undertaken these assessments with the intent of providing the plant designers with some of the phenomenological advantages and disadvantages of various mitigation strategies. This paper will highlight some of the more interesting concepts and summarize the results obtained. (author). 9 refs., 2 tabs

  11. 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.

  12. 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.

  13. Considerations of severe accidents in the design of Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Dong Wook Jerng; Choong Sup Byun

    1998-01-01

    The severe accident is one of the key issues in the design of Korean Next Generation Reactor (KNGR) which is an evolutionary type of pressurized water reactor. As IAEA recommends in TECDOC-801, the design objective of KNGR with regard to safety is provide a sound technical basis by which an imminent off-site emergency response to any circumstance could be practically unnecessary. To implement this design objective, probabilistic safety goals were established and design requirements were developed for systems to mitigate severe accidents. The basic approach of KNGR to address severe accidents is firstly prevent severe accidents by reinforcing its capability to cope with the design basis accidents (DBA) and further with some accidents beyond DBAs caused by multiple failures, and secondly mitigate severe accidents to ensure the retention of radioactive materials in the containment by providing mean to maintain the containment integrity. For severe accident mitigation, KNGR principally takes the concept of ex-vessel corium cooling. To implement this concept, KNGR is equipped with a large cavity and cavity flooding system connected to the in-containment refueling water storage tank. Other major systems incorporated in KNGR are hydrogen igniters and safety depressurization systems. In addition, the KNGR containment is designed to withstand the pressure and temperature conditions expected during the course of severe accidents. In this paper, the design features and status of system designs related with severe accidents will be presented. Also, R and D activities related to severe accident mitigation system design will be briefly described

  14. Understanding and capturing NSSS design basis

    International Nuclear Information System (INIS)

    Palo, W.J.; Miller, B.

    1993-01-01

    Changes to, and technical evaluations of nuclear generating station designs are often warranted. Comprehensive documentation and understanding of the NSSS Design Basis are essential to support these activities. Effective configuration management tools are also needed to maintain the plant within design basis limits. Efficient design basis reconstitution can be realized via: In-depth understanding of the design process; Utilization of effective data collection methodology; State of the art data basing tools. A database can be created to generate a Design Basis Manual (DBM). This database can communicate electronically with other plant databases. A living document vice a static snapshot of the plant design is the goal. A design basis database can serve as the cornerstone for a global electronic information control system

  15. SAFETY BASIS DESIGN DEVELOPMENT CHALLENGES IMECE2007-42747

    International Nuclear Information System (INIS)

    RYAN GW

    2007-01-01

    'Designing in Safety' is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation IOCFR830, Nuclear Safety Management. IOCFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses 'Safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE's Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, (1) flashing spray leaks and (2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents

  16. SAFETY BASIS DESIGN DEVELOPMENT CHALLENGES IMECE2007-42747

    Energy Technology Data Exchange (ETDEWEB)

    RYAN GW

    2007-09-24

    'Designing in Safety' is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation IOCFR830, Nuclear Safety Management. IOCFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses 'Safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE's Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, (1) flashing spray leaks and (2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents.

  17. 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

  18. Technical basis for nuclear accident dosimetry at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kerr, G.D.; Mei, G.T.

    1993-08-01

    The Oak Ridge National Laboratory (ORNL) Environmental, Safety, and Health Emergency Response Organization has the responsibility of providing analyses of personnel exposures to neutrons and gamma rays from a nuclear accident. This report presents the technical and philosophical basis for the dose assessment aspects of the nuclear accident dosimetry (NAD) system at ORNL. The issues addressed are regulatory guidelines, ORNL NAD system components and performance, and the interpretation of dosimetric information that would be gathered following a nuclear accident

  19. Determination of Design Basis Earthquake ground motion

    International Nuclear Information System (INIS)

    Kato, Muneaki

    1997-01-01

    This paper describes principle of determining of Design Basis Earthquake following the Examination Guide, some examples on actual sites including earthquake sources to be considered, earthquake response spectrum and simulated seismic waves. In sppendix of this paper, furthermore, seismic safety review for N.P.P designed before publication of the Examination Guide was summarized with Check Basis Earthquake. (J.P.N.)

  20. Determination of Design Basis Earthquake ground motion

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Muneaki [Japan Atomic Power Co., Tokyo (Japan)

    1997-03-01

    This paper describes principle of determining of Design Basis Earthquake following the Examination Guide, some examples on actual sites including earthquake sources to be considered, earthquake response spectrum and simulated seismic waves. In sppendix of this paper, furthermore, seismic safety review for N.P.P designed before publication of the Examination Guide was summarized with Check Basis Earthquake. (J.P.N.)

  1. Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

    International Nuclear Information System (INIS)

    G. L. Sharp; R. T. McCracken

    2004-01-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

  2. The impact of safety design consideration on future LMFBR developments. (R and D needs related to accident accommodation)

    International Nuclear Information System (INIS)

    Justin, F.

    1985-04-01

    Accident accommodation for design accidents or even beyond design basis accidents is based on components and systems for which important research and development work is needed. Main issues are treated: fuel failure faults, sodium fires, decay heat removal, accommodation of energetics and debris

  3. Planning on a regional basis for a major radiation accident

    International Nuclear Information System (INIS)

    Casey, W.R.

    1981-01-01

    As a part of the Radiological Assistance Program, members of the Safety and Environmental Protection Division of Brookhaven National Laboratory have served as a response team for many years to the northeastern section of the United States. During this time, responses have been made to several significant incidents, including the accident at Three Mile Island. The planning and preparation for emergency response activities will be discussed. Included will be a review of instrument requirements, analytical and support equipment, modes of response, and communication needs. Interaction with and support from other response teams will be discussed. In particular, the lessons from the respone to Three Mile Island will be reviewed

  4. Exploratory Shaft Facility design basis study report

    International Nuclear Information System (INIS)

    Langstaff, A.L.

    1987-01-01

    The Design Basis Study is a scoping/sizing study that evaluated the items concerning the Exploratory Shaft Facility Design including design basis values for water and methane inflow; flexibility of the design to support potential changes in program direction; cost and schedule impacts that could result if the design were changed to comply with gassy mine regulations; and cost, schedule, advantages and disadvantages of a larger second shaft. Recommendations are proposed concerning water and methane inflow values, facility layout, second shaft size, ventilation, and gassy mine requirements. 75 refs., 3 figs., 7 tabs

  5. Design Characteristics as Basis for Design Languages

    DEFF Research Database (Denmark)

    Mortensen, Niels Henrik

    1997-01-01

    The application of modern feature based CAD systems has in many companies lead to significant rationalisation of design, particulary the "down stream" acticities such as NC code generation, FEM analysis, mould flow simulation and documentation. The subject of this paper is the "up stream" activit......The application of modern feature based CAD systems has in many companies lead to significant rationalisation of design, particulary the "down stream" acticities such as NC code generation, FEM analysis, mould flow simulation and documentation. The subject of this paper is the "up stream...

  6. FRG conceptual design and design basis

    International Nuclear Information System (INIS)

    Roethemeyer, H.

    1979-01-01

    For the site-independent conceptual design the following requirements have been laid down: (1) for safety reasons retrievability is not considered; (2) standard mining techniques and experience gained at Asse should be used; (3) two shafts should be sufficient; (4) different waste forms and containers shall be disposed of in different storage areas; (5) ventilated sections must allow the shutting off of each storage area from the rest of the mine; (6) the mining method of retreat working should be applied; (7) the mine works shall have a lateral safety distance to the caprock of 200 m and a vertical safety zone beneath salt level of 300 m; (8) all disposal areas shall be on one level; (9) salt and waste shall be transported in different drifts, mainly in a one way system

  7. Initial basis for agronomic countermeasure selection following a nuclear accident

    International Nuclear Information System (INIS)

    Bonetto, Juan P.; Kunst, Juan J.; Bruno, Hector; Jordan, Osvaldo; Hernandez, Daniel

    2008-01-01

    During the recovery stage, following a nuclear accident, application of agricultural countermeasures will be relevant to the minimization of the radiation induced detriment due to ingestion of locally produced contaminated foodstuff, as long as the magnitude of the averted dose is sufficient to justify their implementation. Nuclear emergency planning in Argentina currently holds food ban as the accepted countermeasure, at least until other measures are taken. Though it may ensure no residual collective dose, food ban may also imply very high costs, compared to other alternatives, specially due to the need of disposing off perishable food such as milk. Therefore, an exhaustive evaluation of all the alternatives, considering both quantitative and qualitative factors is still needed to identify optimal countermeasure strategies, bearing in mind also that decisions made during the early phase of an emergency will affect the fate of the measures to be taken later. As a first step in this direction, a basic quantitative decision-aiding technique, the cost-benefit analysis, is carried out for comparison of countermeasures related to Cesium contaminated cow-milk which are considered feasible for implementation in Argentina. Countermeasures total costs are estimated from various local sources, while their effectiveness are adopted from international bibliography. At this stage, a simple theoretical example considering milk contamination in the surroundings of the Embalse Nuclear Power Plant is used for a generic analysis, since actual collective doses and costs can only be calculated for a specific modelled scenario. (author)

  8. Initial Basis for Agronomic Countermeasure Selection Following a Nuclear Accident

    International Nuclear Information System (INIS)

    Bonetto, J.P.; Kunst, J.J.; Bruno, H.A.; Jordan, O.D.; Hernandez, D.G.

    2011-01-01

    During the recovery stage, following a nuclear accident, application of agricultural countermeasures will be relevant to the minimization of the radiation induced detriment due to ingestion of locally produced contaminated foodstuff, as long as the magnitude of the averted dose is sufficient to justify their implementation. Nuclear emergency planning in Argentina currently holds food ban as the accepted countermeasure, at least until other measures are taken. Though it may ensure no residual collective dose, food ban may also imply very high costs, compared to other alternatives, specially due to the need of disposing off perishable food such as milk. Therefore, an exhaustive evaluation of all the alternatives, considering both quantitative and qualitative factors is still needed to identify optimal countermeasure strategies, bearing in mind also that decisions made during the early phase of an emergency will affect the fate of the measures to be taken later. As a first step in this direction, a basic quantitative decis sion-aiding technique, the cost-benefit analysis, is carried out for comparison of countermeasures related to Cesium contaminated cow-milk which are considered feasible for implementation in Argentina. Countermeasures total costs are estimated from various local sources, while their effectiveness are adopted from international bibliography. At this stage, a simple theoretical example considering milk contamination in the surroundings of the Embalse Nuclear Power Plant is used for a generic analysis, since actual collective doses and costs can only be calculated for a specific modelled scenario. (authors)

  9. PRELIMINARY SELECTION OF MGR DESIGN BASIS EVENTS

    International Nuclear Information System (INIS)

    Kappes, J.A.

    1999-01-01

    The purpose of this analysis is to identify the preliminary design basis events (DBEs) for consideration in the design of the Monitored Geologic Repository (MGR). For external events and natural phenomena (e.g., earthquake), the objective is to identify those initiating events that the MGR will be designed to withstand. Design criteria will ensure that radiological release scenarios resulting from these initiating events are beyond design basis (i.e., have a scenario frequency less than once per million years). For internal (i.e., human-induced and random equipment failures) events, the objective is to identify credible event sequences that result in bounding radiological releases. These sequences will be used to establish the design basis criteria for MGR structures, systems, and components (SSCs) design basis criteria in order to prevent or mitigate radiological releases. The safety strategy presented in this analysis for preventing or mitigating DBEs is based on the preclosure safety strategy outlined in ''Strategy to Mitigate Preclosure Offsite Exposure'' (CRWMS M andO 1998f). DBE analysis is necessary to provide feedback and requirements to the design process, and also to demonstrate compliance with proposed 10 CFR 63 (Dyer 1999b) requirements. DBE analysis is also required to identify and classify the SSCs that are important to safety (ITS)

  10. EPR design features to mitigate severe accident challenges

    International Nuclear Information System (INIS)

    Mazurkiewicz, S.M.; Fischer, M.; Bittermann, D.

    2005-01-01

    The EPR, an evolutionary pressurized water reactor (PWR), is a 4300-4500 MWth that incorporates proven technology within an optimized configuration to enhance safety. EPR was originally developed through a joint effort between Framatome ANP and Siemens by incorporating the best technological features from the French and German nuclear reactor fleets into a cost-competitive product. Commercial EPR units are currently being built in Finland at the Olkiluoto site, and planned for France at the Flamanville site. In recent months, Framatome ANP announced their intention to market the EPR units to China in response to a request for vendor bids as well as their intent to pursue design certification in the United States under 10CFR52. The EPR safety philosophy is based on a deterministic consideration of defense-in-depth complemented by probabilistic analyses. Not only is the EPR designed to prevent and mitigate design basis accidents (DBAs), it employs an extra level of safety associated with severe accident response. Therefore, as a design objective, features are included to ensure that radiological consequences are limited such that the need for stringent counter measures, such as evacuation and relocation of the nearby population, can be reasonably excluded. This paper discusses some of the innovative features of the EPR to address severe accident challenges. (author)

  11. ACR-1000 design provisions for severe accidents

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

  12. Emergency procedures beyond design basis ''Feed and Bleed''

    International Nuclear Information System (INIS)

    Dominguez Bautista, M.T.; Campuzano Pena, F.

    1994-01-01

    The incorporation of Beyond-Design-Basis Emergency Procedures, also called the Emergency Manual or Severe Accident Manual, has been an important step forward in nuclear power plant safety. These procedures cover situations in which the deterministic criteria used in plant design have been contravened. In such situations new accident scenarios, unforeseen system actions or a combination of both, need to be considered. Establishing these procedures is actually the last in a sequence of activities the sequence includes definition of scenarios, study of their phenomena, analysis of optional system actions, verification of their effectiveness and finally, implementation of the procedure. The systematization of these new strategies is supported by the results of the probabilistic analyses which serve in this case to pinpoint the objectives of these strategies. This paper describes the application of this methodology in the definition of a procedure for heat sink recovery on the secondary side (feed and bleed) if this has been totally or partially lost in a beyond-design-basis event. (Author)

  13. Statistical evaluation of design-error related nuclear reactor accidents

    International Nuclear Information System (INIS)

    Ott, K.O.; Marchaterre, J.F.

    1981-01-01

    In this paper, general methodology for the statistical evaluation of design-error related accidents is proposed that can be applied to a variety of systems that evolves during the development of large-scale technologies. The evaluation aims at an estimate of the combined ''residual'' frequency of yet unknown types of accidents ''lurking'' in a certain technological system. A special categorization in incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of U.S. nuclear power reactor technology, considering serious accidents (category 2 events) that involved, in the accident progression, a particular design inadequacy. 9 refs

  14. Prototype Hanford Surface Barrier: Design basis document

    International Nuclear Information System (INIS)

    Myers, D.R.; Duranceau, D.A.

    1994-11-01

    The Hanford Site Surface Barrier Development Program (BDP) was organized in 1985 to develop the technology needed to provide a long-term surface barrier capability for the Hanford Site and other arid sites. This document provides the basis of the prototype barrier. Engineers and scientists have momentarily frozen evolving barrier designs and incorporated the latest findings from BDP tasks. The design and construction of the prototype barrier has required that all of the various components of the barrier be brought together into an integrated system. This integration is particularly important because some of the components of the protective barreir have been developed independently of other barreir components. This document serves as the baseline by which future modifications or other barrier designs can be compared. Also, this document contains the minutes of meeting convened during the definitive design process in which critical decisions affecting the prototype barrier's design were made and the construction drawings

  15. Key issues on safety design basis selection and safety assessment

    International Nuclear Information System (INIS)

    An, S.; Togo, Y.

    1976-01-01

    In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

  16. Establishing 'design basis threat' in Norway

    International Nuclear Information System (INIS)

    Maerli, M.B.; Naadland, E.; Reistad, O.

    2002-01-01

    Full text: INFCIRC 225 (Rev. 4) assumes that a state's physical protection system should be based on the state's evaluation of the threat, and that this should be reflected in the relevant legislation. Other factors should also be considered, including the state's emergency response capabilities and the existing and relevant measures of the state's system of accounting for and control of nuclear material. A design basis threat developed from an evaluation by the state of the threat of unauthorized removal of nuclear material and of sabotage of nuclear material and nuclear facilities is an essential element of a state's system of physical protection. The state should continuously review the threat, and evaluate the implications of any changes in that threat for the required levels and the methods of physical protection. As part of a national design basis threat assessment, this paper evaluates the risk of nuclear or radiological terrorism and sabotage in Norway. Possible scenarios are presented and plausible consequences are discussed with a view to characterize the risks. The need for more stringent regulatory requirements will be discussed, together with the (positive) impact of improved systems and procedures of physical protection on nuclear emergency planning. Special emphasis is placed on discussing the design basis threat for different scenarios in order to systemize regulatory efforts to update the current legislation, requirement for operators' contingency planning, response efforts and the need for emergency exercises. (author)

  17. System 80+ design features for severe accident prevention and mitigation

    International Nuclear Information System (INIS)

    Jacob, M.C.; Schneider, R.E.; Finnicum, D.J.

    1993-01-01

    ABB-CE, in cooperation with the US Department of Energy, is working to develop and certify the System 80+ design, which is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the EPRI's Utility Requirements Document, and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the system is discussed along with its conformance to EPRI URD guidance, as applicable. Computer simulation of a best estimate severe accident scenario is presented to illustrate the acceptable containment performance of the design. It is concluded that by considering severe accident prevention and mitigation early in the design process, the System 80+ design represents a robust plant design that has low core damage frequencies, low containment conditional failure probabilities, and acceptable deterministic containment performance under severe accident conditions

  18. Design Load Basis for Offshore Wind turbines

    DEFF Research Database (Denmark)

    Natarajan, Anand; Hansen, Morten Hartvig; Wang, Shaofeng

    2016-01-01

    DTU Wind Energy is not designing and manufacturing wind turbines and does therefore not need a Design Load Basis (DLB) that is accepted by a certification body. However, to assess the load consequences of innovative features and devices added to existing offshore turbine concepts or new offshore...... turbine concept developed in our research, it is useful to have a full DLB that follows the current design standard and is representative of a general DLB used by the industry. It will set a standard for the offshore wind turbine design load evaluations performed at DTU Wind Energy, which is aligned...... with the challenges faced by the industry and therefore ensures that our research continues to have a strong foundation in this interaction. Furthermore, the use of a full DLB that follows the current standard can improve and increase the feedback from the research at DTU Wind Energy to the international...

  19. Analysis of regulatory requirement for beyond design basis events of SMART

    International Nuclear Information System (INIS)

    Kim, W. S.; Seol, K. W.

    2000-01-01

    To enhance the safety of SMART reactor, safety and regulatory requirements associated with beyond design basis events (beyond BDE), which were developed and applied to advanced light water reactor designs, were analyzed along with a design status of passive reactor. And, based on these requirements, their applicability on the SMART design was evaluated. In the design aspect, severe accident prevention and mitigation features, containment performance, and accident management were analyzed. The evaluation results show that the requirement related to beyond DBE such as ATWS, loss of residual heat removal during shutdown operation, station blackout, fire, inter-system LOCA, and well-known events from severe accident phenomena is applicable to the SMART design. However, comprehensive approach against beyond DBE is not yet provided in the SMART design, and then it is required to designate and analyze the beyond DBE-related features. This study is expected to contribute to efforts to improve plant safety and to establish regulatory requirements for safety review

  20. [Basis for designing a medical course curriculum].

    Science.gov (United States)

    Villarreal, R; Bojalil, L F; Mercer, H

    1977-01-01

    This article sets forth the reasons for the structure given to the Division of Biology and Health on the Xochimilco campus of Metropolitan Autonomous University in Mexico: to adjust the university to the process of social change going forward in the country and gear the university to the problems of the present by avoiding the rigidity of its structure. The basic aspects of curriculum design are cited against a background of an historical analysis of the socioeconomic structure of education and health. The principles underlying the curriculum and the course work are then described on the basis of that analysis.

  1. Configuration management after design basis reconstitution

    International Nuclear Information System (INIS)

    Purcell, J.J.; Livingston, B.R.

    1991-01-01

    Over the last few years, Fort Calhoun station (FCS) has implemented a number of programs to enhance plant operability and readiness. The design basis document (DBD) reconstitution project was the cornerstone of this effort. Vendor manual upgrade, operating procedures upgrade, plant equipment data-base verification, equipment labeling, and warehousing improvements were also implemented as part of this improvement program. With the completion of these programs, plant documentation was current to the baselines established by each program, and a configuration management program (CMP) was established to maintain this level of accuracy throughout the remaining life of FCS. Change control throughout the organization has been reviewed and upgraded to ensure that all changes are evaluated for impact to the design bases

  2. A probabilistic risk assessment of the LLNL Plutonium facility's evaluation basis fire operational accident

    International Nuclear Information System (INIS)

    Brumburgh, G.

    1994-01-01

    The Lawrence Livermore National Laboratory (LLNL) Plutonium Facility conducts numerous involving plutonium to include device fabrication, development of fabrication techniques, metallurgy research, and laser isotope separation. A Safety Analysis Report (SAR) for the building 332 Plutonium Facility was completed rational safety and acceptable risk to employees, the public, government property, and the environment. This paper outlines the PRA analysis of the Evaluation Basis Fire (EDF) operational accident. The EBF postulates the worst-case programmatic impact event for the Plutonium Facility

  3. Acceptable risk as a basis for design

    International Nuclear Information System (INIS)

    Vrijling, J.K.; Hengel, W. van; Houben, R.J.

    1998-01-01

    Historically, human civilisations have striven to protect themselves against natural and man-made hazards. The degree of protection is a matter of political choice. Today this choice should be expressed in terms of risk and acceptable probability of failure to form the basis of the probabilistic design of the protection. It is additionally argued that the choice for a certain technology and the connected risk is made in a cost-benefit framework. The benefits and the costs including risk are weighed in the decision process. A set of rules for the evaluation of risk is proposed and tested in cases. The set of rules leads to technical advice in a question that has to be decided politically

  4. Design and Development of a Severe Accident Training System

    International Nuclear Information System (INIS)

    Kim, Ko Ryu; Park, Sun Hee; Kim, Dong Ha

    2005-01-01

    The nuclear plants' severe accidents have two big characteristics. One is that they are very rare accidents, and the other is that they bring extreme conditions such as the high pressure and temperature in their process. It is, therefore, very hard to get the severe accident data, without inquiring that the data should be real or experimental. In fact, most of severe accident analyses rely on the simulation codes where almost all severe accident knowledge is contained. These codes are, however, programmed by the Fortran language, so that their output are typical text files which are very complicated. To avoid this kind of difficulty in understanding the code output data, several kinds of graphic user interface (GUI) programs could be developed. In this paper, we will introduce a GUI system for severe accident management and training, partly developed and partly in design stage

  5. Preliminary Design of Optimized Reactor Insulator for Severe Accident Mitigation of APR1400

    International Nuclear Information System (INIS)

    Heo, Sun; Lee, Jae-Gon; Kang, Yong-Chul

    2007-01-01

    APR1400, a Korean evolutionary advance light water reactor, has many advanced safety feature to prevent and mitigate of design basis accident (DBA) and severe accident. When reactor cooling system (RCS) fails to cooling its core, the core melted down and the molten core gathers together on bottom of reactor vessel. The molten core hurts reactor vessel and is released to containment, which raises the release of radioactive isotopes and the heating of the containment atmosphere. Finally, the corium is accumulated in the bottom of reactor cavity and it also raises the Molten Core and Concrete Interaction (MCCI) and the heating of containment atmosphere. There are two strategies to cooling molten core. Those are in-vessel retention and ex-vessel cooling. At the early stage of APR1400 design, only ex-vessel cooling which is cooling of the molten core outside the vessel after vessel failure is considered based on EPRI Utility Requirement Document (URD) for Evolutionary LWR. However, a need has been arisen to reflect current research findings on severe accident phenomena and mitigation technologies to Korean URD and IVRERVC (In-Vessel corium Retention using Ex-Reactor Vessel Cooling) was adopted APR1400. The ERVC is not considered as a licensing design basis but based on the defense-in-depth principle and safety margin basis, which is the top-tier requirement of the severe accident mitigation design as stated in the KURD. The Severe Accident Management strategy for APR1400 is intended to aid the plant operating staff to secure reactor vessel integrity in the early stage of the severe accident. As a part of a design implementation of IVR-ERVC for APR1400, we developed the preliminary design requirement, design specification and conceptual design

  6. Statistical evaluation of design-error related accidents

    International Nuclear Information System (INIS)

    Ott, K.O.; Marchaterre, J.F.

    1980-01-01

    In a recently published paper (Campbell and Ott, 1979), a general methodology was proposed for the statistical evaluation of design-error related accidents. The evaluation aims at an estimate of the combined residual frequency of yet unknown types of accidents lurking in a certain technological system. Here, the original methodology is extended, as to apply to a variety of systems that evolves during the development of large-scale technologies. A special categorization of incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of the nuclear power reactor technology, considering serious accidents that involve in the accident-progression a particular design inadequacy

  7. Taking into account a reactivity accident in research reactors design

    International Nuclear Information System (INIS)

    Abou Yehia, H.; Berry, J.L.; Sinda, T.

    1989-11-01

    The particular studies realized in France for research reactors design at a Borax accident type are described. The cases of ORPHEE and RHF reactors are particularly developed. The evolution of the studies and the conservatism used are given [fr

  8. Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-11-01

    This report develops and applies a method for estimating strong earthquake ground motion. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Specifically considered are ground motions resulting from earthquakes with magnitudes from 5 to 8, fault distances from 0 to 500 km, and frequencies from 1 to 35 Hz. The two main objectives were: (1) to develop generic relations for estimating ground motion appropriate for site screening; and (2) to develop a guideline for conducting a thorough site investigation needed to define the seismic design basis. For the first objective, an engineering model was developed to predict the expected ground motion on rock sites, with an additional set of amplification factors to account for the response of the soil column over rock at soil sites. The results incorporate best estimates of ground motion as well as the randomness and uncertainty associated with those estimates. For the second objective, guidelines were developed for gathering geotechnical information at a site and using this information in calculating site response. As a part of this development, an extensive set of geotechnical and seismic investigations was conducted at three reference sites. Together, the engineering model and guidelines provide the means to select and assess the seismic suitability of a site

  9. Design consideration on severe accident for future LWR

    International Nuclear Information System (INIS)

    Omoto, A.

    1998-01-01

    Utilities' Severe Accident Management strategies, selected based on Individual Plant Examination, are in the process of implementation for each operating plant. Activities for the next generation LWR design are going on by Utilities, NSSS vendors and Research Institutes. The proposed new designs vary from evolutionary design to revolutionary design such as the supercritical LWR. Discussion on the consideration of Severe Accident in the design of next generation LWR is being held to establish the industry's self-regulatory document on containment design and its performance, which ABWR-IER (Improved Evolutionary Reactor) on the part of BWR and Evolutionary APWR and New PWR21 on the part of PWR are expected to comply. Conceptual design study for ABWR-IER will illustrate an example of design approach for the prevention and mitigation of Severe Accident and its impact on capital cost

  10. The next nuclear power station generation: Beyond-design accident concepts, methods, and action sequence

    International Nuclear Information System (INIS)

    Asmolov, V.G.; Khakh, O.Ya.; Shashkov, M.G.

    1993-01-01

    The problem of beyond-design accidents at nuclear stations will not be solved unless a safety culture becomes a basic characteristic of all lines of activity. Only then can the danger of accidents as an objective feature of nuclear stations be eliminated by purposive skilled and responsible activities of those implementing safety. Nuclear-station safety is provided by the following interacting and complementary lines of activity: (1) the design and construction of nuclear stations by properly qualified design and building organizations; (2) monitoring and supervision of safety by special state bodies; (3) control of the station by the exploiting organization; and (4) scientific examination of safety within the above framework and by independent organizations. The distribution of the responsibilities, powers, and right in these lines should be defined by a law on atomic energy, but there is not such law in Russian. The beyond-design accident problem is a key one in nuclear station safety, as it clear from the serious experience with accidents and numerous probabilistic studies. There are four features of the state of this topic in Russia that are of major significance for managing accidents: the lack of an atomic energy law, the inadequacy of the technical standards, the lack of a verified program package for nuclear-station designs in order to calculate the beyond-design accidents and analyze risks, and a lack of approach by designers to such accidents on the basis of international recommendations. This paper gives a brief description of three-forming points in the scientific activity: the general concept of nuclear-station safety, methods of analyzing and providing accident management, and the sequence of actions developed by specialists at this institute in recent years

  11. A probabilistic risk assessment of the LLNL Plutonium Facility's evaluation basis fire operational accident. Revision 1

    International Nuclear Information System (INIS)

    Brumburgh, G.P.

    1995-01-01

    The Lawrence Livermore National Laboratory (LLNL) Plutonium Facility conducts numerous programmatic activities involving plutonium to include device fabrication, development of improved and/or unique fabrication techniques, metallurgy research, and laser isotope separation. A Safety Analysis Report (SAR) for the building 332 Plutonium Facility was completed in July 1994 to address operational safety and acceptable risk to employees, the public, government property, and the environmental. This paper outlines the PRA analysis of the Evaluation Basis Fire (EBF) operational accident. The EBF postulates the worst-case programmatic impact event for the Plutonium Facility

  12. Methods for air cleaning system design and accident analysis

    International Nuclear Information System (INIS)

    Gregory, W.S.; Nichols, B.D.

    1987-01-01

    This paper describes methods, in the form of a handbook and five computer codes, that can be used for nuclear facility air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ventilation systems in the mining industry but do not seem to be commonly used in the nuclear industry. For example, the Nuclear Air Cleaning Handbook is an excellent design reference, but it fails to include information on computer codes that can be used to aid in the design process. These computer codes allow the analyst to use the handbook information to form all the elements of a complete system design. Because these analysis methods are in the form of computer codes they allow the analyst to investigate many alternative designs. In addition, the effects of many accident scenarios on the operation of the air cleaning system can be evaluated. These tools originally were intended for accident analysis, but they have been used mostly as design tools by several architect-engineering firms. The Cray, VAX, and personal computer versions of the codes, an accident analysis handbook, and the codes availability will be discussed. The application of these codes to several design operations of nuclear facilities will be illustrated, and their use to analyze the effect of several accident scenarios also will be described

  13. System requirements and design description for the document basis database interface (DocBasis)

    International Nuclear Information System (INIS)

    Lehman, W.J.

    1997-01-01

    This document describes system requirements and the design description for the Document Basis Database Interface (DocBasis). The DocBasis application is used to manage procedures used within the tank farms. The application maintains information in a small database to track the document basis for a procedure, as well as the current version/modification level and the basis for the procedure. The basis for each procedure is substantiated by Administrative, Technical, Procedural, and Regulatory requirements. The DocBasis user interface was developed by Science Applications International Corporation (SAIC)

  14. Addressing severe accidents in the CANDU 9 design

    International Nuclear Information System (INIS)

    Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

    1998-01-01

    CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

  15. Design of research reactors to take into account a reactivity accident

    International Nuclear Information System (INIS)

    Abou Yehia, H.; Berry, J.L.; Sinda, T.

    1990-01-01

    A description is given of the procedures followed and the studies performed in France with regard to the design of pool-type research reactors to cope with an explosive accident of the BORAX type. The examples of the high-flux reactor and of ORPHEE, the last reactor constructed, are developed at length. The development of the procedures and studies on the basis of results obtained by others is shown, and the conservative assumptions used when taking into account such an accident are described

  16. Design features of ACR in severe accident mitigation

    International Nuclear Information System (INIS)

    Shapiro, H.; Krishnan, V.S.; Santamaura, P.; Lekakh, B.; Blahnik, C.

    2007-01-01

    New reactor designs require the evaluation of design alternatives to reduce the radiological risk by preventing severe accidents or by limiting releases from the plant in the event of such accidents. The Advanced CANDU Reactor TM (ACR TM ) design has provisions to prevent and mitigate severe accidents. This paper describes key ACR design features for severe accident mitigation. It provides a high-level overview of the findings to date. Several design provisions have not yet been finalized or decided, but the designers are keenly aware of the SAM concepts and their requirements. The active heat sinks for 'vessels' (i.e., the fuel channels, the calandria vessel, the calandria end-shields and the calandria vault) are all amply capable of dissipating the severe accident heat loads. These heat sinks are designed to be operable under severe accident environmental conditions; however, their operability is yet to be confirmed by assessments. The active heat sinks for the various process vessels are 'backed up' by passive heat sinks (i.e., steaming plus water make-up from the RWS). The supply side of passive heat sinks is simple, rugged, and not vulnerable to failures of plant systems. The importance of the steam relief side is recognized, and the adequate relief capacity will be provided. The passive heat sinks will give the SAM more than 1 day (likely several days) to diagnose the accident and to establish the ultimate heat sinks. The spray system for containment pressure suppression is designed for high reliability and has ample capacity to ensure low containment leakage without external intervention, after which time alternative supply to the sprays can be brought on line manually. The sprays are backed up by the LACs which are assessed for operability following a severe accident. The strong ACR containment will provide a long time of completely passive protection for any severe accident at decay power. Its characteristics are not prone to catastrophic failures. The

  17. Design basis for the NRC Operations Center

    Energy Technology Data Exchange (ETDEWEB)

    Lindell, M.K.; Wise, J.A.; Griffin, B.N.; Desrosiers, A.E.; Meitzler, W.D.

    1983-05-01

    This report documents the development of a design for a new NRC Operations Center (NRCOC). The project was conducted in two phases: organizational analysis and facility design. In order to control the amount of traffic, congestion and noise within the facility, it is recommended that information flow in the new NRCOC be accomplished by means of an electronic Status Information Management System. Functional requirements and a conceptual design for this system are described. An idealized architectural design and a detailed design program are presented that provide the appropriate amount of space for operations, equipment and circulation within team areas. The overall layout provides controlled access to the facility and, through the use of a zoning concept, provides each team within the NRCOC the appropriate balance of ready access and privacy determined from the organizational analyses conducted during the initial phase of the project.

  18. Design basis for the NRC Operations Center

    International Nuclear Information System (INIS)

    Lindell, M.K.; Wise, J.A.; Griffin, B.N.; Desrosiers, A.E.; Meitzler, W.D.

    1983-05-01

    This report documents the development of a design for a new NRC Operations Center (NRCOC). The project was conducted in two phases: organizational analysis and facility design. In order to control the amount of traffic, congestion and noise within the facility, it is recommended that information flow in the new NRCOC be accomplished by means of an electronic Status Information Management System. Functional requirements and a conceptual design for this system are described. An idealized architectural design and a detailed design program are presented that provide the appropriate amount of space for operations, equipment and circulation within team areas. The overall layout provides controlled access to the facility and, through the use of a zoning concept, provides each team within the NRCOC the appropriate balance of ready access and privacy determined from the organizational analyses conducted during the initial phase of the project

  19. Review of current status for designing severe accident management support system

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwang Sub

    2000-05-01

    The development of operator support system (OSS) is ongoing in many other countries due to the complexity both in design and in operation for nuclear power plant. The computerized operator support system includes monitoring of some critical parameters, early detection of plant transient, monitoring of component status, plant maintenance, and safety parameter display, and the operator support system for these areas are developed and are being used in some plants. Up to now, the most operator support system covers the normal operation, abnormal operation, and emergency operation. Recently, however, the operator support system for severe accident is to be developed in some countries. The study for the phenomena of severe accident is not performed sufficiently, but, based on the result up to now, the operator support system even for severe accident will be developed in this study. To do this, at first, the current status of the operator support system for normal/abnormal/emergency operation is reviewed, and the positive aspects and negative aspects of systems are analyzed by their characteristics. And also, the major items that should be considered in designing the severe accident operator support system are derived from the review. With the survey of domestic and foreign operator support systems, they are reviewed in terms of the safety parameter display system, decision-making support system, and procedure-tracking system. For the severe accident, the severe accident management guideline (SAMG) which is developed by Westinghouse is reviewed; the characteristics, structure, and logical flow of SAMG are studied. In addition, the critical parameters for severe accident, which are the basis for operators decision-making in severe accident management and are supplied to the operators and the technical support center, are reviewed, too.

  20. Review of current status for designing severe accident management support system

    International Nuclear Information System (INIS)

    Jeong, Kwang Sub

    2000-05-01

    The development of operator support system (OSS) is ongoing in many other countries due to the complexity both in design and in operation for nuclear power plant. The computerized operator support system includes monitoring of some critical parameters, early detection of plant transient, monitoring of component status, plant maintenance, and safety parameter display, and the operator support system for these areas are developed and are being used in some plants. Up to now, the most operator support system covers the normal operation, abnormal operation, and emergency operation. Recently, however, the operator support system for severe accident is to be developed in some countries. The study for the phenomena of severe accident is not performed sufficiently, but, based on the result up to now, the operator support system even for severe accident will be developed in this study. To do this, at first, the current status of the operator support system for normal/abnormal/emergency operation is reviewed, and the positive aspects and negative aspects of systems are analyzed by their characteristics. And also, the major items that should be considered in designing the severe accident operator support system are derived from the review. With the survey of domestic and foreign operator support systems, they are reviewed in terms of the safety parameter display system, decision-making support system, and procedure-tracking system. For the severe accident, the severe accident management guideline (SAMG) which is developed by Westinghouse is reviewed; the characteristics, structure, and logical flow of SAMG are studied. In addition, the critical parameters for severe accident, which are the basis for operators decision-making in severe accident management and are supplied to the operators and the technical support center, are reviewed, too

  1. A risk-informed framework for establishing a beyond design basis safety basis for external hazards

    Energy Technology Data Exchange (ETDEWEB)

    Amico, P. [Hughes Associates, Inc, Baltimore, MD (United States); Anoba, R. [Hughes Associates, Inc, Raleigh, NC (United States); Najafi, B. [Hughes Associates, Inc., Los Gatos, CA (United States)

    2014-07-01

    The events at Fukushima Daiichi taught us that meeting a deterministic design basis requirement for external hazards does not assure that the risk is low. As observed at the plant, the two primary reasons for this are failure cliffs above the design basis event and that combined hazard effects are not considered in design. Because the possible combinations of design basis exceedences and external hazard combinations are very large and complex, an approach focusing only on the most important ones is needed. For this reason, a risk informed approach is the most effective approach, which is discussed in this paper. (author)

  2. Data base pertinent to earthquake design basis

    International Nuclear Information System (INIS)

    Sharma, R.D.

    1988-01-01

    Mitigation of earthquake risk from impending strong earthquakes is possible provided the hazard can be assessed, and translated into appropriate design inputs. This requires defining the seismic risk problem, isolating the risk factors and quantifying risk in terms of physical parameters, which are suitable for application in design. Like all other geological phenomena, past earthquakes hold the key to the understanding of future ones. Quantificatio n of seismic risk at a site calls for investigating the earthquake aspects of the site region and building a data base. The scope of such investigations is il lustrated in Figure 1 and 2. A more detailed definition of the earthquake problem in engineering design is given elsewhere (Sharma, 1987). The present document discusses the earthquake data base, which is required to support a seismic risk evaluation programme in the context of the existing state of the art. (author). 8 tables, 10 figs., 54 refs

  3. Integral Monitored Retrievable Storage (MRS) Facility conceptual basis for design

    International Nuclear Information System (INIS)

    1985-10-01

    The purpose of the Conceptual Basis for Design is to provide a control document that establishes the basis for executing the conceptual design of the Integral Monitored Retrievable Storage (MRS) Facility. This conceptual design shall provide the basis for preparation of a proposal to Congress by the Department of Energy (DOE) for construction of one or more MRS Facilities for storage of spent nuclear fuel, high-level radioactive waste, and transuranic (TRU) waste. 4 figs., 25 tabs

  4. Discussion about design basis flood of site of research reactors by river

    International Nuclear Information System (INIS)

    Rong Feng; Zhao Jianjun; Du Qiaomin; Zhang Lingyan

    2006-01-01

    This paper presents the well-defined standard in relation to design the basis flood of the sites of research reactors by river. It is based on the concept of some relational standards, analysis of hydrological calculation technology and methods, and analysis of accident dangerous degrees of research reactor, as well as in combination with the engineering practices. The flood preventing standard for research reactors with higher power should be the same with that of the nuclear power plants. (authors)

  5. Design features which mitigate severe accident challenges in the GE ABWR and SBWR

    International Nuclear Information System (INIS)

    Buchholz, Carol E.

    2004-01-01

    A reduction of the requirements for the emergency planning zone (EPZ) is a goal of advanced light water reactors. The technical basis for reducing the EPZ requirements is based on a very low frequency of a severe accident and high confidence that the offsite dose would be low even if a severe accident was to occur. Design features have been included in both the ABWR and SBWR to ensure that both of these goals are achieved. Probabilistic Risk Assessments (PRAs) have been performed for both plants. The PRAs indicate a core damage frequency on the order of IE-7 for both plants. The PRAs also show that the containments will not fail even if a severe accident should occur. The potential offsite is extremely low. (author)

  6. Accident sequence analysis of human-computer interface design

    International Nuclear Information System (INIS)

    Fan, C.-F.; Chen, W.-H.

    2000-01-01

    It is important to predict potential accident sequences of human-computer interaction in a safety-critical computing system so that vulnerable points can be disclosed and removed. We address this issue by proposing a Multi-Context human-computer interaction Model along with its analysis techniques, an Augmented Fault Tree Analysis, and a Concurrent Event Tree Analysis. The proposed augmented fault tree can identify the potential weak points in software design that may induce unintended software functions or erroneous human procedures. The concurrent event tree can enumerate possible accident sequences due to these weak points

  7. GIS tools for analyzing accidents and road design: A review

    Energy Technology Data Exchange (ETDEWEB)

    Satria, R.

    2016-07-01

    A significant unexpected outcome of transportation systems is road accidents with injuries and loss of lives. In recent years, the number of studies about the tools for analyzing accidents and road design has increased considerably. Among these tools, Geographical Information Systems (GIS) stand out for their ability to perform complex spatial analyses. However, sometimes the GIS, has been used only as a geographical database to store and represent data about accidents and road characteristics. It has also been used to represent the results of statistical studies of accidents but, these statistical studies have not been carried out with GIS. Owing to its integrated statistical-analysis capabilities GIS provides several advantages. First, it allows a more careful and accurate data selection, screening and reduction. Also, it allows a spatial analysis of the results in pre and post-processing. Second, GIS allows the development of spatial statistics that rely on geographically-referenced data. In this paper, several GIS tools used to model accidents have been examined. The understanding of these tools will help the analyst to make a better decision about which tool could be applied in each particular condition and context. (Author)

  8. Rupture of DN 500 - design basic accident at units 3 and 4 of Kozloduy NPP

    International Nuclear Information System (INIS)

    Uruchev, V.; Vassilev, P.; Ivanova, A.; Sartmadjiev, A.

    2005-01-01

    The original design of Kozloduy NPP Units 3 and 4 assumes as Design Basis Accident (DBA) the rupture of DN 32 mm primary pipeline, while an initial event of double-sided guillotine break of primary pipeline with maximal diameter is not considered. In the course of units modernization it have been demonstrated once and again that both the emergency core cooling systems and the localization systems can cope with larger and larger primary circuit leaks. After the installation of a Jet-Vortex Condenser (JVC) at Units 3 and 4 it was substantiated that, the integrity of the hermetic rooms is ensured even in case of double-sided guillotine break of a primary circuit pipeline with maximal diameter (DEGB). The technical justification of the jet-vortex condenser, elaborated by VNIAEC, contains calculations determining both the source term and the doses obtained outside the NPP site after LOCA DN 500. LOCA DN 500 is considered in these analyses as a beyond design basis accident and it is so included in the SAR and approved by the Nuclear Regulatory Agency (NRA). The thermo-hydraulic calculations performed later on show that the emergency core cooling systems can cope with this initial event at conservative assumptions. In order to classify this initiating event as a design basis accident it is necessary to demonstrate that the core cooling criteria are fulfilled and the internal and external doses outside the NPP site are within the permissible limits fixed for design basis accident by the Bulgarian regulatory body (NRA), when using conservative assumptions. For this purpose two consecutive studies were performed - evaluation of the DEGB probability and categorization of the initial event according to the contemporary regulations acting in Republic of Bulgaria. The presented report summarizes the results of the performed conservative analyses of double-sided guillotine break accident of main circulation line taking into account the probability of rupture of large diameter

  9. The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M. G.; Mouser, M. R.; Simon, J. B. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been

  10. The power of simplification: Operator interface with the AP1000R during design-basis and beyond design-basis events

    International Nuclear Information System (INIS)

    Williams, M. G.; Mouser, M. R.; Simon, J. B.

    2012-01-01

    The AP1000 R plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed

  11. Design feasibility study on corium stabilization in bottom end-fitting for AHWR under accident condition

    International Nuclear Information System (INIS)

    Gokhale, Onkar; Mukhopadhyay, D.; Chatterjee, B.; Singh, R.K.

    2015-01-01

    Advanced Heavy Water Reactor (AHWR) is being designed in a robust way to cater both Design and Beyond Design Basis Accidents to meet all the safety functions. All the functions are met by passive means with special emphasis on 'residual heat removal' which is catered by passive natural circulation mode. In context to Design Basis Accidents, several features are designed to handle worst kind of scenario like Station Black Out. For Design Extension Conditions (DEC), the means of passive natural circulation is adopted as a design means to meet the DEC-A conditions like cooling of moderator by natural circulation means with GDWP inventory. Under the DEC-B condition where large scale of fuel melting is envisaged, a core catcher is designed with active/passive cooling modes to take care of the residual heat of the core. All the mentioned features utilizes the natural mode of heat transfer to meet one of the safety function i.e. 'residual heat removal'. The analysis shows that the tube sheet as well as lattice tube temperatures remain low and are able to take out the heat from corium through sub-cooled nucleate boiling. The ES cooling is sufficient to maintain the cooling water in subcooled condition. The integrity of tube sheet and lattice tube is maintained

  12. Model review and evaluation for application in DOE safety basis documentation of chemical accidents - modeling guidance for atmospheric dispersion and consequence assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Woodarad, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanna, S. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hesse, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Huang, J. -C. [Argonne National Lab. (ANL), Argonne, IL (United States); Lewis, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mazzola, C. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    1997-09-01

    The U.S. Department of Energy (DOE), through its Defense Programs (DP), Office of Engineering and Operations Suppon, established the Accident Phenomenology and Consequence (AP AC) Methodology Evaluation Program to identify and evaluate methodologies and computer codes to support accident phenomenological and consequence calculations for both radiological and nonradiological materials at DOE facilities and to identify development needs. The program is also intended to define and recommend "best or good engineering/safety analysis practices" to be followed in preparing ''design or beyond design basis" assessments to be included in DOE nuclear and nonnuclear facility safety documents. The AP AC effort is intended to provide scientifically sound and more consistent analytical approaches, by identifying model selection procedures and application methodologies, in order to enhance safety analysis activities throughout the DOE complex.

  13. Noble gas control room accident filtration system for severe accident conditions N-CRAFT. System design

    International Nuclear Information System (INIS)

    Hill, Axel

    2014-01-01

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP. This can either be due to leakages of the containment or due to a filtered containment venting in order to ensure the overall integrity of the containment. During the containment venting process aerosols and iodine can be retained by the FCVS which prevents long term ground contamination. Noble gases are not retainable by the FCVS. From this it follows that a large amount of radioactive noble gases (e.g. xenon, krypton) might be present in the nearby environment of the plant dominating the activity release, depending on the venting procedure and the weather conditions. Accident management measures are necessary in case of severe accidents and the prolonged stay of staff inside the main control room (MCR) or emergency response center (ERC) is essential. Therefore, the in leakage and contamination of the MRC and ERC with airborne activity has to be prevented. The radiation exposure of the crises team needs to be minimized. The entrance of noble gases cannot be sufficiently prevented by the conventional air filtration systems such as HEPA filters and iodine absorbers. With the objective to prevent an unacceptable contamination of the MCR/ERC atmosphere by noble gases AREVA GmbH has developed a noble gas retention system. The noble gas control room accident filtration system CRAFT is designed for this case and provides supply of fresh air to the MCR/ERC without time limitation. The retention process of the system is based on the dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. These cycles ensure a periodic load and flushing of the delay lines retaining the noble gases from entering the MCR. CRAFT allows a minimization of the dose rate inside MCR/ERC and ensures a low radiation exposure to the staff on shift maintaining

  14. Maximum credible accident analysis for TR-2 reactor conceptual design

    International Nuclear Information System (INIS)

    Manopulo, E.

    1981-01-01

    A new reactor, TR-2, of 5 MW, designed in cooperation with CEN/GRENOBLE is under construction in the open pool of TR-1 reactor of 1 MW set up by AMF atomics at the Cekmece Nuclear Research and Training Center. In this report the fission product inventory and doses released after the maximum credible accident have been studied. The diffusion of the gaseous fission products to the environment and the potential radiation risks to the population have been evaluated

  15. Upwind design basis (WP4 : Offshore foundations and support structures)

    NARCIS (Netherlands)

    Fischer, T.; De Vries, W.E.; Schmidt, B.

    2010-01-01

    The presented design basis gives a summarized overview of relevant design properties for a later offshore wind turbine design procedures within work package 4. The described offshore site is located in the Dutch North Sea and has a water depth of 21m. Therefore it will be chosen as shallow site

  16. Fuel behaviour in the case of severe accidents and potential ATF designs. Fuel Behavior in Severe Accidents and Potential Accident Tolerance Fuel Designs

    International Nuclear Information System (INIS)

    Cheng, Bo

    2013-01-01

    This presentation reviews the conditions of fuel rods under severe loss of coolant conditions, approaches that may increase coping time for plant operators to recover, requirements of advanced fuel cladding to increase tolerance in accident conditions, potential candidate alloys for accident-tolerant fuel cladding and a novel design of molybdenum (Mo) -based fuel cladding. The current Zr-alloy fuel cladding will lose all its mechanical strength at 750-800 deg. C, and will react rapidly with high-pressure steam, producing significant hydrogen and exothermic heat at 700-1000 deg. C. The metallurgical properties of Zr make it unlikely that modifications of the Zr-alloy will improve the behaviour of Zr-alloys at temperatures relevant to severe accidents. The Mo-based fuel cladding is designed to (1) maintain fuel rod integrity, and reduce the release rate of hydrogen and exothermic heat in accident conditions at 1200-1500 deg. C. The EPRI research has thus far completed the design concepts, demonstration of feasibility of producing very thin wall (0.2 mm) Mo tubes. The feasibility of depositing a protective coating using various techniques has also been demonstrated. Demonstration of forming composite Mo-based cladding via mechanical reduction has been planned

  17. Simulations of the design basis accident at conditions of power increase and the o transient of MSIV at overpressure conditions of the Laguna Verde Power Station; Simulaciones del accidente base de diseno a condiciones de aumento de potencia y del transitorio de cierre de MSIV a condiciones de sobrepresion de la Central Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Araiza M, E.; Nunez C, A. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    2001-07-01

    This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is concluded that the integrity of the containment is not challenged during a loss of coolant accident at uprate power conditions.The analysis of the main steam isolation valve transients at overpressure conditions, and the analysis of the particular cases of the failure of one to six safety relief valves to open, show that the vessel peak pressures are below the design pressure and have no significant effect on vessel integrity. (Author)

  18. Protection of the Population in the event of a Nuclear accident. A Basis for Intervention

    International Nuclear Information System (INIS)

    1990-01-01

    During the years following the Chernobyl accident in 1986, the NEA actively participated in the international effort towards the improvement and better harmonization of the international and national criteria for the protection of the public in the event of a nuclear accident. A first report on this matter, titled Nuclear Accidents: Intervention Levels for Protection of the Public was published by the NEA in 1989. Subsequently, the NEA Committee on Radiation Protection and Public Health set up a small Task Group to provide additional guidance, and to take into account recent developments in other international organizations. The report outlines the status of relevant international activities in the period following the preparation of the 1989 report, discusses the intervention principles and describes both the proposed accident management system and a general scheme for its application. It is to be noted that the principles and criteria for intervention discussed in this report, although developed with specific reference to reactor accidents, apply equally well to activities and possible accidents at other nuclear facilities. The report briefly describes the transition from an accident management situation back to a normal situation and the related problem of changing criteria for the protection of the public. In addition to the traditional exposure pathways -inhalation from the cloud, external irradiation from the cloud and the ground and ingestion of food - the report acknowledges the existence of special pathways, proposing criteria for protecting workers and the public and some examples of their application

  19. Technical basis for the ITER-FEAT outline design

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-11-01

    This ITER EDA Documentation Series issue summarizes the results of the ITER Engineering Design Activities on the technical basis for the ITER-FEAT outline design. This issue also comprises some physical analysis activities as well as structure and goals of the Physics Expert Group activities.

  20. Technical basis for the ITER-FEAT outline design

    International Nuclear Information System (INIS)

    2000-01-01

    This ITER EDA Documentation Series issue summarizes the results of the ITER Engineering Design Activities on the technical basis for the ITER-FEAT outline design. This issue also comprises some physical analysis activities as well as structure and goals of the Physics Expert Group activities

  1. Toxic industrial chemicals (TICs) as asymmetric weapons: the design basis threat

    International Nuclear Information System (INIS)

    Skinner, L.

    2009-01-01

    Asymmetric warfare concepts relate well to the use of improvised chemical weapons against urban targets. Sources of information on toxic industrial chemicals (TICs) and lists of high threat chemicals are available that point to likely choices for an attack. Accident investigations can be used as a template for attacks, and to judge the possible effectiveness of an attack using TICs. The results of a chlorine rail car accident in South Carolina, USA and the Russian military assault on a Moscow theater provide many illustrative points for similar incidents that mighty be carried out deliberately. Computer modeling of outdoor releases shows how an attack might take into consideration issues of stand-off distance and dilution. Finally, the preceding may be used to estimate with some accuracy the design basis threat posed by the used of TICs as weapons.(author)

  2. Solar Power Tower Design Basis Document, Revision 0

    Energy Technology Data Exchange (ETDEWEB)

    ZAVOICO,ALEXIS B.

    2001-07-01

    This report contains the design basis for a generic molten-salt solar power tower. A solar power tower uses a field of tracking mirrors (heliostats) that redirect sunlight on to a centrally located receiver mounted on top a tower, which absorbs the concentrated sunlight. Molten nitrate salt, pumped from a tank at ground level, absorbs the sunlight, heating it up to 565 C. The heated salt flows back to ground level into another tank where it is stored, then pumped through a steam generator to produce steam and make electricity. This report establishes a set of criteria upon which the next generation of solar power towers will be designed. The report contains detailed criteria for each of the major systems: Collector System, Receiver System, Thermal Storage System, Steam Generator System, Master Control System, and Electric Heat Tracing System. The Electric Power Generation System and Balance of Plant discussions are limited to interface requirements. This design basis builds on the extensive experience gained from the Solar Two project and includes potential design innovations that will improve reliability and lower technical risk. This design basis document is a living document and contains several areas that require trade-studies and design analysis to fully complete the design basis. Project- and site-specific conditions and requirements will also resolve open To Be Determined issues.

  3. Simulant Basis for the Standard High Solids Vessel Design

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-09-30

    The Waste Treatment and Immobilization Plant (WTP) is working to develop a Standard High Solids Vessel Design (SHSVD) process vessel. To support testing of this new design, WTP engineering staff requested that a Newtonian simulant and a non-Newtonian simulant be developed that would represent the Most Adverse Design Conditions (in development) with respect to mixing performance as specified by WTP. The majority of the simulant requirements are specified in 24590-PTF-RPT-PE-16-001, Rev. 0. The first step in this process is to develop the basis for these simulants. This document describes the basis for the properties of these two simulant types. The simulant recipes that meet this basis will be provided in a subsequent document.

  4. Information management needs for Fort Calhoun's design basis reconstitution project

    International Nuclear Information System (INIS)

    Beach, D.R.; Erickson, E.A.; Gambhir, S.K.; Parsons, R.D.

    1989-01-01

    While the need for information management is not new to the nuclear industry or Omaha Public Power District (OPPD), the interrelationship among design information, multiple systems, and design basis issues has necessitated the management of this information in new ways. The project team involved in the reconstitution of the design basis for OPPD's Fort Calhoun nuclear station has experienced the need for the developed effective methods for managing the vast amount of interrelated information associated with this effort. This management of information has been necessary to ensure that design basis documents (DBDs) adequately reflect the interrelated nature of component, system, and plant design; are complete and accurate; and are produced and maintained in a cost-effective manner. Fort Calhoun's aggressive design basis reconstitution project began in early 1987. The present scope of the project includes the production of 52 system and plant level DBDs; currently the project is ∼50% complete with DBDs in various stages of completion, from pilot DBDs through DBDs with approved formats, which have been issued for use. The experience in producing these documents has lead to a growing understanding of the special need for information management in each stage of the project. The development of the information tracking and management processes for the various stages of DBD development has proven to be cost-effective and gives a level of assurance that information has been included in the DBDs consistently and accurately

  5. Post accident training program design at Three Mile Island

    International Nuclear Information System (INIS)

    Lawyer, L.L.

    1981-01-01

    The TMI preaccident training staff typically consisted of 9 professional and 3 administrative support persons. Procedures were prepared and facilities designated for operator training. The thrust of the post accident effort was directed to expanding the training function to include all other personnel while modifying the operator training to address lessons learned. Significant experiences were encountered in part task simulation, job and task analysis, decision analysis and with various external committees. These experiences led to specific opinions on industry needs in the areas of staffing, regulation, importance of training and contractor assistance

  6. Design basis document open-item resolution and reportability

    International Nuclear Information System (INIS)

    Gambhir, S.K.; Livingston, B.R.; Purcell, J.J.; Erickson, E.A.

    1989-01-01

    In the process of reconstituting the design bases for older nuclear power plants, information or references may not be available to fully define the design requirements or to document and verify the adequacy of the design. Also, information that is in conflict with other data is identified. The missing and conflicting information must be reconstituted in order to adequately document the design bases of the plant. For these operating facilities, the identification, tracking, and resolution of missing or conflicting information is very important when the reporting requirements stipulated by 10CFR21, 10CFR50.72, and 10CFR50.73 are considered. Additionally, controlled documentation (calculations, drawings, etc.) used to develop the design basis documents may contain conflicting data. In some cases, conflicts between the as-built design and licensing or design basis requirements established in specific commitments to the U.S. Nuclear Regulatory Commission may be identified. Furthermore, concerns regarding the adequacy of safety-related systems or components to perform their required function may be identified that would warrant prompt action by the licensee. The approach discussed in this paper was used by Omaha Public Power District for the ongoing design basis reconstitution effort at the Fort Calhoun nuclear plant

  7. Design basis programs and improvements in plant operation

    International Nuclear Information System (INIS)

    Metcalf, M.F.

    1991-01-01

    Public Service Electric and Gas (PSE and G) Company operates three commercial nuclear power plants in southern New Jersey. The three plants are of different designs and vintages (two pressurized water reactors licensed in 1976 and 1980 and one boiling water reactor licensed in 1986). As the industry recognized the need to develop design basis programs, PSE and G also realized the need after a voluntary 52-day shutdown of one unit because of electrical design basis problems. In its drive to be a premier electric utility, PSE and G has been aggressively active in developing design basis documents (DBDs) with supporting projects and refined uses to obtain the expected value and see the return on investment. Progress on Salem is nearly 75% complete, while Hope Creek is 20% complete. To data, PSE and G has experienced success in the use of DBDs in areas such as development of plant modifications, development of the reliability-centered maintenance program, procedure upgrades, improved document retrieval, resolution of regulatory issues, and training. The paper examines the design basis development process, supporting projects, and expected improvements in plant operations as a result of these efforts

  8. Postulated accidents

    International Nuclear Information System (INIS)

    Ullrich, W.

    1980-01-01

    This lecture on 'Postulated Accidents' is the first of a series of lectures on the dynamic and transient behaviour of nuclear power plants, especially pressurized water reactors. The main points covered will be: Reactivity Accidents, Transients (Intact Loop) and Loss of Cooland Accidents (LOCA) including small leak. This lecture will discuss the accident analysis in general, the definition of the various operational phases, the accident classification, and, as an example, an accident sequence analysis on the basis of 'Postulated Accidents'. (orig./RW)

  9. Design basis ground motion (Ss) required on new regulatory guide

    International Nuclear Information System (INIS)

    Kamae, Katsuhiro

    2013-01-01

    New regulatory guide is enforced on July 8. Here, it is introduced how the design basis ground motion (Ss) for seismic design of nuclear power reactor facilities was revised on the new guide. Ss is formulated as two types of earthquake ground motions, earthquake ground motions with site specific earthquake source and with no such specific source locations. The latter is going to be revised based on the recent observed near source ground motions. (author)

  10. An information-theoretic basis for uncertainty analysis: application to the QUASAR severe accident study

    International Nuclear Information System (INIS)

    Unwin, S.D.; Cazzoli, E.G.; Davis, R.E.; Khatib-Rahbar, M.; Lee, M.; Nourbakhsh, H.; Park, C.K.; Schmidt, E.

    1989-01-01

    The probabilistic characterization of uncertainty can be problematic in circumstances where there is a paucity of supporting data and limited experience on which to base engineering judgement. Information theory provides a framework in which to address this issue through reliance upon entropy-related principles of uncertainty maximization. We describe an application of such principles in the United States Nuclear Regulatory Commission-sponsored program QUASAR (Quantification and Uncertainty Analysis of Source Terms for Severe Accidents in Light Water Reactors). (author)

  11. Loss-of-coolant accident analysis of the Savannah River new production reactor design

    International Nuclear Information System (INIS)

    Maloney, K.J.; Pryor, R.J.

    1990-11-01

    This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

  12. Basis for NGNP Reactor Design Down-Selection

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2010-08-01

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  13. Design-Load Basis for LANL Structures, Systems, and Components

    Energy Technology Data Exchange (ETDEWEB)

    I. Cuesta

    2004-09-01

    This document supports the recommendations in the Los Alamos National Laboratory (LANL) Engineering Standard Manual (ESM), Chapter 5--Structural providing the basis for the loads, analysis procedures, and codes to be used in the ESM. It also provides the justification for eliminating the loads to be considered in design, and evidence that the design basis loads are appropriate and consistent with the graded approach required by the Department of Energy (DOE) Code of Federal Regulation Nuclear Safety Management, 10, Part 830. This document focuses on (1) the primary and secondary natural phenomena hazards listed in DOE-G-420.1-2, Appendix C, (2) additional loads not related to natural phenomena hazards, and (3) the design loads on structures during construction.

  14. CE/Bechtel design containment response to severe accident phenomenology: A comparison among several combustion engineering plants

    International Nuclear Information System (INIS)

    Khalil, Y.F.; Schneider, R.E.

    1995-01-01

    The objectives of this paper are to: (1) discuss the types of severe accident phenomena that drive containment failure modes in CE plants and (2) contribute to the current state of knowledge of CE/Bechtel-design containment response to severe accident phenomenology. The second objective is addressed by providing a comparative study of containment response to severe accidents among several CE plants including Millstone Unit 2 (MP2), Palisades (Consumers Power), Calvert Cliffs (Baltimore Gas and Electric Company), Palo Verde (Arizona Public Service), and SONGS Units 2 and 3 (Southern California Edison). The motivation for addressing the second objective is based on the current lack of comprehensive literature on CE/Bechtel design containment failure modes and mechanisms for accidents that progress beyond the design basis limits. The first part of this paper addresses severe accident phenomena-related failure mechanisms in CE/Bechtel-designed containments. The second part of this work provides a comparative study of containment response among several CE plants

  15. The effect of roundabout design features on cyclist accident rate

    DEFF Research Database (Denmark)

    Hels, Tove; Orozova-Bekkevold, Ivanka

    2007-01-01

    Roundabouts are known to result in fewer traffic accidents than traditional intersections. However, this is to a lesser degree true for bicycles than for vehicles. In this paper, we aimed at establishing statistical relationships through Poisson regression and logistic regression analyses between...... was age of the roundabout-older roundabouts related to more accidents and higher accident probability. Excluding 48 single cyclist accidents strengthened the relationship between accidents on one hand and vehicle and cyclist volume and potential vehicle speed on the other. This stresses the significance...

  16. Advances in the physics basis for the European DEMO design

    Science.gov (United States)

    Wenninger, R.; Arbeiter, F.; Aubert, J.; Aho-Mantila, L.; Albanese, R.; Ambrosino, R.; Angioni, C.; Artaud, J.-F.; Bernert, M.; Fable, E.; Fasoli, A.; Federici, G.; Garcia, J.; Giruzzi, G.; Jenko, F.; Maget, P.; Mattei, M.; Maviglia, F.; Poli, E.; Ramogida, G.; Reux, C.; Schneider, M.; Sieglin, B.; Villone, F.; Wischmeier, M.; Zohm, H.

    2015-06-01

    In the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

  17. Design basis tropical cyclone for nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

    The general characteristics of tropical cyclones are discussed in this Safety Guide, with particular emphasis on their pressure and wind structures in the light of available data. General methods are given for the evaluation of the relevant parameters of a Probable Maximum Tropical Cyclone (PMTC), which can be used as the Design Basis Tropical Cyclone (DBTC); these parameters then serve as inputs for the derivation of a design basis surge and a design basis wind. A possible method is also given for the evaluation of the PMTC pressure and wind field based on an approach valid primarily for a particular region. This method depends on the results of a theoretical study on the tropical cyclone structure and makes use of a large amount of data, including aircraft reconnaissance observations for 170 most intense tropical cyclones near the coast of Japan, Taiwan and the Philippines for the period 1960-1974, as well as detailed analyses of all the extreme storms along the Gulf of Mexico and the east coast of the USA during 1900-1978, for the determination of the necessary parameters

  18. Design basis event consequence analyses for the Yucca Mountain project

    International Nuclear Information System (INIS)

    Orvis, D.D.; Haas, M.N.; Martin, J.H.

    1997-01-01

    Design basis event (DBE) definition and analysis is an ongoing and integrated activity among the design and analysis groups of the Yucca Mountain Project (YMP). DBE's are those that potentially lead to breach of the waste package and waste form (e.g., spent fuel rods) with consequent release of radionuclides to the environment. A Preliminary Hazards Analysis (PHA) provided a systematic screening of external and internal events that were candidate DBE's that will be subjected to analyses for radiological consequences. As preparation, pilot consequence analyses for the repository subsurface and surface facilities have been performed to define the methodology, data requirements, and applicable regulatory limits

  19. Air temperature determination inside residual heat removal pump room of Angra-1 nuclear power plant after a design basic accident

    International Nuclear Information System (INIS)

    Siniscalchi, Marcio Rezende

    2005-01-01

    This work develops heat transfer theoretical models for determination of air temperature inside the Residual Heat Removal Pump Room of Angra 1 Nuclear Power Plant after a Design Basis Accident without forced ventilation. Two models had been developed. The differential equations are solved by analytical methods. A software in FORTRAN language are developed for simulations of temperature inside rooms for different geometries and materials. (author)

  20. Overview of core disruptive accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.

    1977-01-01

    An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods are used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described

  1. Conceptual design considerations for providing hook-up type schemes for tracking beyond design basis events (BDBE) for 700 MWe PHWR project

    International Nuclear Information System (INIS)

    Vhora, S.F.; Inder Jit; Bhardwaj, S.A.

    2005-01-01

    A broad review of major nuclear accidents such as Chernobyl reveals that provision of access to the reactor core for cooling purpose had to be made from outside the reactor building by tunneling. Also the NAPS fire incident could be mitigated once the fire water injection to the steam generators could be ensured. In this case the boiler room which was outside the primary containment was accessible relatively easily for mitigation after the initial period. Both of the above had accident scenarios which can be termed Beyond Design Basis (BDBE) since the accident initiation/scenario did not fit into the events under postulated initiating events (PIES) or Design Basis Events (DBEs). These accidents or events reveal that some sort of access to the core or the components inside the Reactor building becomes necessary. It is also to be noted that manual intervention beyond the initial period of half an hour or earlier in the Emergency operating procedure (EOP) is inevitably called for as a recovery action in order to mitigate the severity and minimize long term consequences. This paper attempts to discuss the type of concepts which can give access to the core or associated systems which can then provide continued heat sink. The discussions would include the criteria for design of such concepts and give examples of such concepts already implemented and proposes schemes to be implemented in the 700 MWe Project. (author)

  2. Application of the MELCOR code to design basis PWR large dry containment analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  3. Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study

    International Nuclear Information System (INIS)

    CLEVELAND, K.J.

    2000-01-01

    This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage

  4. Retrofitting a spent fuel pool spray system for alternative cooling as a strategy for beyond design basis events

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph; Vujic, Zoran [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2017-06-15

    Due to requirements for nuclear power plants to withstand beyond design basis accidents, including events such as happened in 2011 in the Fukushima Daiichi Nuclear Power Plant in Japan, alternative cooling of spent fuel is needed. Alternative spent fuel cooling can be provided by a retrofitted spent fuel pool spray system based on the AP1000 plant design. As part of Krsko Nuclear Power Plant's Safety Upgrade Program, Krsko Nuclear Power Plant decided on, and Westinghouse successfully designed a retrofit of the AP1000 {sup registered} plant spent fuel pool spray system to provide alternative spent fuel cooling.

  5. Possible emission of radioactive fission products during off-design accidents at a nuclear power plant with VVER-1000 reactor

    International Nuclear Information System (INIS)

    Dubkov, A.P.; Kozlov, V.F.; Luzanova, L.M.

    1995-01-01

    It is well known that eight nuclear power plants with VVER-1000 reactors have been constructed in Russia, Ukraine, and in the Republic of Belarus and they have been operating successfully without any serious accidents since 1980. These facilities have been analyzed for various accident scenarios, and measures have been incorporated which will prevent core damage during these possible events. However, an off-design accident can occur, and in such a case, the radiological consequences would exceed the worst design accidents. This paper reviews a number of potential off-design accidents in order to develop an accident plan to mitigate the consequences of such an accident

  6. Archaeological data as a basis for repository marker design

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1982-10-01

    This report concerns the development of a marking system for a nuclear waste repository which is very likely to survive for 10,000 years. In order to provide a background on the subject, and for the preliminary design presented in this report, a discussion is presented about the issues involved in human interference with the repository system and the communication of information. A separate chapter summarizes six ancient man-made monuments including: materials, effects of associated textual information on our understanding of the monument, and other features of the ancient monument relevant to marking a repository site. The information presented in the two chapters is used to provide the basis and rationale for a preliminary marker system design presented in a final chapter. 86 refs., 22 figs., 1 tab

  7. Reactor safety under design basis flood condition for inland sites

    International Nuclear Information System (INIS)

    Hajela, S.; Bajaj, S.S.; Samota, A.; Verma, U.S.P.; Warudkar, A.S.

    2002-01-01

    Full text: In June 1994, there was an incident of flooding at Kakrapar Atomic Power Station (KAPS) due to combination of heavy rains and mechanical failure in the operation of gates at the adjoining weir. An indepth review of the incident was carried out and a number of flood protection measures were recommended and were implemented at site. As part of this review, a safety analysis was also done to demonstrate reactor safety with a series of failures considered in the flood protection features. For each inland NPP site, as part of design, different flood scenarios are analysed to arrive at design basis flood (DBF) level. This level is estimated based on worst combination of heavy local precipitation, flooding in river, failure of upstream/downstream water control structures

  8. Archaeological data as a basis for repository marker design

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, M.F.

    1982-10-01

    This report concerns the development of a marking system for a nuclear waste repository which is very likely to survive for 10,000 years. In order to provide a background on the subject, and for the preliminary design presented in this report, a discussion is presented about the issues involved in human interference with the repository system and the communication of information. A separate chapter summarizes six ancient man-made monuments including: materials, effects of associated textual information on our understanding of the monument, and other features of the ancient monument relevant to marking a repository site. The information presented in the two chapters is used to provide the basis and rationale for a preliminary marker system design presented in a final chapter. 86 refs., 22 figs., 1 tab.

  9. Design basis for resistance to shock and vibration

    International Nuclear Information System (INIS)

    Glass, R.E.; Gwinn, K.W.

    1989-01-01

    Sandia National Laboratories, in conjunction with its participation in the American National Standards Institute (ANSI) writing groups, has undertaken to provide an experimental and analytical basis for the design of components of radioactive materials packages to resist normal transport shock and vibration loads. Previous efforts have resulted in an overly conservative shock spectra description of the loads in the tie-downs and cask attachment points anticipated during normal shipment. The present effort is aimed at predicting the actual loads so that the design basis can be accurately determined. This goal is being accomplished with road simulator and over-the-road tests and the development of an analytical model. This model is used to parametrically evaluate and envelop the transportation systems' responses. The parameters to be varied include damping, stiffness, geometry, and cargo mass. The over-the-road tests provide operational data that are used to validate the selection of environments for the road simulator tests. The road simulator tests provide verification for the model. This verification is accomplished since the road simulator tests provide not only the system response which can be measured in over-the-road tests but also the system input. Finally, when the model has been verified, it can be used to vary parameters to envelop a wide range of normal transport conditions

  10. Design basis for resistance to shock and vibration

    International Nuclear Information System (INIS)

    Glass, R.E.; Gwinn, K.W.

    1989-01-01

    Sandia National Laboratories, in conjunction with its participation in the American National Standards Institute (ANSI) writing groups, has undertaken to provide an experimental and analytical basis for the design of components of radioactive materials packages to resist normal transport shock and vibration loads. Previous efforts have resulted in an overly conservative shock spectra description of the loads in the tie-downs and cask attachment points anticipated during normal shipment. The present effort is aimed at predicting the actual loads so that the design basis can be accurately determined. This goal is being accomplished with road simulator and over-the-road tests and the development of an analytical model. This model is used to parametrically evaluate and envelop the transportation systems responses. The parameters to be varied include damping, stiffness, geometry, and cargo mass. The over-the-road tests provide operational data that are used to validate the selection of environments for the road simulator tests. The road simulator tests provide verification for the model. This verification is accomplished since the road simulator tests provide not only the system response which can be measured in over-the-road tests but also the system input. Finally, when the model has been verified, it can be used to vary parameters to envelope a wide range of normal transport conditions

  11. WIPP conceptual design report. Addendum G. Accident analysis for Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Shefelbine, H.C.; Metcalf, J.H.

    1977-06-01

    The types of accidents or risks pertinent to the Waste Isolation Pilot Plant (WIPP) are presented. Design features addressing these risks are discussed. Also discussed are design features that protect the public

  12. Probabilistic Accident Progression Analysis with application to a LMFBR design

    International Nuclear Information System (INIS)

    Jamali, K.M.

    1982-01-01

    A method for probabilistic analysis of accident sequences in nuclear power plant systems referred to as ''Probabilistic Accident Progression Analysis'' (PAPA) is described. Distinctive features of PAPA include: (1) definition and analysis of initiator-dependent accident sequences on the component level; (2) a new fault-tree simplification technique; (3) a new technique for assessment of the effect of uncertainties in the failure probabilities in the probabilistic ranking of accident sequences; (4) techniques for quantification of dependent failures of similar components, including an iterative technique for high-population components. The methodology is applied to the Shutdown Heat Removal System (SHRS) of the Clinch River Breeder Reactor Plant during its short-term (0 -2 . Major contributors to this probability are the initiators loss of main feedwater system, loss of offsite power, and normal shutdown

  13. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1997-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  14. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1998-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  15. Data Requirements and the Basis for Designing Health Information Kiosks.

    Science.gov (United States)

    Afzali, Mina; Ahmadi, Maryam; Mahmoudvand, Zahra

    2017-09-01

    Health kiosks are an innovative and cost-effective solution that organizations can easily implement to help educate people. To determine the data requirements and basis for designing health information kiosks as a new technology to maintain the health of society. By reviewing the literature, a list of information requirements was provided in 4 sections (demographic information, general information, diagnostic information and medical history), and questions related to the objectives, data elements, stakeholders, requirements, infrastructures and the applications of health information kiosks were provided. In order to determine the content validity of the designed set, the opinions of 2 physicians and 2 specialists in medical informatics were obtained. The test-retest method was used to measure its reliability. Data were analyzed using SPSS software. In the proposed model for Iran, 170 data elements in 6 sections were presented for experts' opinion, which ultimately, on 106 elements, a collective agreement was reached. To provide a model of health information kiosk, creating a standard data set is a critical point. According to a survey conducted on the various literature review studies related to the health information kiosk, the most important components of a health information kiosk include six categories; information needs, data elements, applications, stakeholders, requirements and infrastructure of health information kiosks that need to be considered when designing a health information kiosk.

  16. DART - for design basis justification and safety related information management

    International Nuclear Information System (INIS)

    Billington, A.; Blondiaux, P.; Boucau, J.; Cantineau, B.; Doumont, C.; Mared, A.

    2000-01-01

    DART is the acronym for Design Analysis Re-engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

  17. DART - for design basis justification and safety related information management

    International Nuclear Information System (INIS)

    Billington, A.; Blondiaux, B.; Boucau, J.; Cantineau, B.; Mared, A.

    2001-01-01

    DART is the acronym for Design Analysis Re-Engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can then be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

  18. Unites States position paper on sodium fires. Design basis and testing

    International Nuclear Information System (INIS)

    Lancet, R.T.; Johnson, R.P.; Matlin, E.; Vaughan, E.U.; Fields, D.E.; Glueckler, E.; McCormack, J.D.; Miller, C.W.; Pedersen, D.R.

    1989-01-01

    This paper focuses on designs, analyses, and tests performed since the last Sodium Fires Meeting of the IAEA International Working Group on Fast Reactors in May 1982. Since the U.S. Liquid Metal Reactor (LMR) program is focused on the two advanced LMRs, SAFR and PRISM, the paper relates this work to these designs. First, the design philosophy and approach taken by these advanced pool reactors are described. This includes methods of leak detection, the design basis leaks, and passive accommodation of sodium fires. Then the small- and large-scale sodium fire tests performed in support of the Clinch River Breeder Reactor Plant (CRBRP) program, including post-accident cleanup, are presented and related to the advanced LMR designs. Next, the assessment and behavior of the aerosols generated are discussed including generation rate, behavior within structures, release and dispersal, and deposition on safety-grade equipment. Finally, the impact of these aerosols on the performance of safety-grade decay heat removal heat exchange surfaces is discussed including some test results as well as planned tests. (author)

  19. Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Sevy, R.H.; Su, S.F.

    1985-01-01

    This paper presents the results of a study of the effectivness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). Results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs

  20. Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

    Science.gov (United States)

    Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

    2015-04-01

    Preparedness of nuclear power plants to beyond design base external effects became high importance after 11th of March 2011 Great Tohoku Earthquakes. In case of some nuclear power plants constructed at the soft soil sites, liquefaction should be considered as a beyond design basis hazard. The consequences of liquefaction have to be analysed with the aim of definition of post-event plant condition, identification of plant vulnerabilities and planning the necessary measures for accident management. In the paper, the methodology of the analysis of liquefaction effects for nuclear power plants is outlined. The case of Nuclear Power Plant at Paks, Hungary is used as an example for demonstration of practical importance of the presented results and considerations. Contrary to the design, conservatism of the methodology for the evaluation of beyond design basis liquefaction effects for an operating plant has to be limited to a reasonable level. Consequently, applicability of all existing methods has to be considered for the best estimation. The adequacy and conclusiveness of the results is mainly limited by the epistemic uncertainty of the methods used for liquefaction hazard definition and definition of engineering parameters characterizing the consequences of liquefaction. The methods have to comply with controversial requirements. They have to be consistent and widely accepted and used in the practice. They have to be based on the comprehensive database. They have to provide basis for the evaluation of dominating engineering parameters that control the post-liquefaction response of the plant structures. Experience of Kashiwazaki-Kariwa plant hit by Niigata-ken Chuetsu-oki earthquake of 16 July 2007 and analysis of site conditions and plant layout at Paks plant have shown that the differential settlement is found to be the dominating effect in case considered. They have to be based on the probabilistic seismic hazard assessment and allow the integration into logic

  1. Technical basis for the ITER-FEAT outline design. Progress in resolving open design issues from the outline design report

    International Nuclear Information System (INIS)

    2000-01-01

    In this publication the technical basis for the ITER-FEAT outline design is presented. It comprises the Plant Design Specifications, the Safety Principles and Environmental Criteria, the Site Requirements and Site Design Assumptions. The outline of the key features of the ITER-FEAT design includes main physical parameters and assessment, design overview and preliminary safety assessment, cost and schedule

  2. The concept of risk in the design basis threat

    International Nuclear Information System (INIS)

    Reynolds, J.M.

    2001-01-01

    Full text: Mathematically defined, risk is a product of one or more probability factors and one or more consequences. Actuarial analysis of risk requires the creation of a numeric algorithm that reflects the interaction of different probability factors, where probability data usually draws on direct measurements of incidence. For physical protection purposes, the algorithms take the general form: Risk = Probability of successful attack x Consequence where the overall probability of a successful attack will be determined by the product of, amongst other things, the probability of there being sufficient intent, the probability of there being available hostile resources, the probability of deterrence, and the probability that a hostile act will be detected and prevented. Deliberate, malevolent acts against nuclear facilities are rare. In so far as it is possible to make an actuarial type of judgement, the probability of malevolent activity against a nuclear facility is almost zero. This creates a problem for a numerical assessment of risk for nuclear facilities where the value (consequence) term could be almost infinite. As can be seen from the general equation above, a numerical algorithm of risk of malevolent activity affecting nuclear facilities could only yield a zero or infinite result. In such circumstances, intelligence-based threat assessments are sometimes thought of as a substitute for historic data in the determination of probability. However, if the paucity of historic data reflects the actual threat - which by and large it should - no amount of intelligence is likely to yield a substantially different conclusion. This mathematical approach to analysing risk appears to lead us either to no risk and no protection or to an infinite risk demanding every conceivable protective measure. The Design Basis Threat (DBT) approach offers a way out of the dilemma. Firstly, it allows us to eliminate from further consideration all zero or near zero probabilities

  3. Some Lessons Learnt From the Fukushima Daiichi Accident, as Regards Defence in Depth and its Implementation in New or Existing Designs – An Industry Example

    Energy Technology Data Exchange (ETDEWEB)

    De L’Epinois, B.; Bouteille, F.; Nicaise, N., E-mail: bertrand.delepinois@areva.com [AREVA, Paris (France)

    2014-10-15

    Defence-in-Depth (DiD) concept has been the overarching principle of the nuclear safety since the design of the first power reactors and it remains so more than ever. The Fukushima accident, characterised by a massive common mode failure induced by the flooding due to the tsunami, reminds the need for a very careful implementation, in the engineering, construction and operation of nuclear facilities, of the principles and rules concerning DiD. In particular, this accident highlights the need for appropriate consideration of DiD in two domains: protection against external hazards and severe accident management. In terms of external hazards, Fukushima reminds that the site design basis hazards must be as complete as possible and incorporate all relevant, including new, knowledge. In addition, in case nature would reveal to be more imaginative than us, or in case some hazards would have been under-evaluated despite all precautions, adequate extra margins should be ensured beyond the design basis hazards, as a defence in depth provision to avoid cliff edge effects. In other words, a minimal set of essential safety functions needed to prevent a severe accident or to mitigate its consequences should show sufficient robustness and safety margins to cope with external hazards exceeding the design basis hazards. Concerning severe accident, comprehensive R&D has been performed for several decades, the occurrence of such situations has been included in the DiD principles (level 4 as defined by INSAG 10) and has led to substantial safety improvements at many plants. Fukushima reminds the importance of a thorough implementation of these mitigation provisions. These two DiD provisions can be expressed as the need to show that the nuclear facilities and the crisis organizations can cope with extreme, beyond design hazards or accidents, precluding unacceptable offsite radiological impact and contamination. The development of Gen 3+ reactors targeted a significant safety step, by

  4. Design of a High Power Robotic Manipulator for Emergency Response to the Nuclear Accidents

    International Nuclear Information System (INIS)

    Park, Jongwon; Bae, Yeong-Geol; Kim, Myoung Ho; Choi, Young Soo

    2016-01-01

    An accident in a nuclear facility causes a great social cost. To prevent an unexpected nuclear accident from spreading to the catastrophic disaster, emergency response action in early stage is required. However, high radiation environment has been proved as a challenging obstacle for human workers to access to the accident site and take an action in previous accident cases. Therefore, emergency response robotic technology to be used in a nuclear accident site instead of human workers are actively conducted in domestically and internationally. Robots in an accident situation are required to carry out a variety of tasks depend on the types and patterns of accidents. An emergency response usually includes removing of debris, make an access road to a certain place and handling valves. These tasks normally involve high payload handling. A small sized high power robotic manipulator can be an appropriate candidate to deal with a wide spectrum of tasks in an emergency situation. In this paper, we discuss about the design of a high power robotic manipulator, which is capable of handling high payloads for an initial response action to the nuclear facility accident. In this paper, we presented a small sized high power robotic manipulator design. Actuator types of manipulator was selected and mechanical structure was discussed. In the future, the servo valve and hydraulic pump systems will be determined. Furthermore, control algorithms and test bed experiments will be also conducted

  5. Design of a High Power Robotic Manipulator for Emergency Response to the Nuclear Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jongwon; Bae, Yeong-Geol; Kim, Myoung Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    An accident in a nuclear facility causes a great social cost. To prevent an unexpected nuclear accident from spreading to the catastrophic disaster, emergency response action in early stage is required. However, high radiation environment has been proved as a challenging obstacle for human workers to access to the accident site and take an action in previous accident cases. Therefore, emergency response robotic technology to be used in a nuclear accident site instead of human workers are actively conducted in domestically and internationally. Robots in an accident situation are required to carry out a variety of tasks depend on the types and patterns of accidents. An emergency response usually includes removing of debris, make an access road to a certain place and handling valves. These tasks normally involve high payload handling. A small sized high power robotic manipulator can be an appropriate candidate to deal with a wide spectrum of tasks in an emergency situation. In this paper, we discuss about the design of a high power robotic manipulator, which is capable of handling high payloads for an initial response action to the nuclear facility accident. In this paper, we presented a small sized high power robotic manipulator design. Actuator types of manipulator was selected and mechanical structure was discussed. In the future, the servo valve and hydraulic pump systems will be determined. Furthermore, control algorithms and test bed experiments will be also conducted.

  6. Innovations in systems engineering and analysis for the simulation of beyond design-base accidents

    International Nuclear Information System (INIS)

    Frisch, W.; Beraha, D.

    1990-01-01

    An important target in improving reactor safety is to have the most realistic simulation possible of beyond design-base accidents in the computer. This paper presents new developments in ATHLET and further developments (description of the thermo-fluid-dynamic conditions in the core and cooling circuits during serious incidents in the computer programme ATHLET-SA) and extensions (link-up to RALOC). RALOC is a computer programme for describing thermodynamic conditions inside the containment during design-base accidents and accidents involving core meltdown. Further research is dedicated to code acceleration. (DG) [de

  7. Environmental conditions using thermal-hydraulics computer code GOTHIC for beyond design basis external events

    International Nuclear Information System (INIS)

    Pleskunas, R.J.

    2015-01-01

    In response to the Fukushima Dai-ichi beyond design basis accident in March 2011, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, 'Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies Beyond-Design-Basis-External-Events'. To outline the process to be used by individual licensees to define and implement site-specific diverse and flexible mitigation strategies (FLEX) that reduce the risks associated with beyond design basis conditions, Nuclear Energy Institute document NEI 12-06, 'Diverse and Flexible Coping Strategies (FLEX) Implementation Guide', was issued. A beyond design basis external event (BDBEE) is postulated to cause an Extended Loss of AC Power (ELAP), which will result in a loss of ventilation which has the potential to impact room habitability and equipment operability. During the ELAP, portable FLEX equipment will be used to achieve and maintain safe shutdown, and only a minimal set of instruments and controls will be available. Given these circumstances, analysis is required to determine the environmental conditions in several vital areas of the Nuclear Power Plant. The BDBEE mitigating strategies require certain room environments to be maintained such that they can support the occupancy of personnel and the functionality of equipment located therein, which is required to support the strategies associated with compliance to NRC Order EA-12-049. Three thermal-hydraulic analyses of vital areas during an extended loss of AC power using the GOTHIC computer code will be presented: 1) Safety-related pump and instrument room transient analysis; 2) Control Room transient analysis; and 3) Auxiliary/Control Building transient analysis. GOTHIC (Generation of Thermal-Hydraulic Information for Containment) is a general purpose thermal-hydraulics software package for the analysis of nuclear power plant containments, confinement buildings, and system components. It is a volume/path/heat sink

  8. Design study on dose evaluation method for employees at severe accident

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka; Irie, Takashi; Kohriyama, Tamio; Kudo, Seiichi; Nishimura, Kazuya

    2001-01-01

    When we assume a severe accident in a nuclear power plant, it is required for rescue activity in the plant, accident management, repair work of failed parts and evaluation of employees to obtain radiation dose rate distribution or map in the plant and estimated dose value for the above works. However it might be difficult to obtain them accurately along the progress of the accident, because radiation monitors are not always installed in the areas where the accident management is planned or the repair work is thought for safety-related equipments. In this work, we analyzed diffusion of radioactive materials in case of a severe accident in a pressurized water reactor plant, investigated a method to obtain radiation dose rate in the plant from estimated radioactive sources, made up a prototype analyzing system by modeling a specific part of components and buildings in the plant from this design study on dose evaluation method for employees at severe accident, and then evaluated its availability. As a result, we obtained the followings: (1) A new dose evaluation method was established to predict the radiation dose rate in any point in the plant during a severe accident scenario. (2) This evaluation of total dose including moving route and time for the accident management and the repair work is useful for estimating radiation dose limit for these actions of the employees. (3) The radiation dose rate map is effective for identifying high radiation areas and for choosing a route with lower radiation dose rate. (author)

  9. Design study on dose evaluation method for employees at severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Yoshitaka; Irie, Takashi; Kohriyama, Tamio [Institute of Nuclear Safety Systems Inc., Mihama, Fukui (Japan); Kudo, Seiichi [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Nishimura, Kazuya [Computer Software Development Co., Ltd., Tokyo (Japan)

    2001-09-01

    When we assume a severe accident in a nuclear power plant, it is required for rescue activity in the plant, accident management, repair work of failed parts and evaluation of employees to obtain radiation dose rate distribution or map in the plant and estimated dose value for the above works. However it might be difficult to obtain them accurately along the progress of the accident, because radiation monitors are not always installed in the areas where the accident management is planned or the repair work is thought for safety-related equipments. In this work, we analyzed diffusion of radioactive materials in case of a severe accident in a pressurized water reactor plant, investigated a method to obtain radiation dose rate in the plant from estimated radioactive sources, made up a prototype analyzing system by modeling a specific part of components and buildings in the plant from this design study on dose evaluation method for employees at severe accident, and then evaluated its availability. As a result, we obtained the followings: (1) A new dose evaluation method was established to predict the radiation dose rate in any point in the plant during a severe accident scenario. (2) This evaluation of total dose including moving route and time for the accident management and the repair work is useful for estimating radiation dose limit for these actions of the employees. (3) The radiation dose rate map is effective for identifying high radiation areas and for choosing a route with lower radiation dose rate. (author)

  10. Design basis flood for nuclear power plants on river sites

    International Nuclear Information System (INIS)

    1983-01-01

    The Guide presents techniques for determining the design basis flood (DBF) to be used for siting nuclear power plants at or near non-tidal reaches of rivers and for protecting nuclear power plants against floods. Since flooding of a nuclear power plant can have repercussions on safety, the DBF is always chosen to have a very low probability of exceedance per annum. The DBF may result from one or more of the following causes: (1) Precipitation, snowmelt; (2) Failure of water control structures, either from seismic or hydrological causes or from faulty operation of these structures; (3) Channel obstruction such as landslide, ice effects, log or debris jams, and effects of vulcanism. Normally the DBF is not less than any recorded or historical flood occurrence. For flood evaluation two types of methods are discussed in this Guide: probabilistic and deterministic. Simple probabilistic methods to determine floods of such low exceedance probability have a great degree of uncertainty and are presented for use only during the site survey. However, the more sophisticated probabilistic methods, the so-called stochastic methods, may give an acceptable result, as outlined in this Guide. The preferred method of evaluating the component of the DBF due to precipitation, as described in this Guide, is the deterministic one, based on the concept of a limit to the probable maximum precipitation (PMP) and on the unit hydrograph technique. Dam failures may generate a flood substantially more severe than that due to precipitation. The methodology for evaluating these types of floods is therefore presented in this Guide. Making allowance for the possible simultaneous occurrence of two or more important flood-producing events is also discussed here. The Guide does not deal with floods caused by sabotage

  11. MOV motor and gearbox performance under design basis loads

    International Nuclear Information System (INIS)

    DeWall, K.G.; Watkins, J.C.

    1998-01-01

    This paper describes the results of valve testing sponsored by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research and conducted at the Idaho National Engineering and Environmental Laboratory. The research objective was to evaluate the capabilities of specific actuator motor and gearbox assemblies under various design basis loading conditions. The testing was performed using the motor-operated valve load simulator, a test fixture that simulates the stem load profiles a valve actuator would experience when closing a valve against flow and pressure loadings. The authors tested five typical motors (four ac motors and one dc motor) with three gearbox assemblies at conditions a motor might experience in a power plant, including such off-normal conditions as operation at high temperature and reduced voltage. The authors also determined the efficiency of the actuator gearbox. The testing produced the following significant results: all five motors operated at or above their rated torque during tests at full voltage and ambient temperature; for all five motors (dc as well as ac), the actual torque loss due to voltage degradation was greater than the torque loss predicted using common methods; startup torques in locked rotor tests compared well with stall torques in dynamometer-type tests; the methods commonly used to predict torque losses due to elevated operating temperatures sometimes bounded the actual losses, but not in all cases; the greatest discrepancy involved the prediction for the dc motor; running efficiencies published by the manufacturer for actuator gearboxes were higher than the actual efficiencies determined from testing, in some instances, the published pullout efficiencies were also higher than the actual values; operation of the gearbox at elevated temperature did not affect the operating efficiency

  12. Session 1 theme: Various forms of design basis knowledge and effects of its loss on Safety. Views from EDF

    International Nuclear Information System (INIS)

    Servière, Georges

    2013-01-01

    Design basis knowledge - What happens or may happen and corresponding required knowledge: • Unexpected events or failures of equipment; • Spare part issues (no longer availaible,…); • Change in applicable regulations / requirements; • Change of operating conditions; • Change of plant performances; • Evolution of external environment and conditions; • Events and accidents on other plants, worldwide; • New knowledge availaible; • Periodic safety reviews and upgrades; • Extension of plant operation life; • Decommissioning and dismantling; • Some of those you may choose not to do, but most of them have to be faced and need appropriate knowledge

  13. Accident management approach in Armenia

    International Nuclear Information System (INIS)

    Ghazaryan, K.

    1999-01-01

    In this lecture the accident management approach in Armenian NPP (ANPP) Unit 2 is described. List of BDBAs had been developed by OKB Gydropress in 1994. 13 accident sequences were included in this list. The relevant analyses had been performed in VNIIAES and the 'Guidelines on operator actions for beyond design basis accident (BDBA) management at ANPP Unit 2' had been prepared. These instructions are discussed

  14. NPP post-accident monitoring system based on unmanned aircraft vehicle:concept, design principles

    International Nuclear Information System (INIS)

    Sachenko, A.A.; Kochan, V.V.; Kharchenko, V.S.; Yanovskij, M.Eh.; Yastrebenetskij, M.A.; Fesenko, G.V.

    2016-01-01

    The paper presents a concept of designing the post-accident system for monitoring the equipment and territory of nuclear power plant after a severe accident based on unmanned aircraft vehicle (UAVs). Wired power and communications networks are found out as the most vulnerable ones during the accident monitoring, and informativity, reliability and veracity are recognized as system basic parameters. It is proposed to equip measurement and control modules with backup wireless communication channels and deploy the repeaters network based on UAVs to ensure the informativity. Modules possess the backup power battery, and repeaters appear in the appropriate places after the accident to provide the survivability. Moreover, an optimization of UAVs' location is proposed according to the minimum energy consumption criterion. To ensure the veracity, it is expected to design the noise-immune protocol for message exchange and archiving and self-diagnostics of all system components

  15. Impact of severe accidents on the European pressurized water reactor (ERP) design and layout

    International Nuclear Information System (INIS)

    Yvon, M.; Lohnert, G.; Lauret, P.; Bittermann, D.

    1998-01-01

    The purpose of this presentation is to describe the impact of severe accidents on the EPR design and layout. After a summary of the safety requirements specified in accordance with the recommendations expressed by the French and German safety authorities, the main EPR features corresponding to the prevention and the mitigation of severe accidents will be described. Considerations with regard to R and D and cost impacts are also provided

  16. The earthquake problem in engineering design: generating earthquake design basis information

    International Nuclear Information System (INIS)

    Sharma, R.D.

    1987-01-01

    Designing earthquake resistant structures requires certain design inputs specific to the seismotectonic status of the region, in which a critical facility is to be located. Generating these inputs requires collection of earthquake related information using present day techniques in seismology and geology, and processing the collected information to integrate it to arrive at a consolidated picture of the seismotectonics of the region. The earthquake problem in engineering design has been outlined in the context of a seismic design of nuclear power plants vis a vis current state of the art techniques. The extent to which the accepted procedures of assessing seismic risk in the region and generating the design inputs have been adherred to determine to a great extent the safety of the structures against future earthquakes. The document is a step towards developing an aproach for generating these inputs, which form the earthquake design basis. (author)

  17. Reduced design load basis for ultimate blade loads estimation in multidisciplinary design optimization frameworks

    DEFF Research Database (Denmark)

    Pavese, Christian; Tibaldi, Carlo; Larsen, Torben J.

    2016-01-01

    The aim is to provide a fast and reliable approach to estimate ultimate blade loads for a multidisciplinary design optimization (MDO) framework. For blade design purposes, the standards require a large amount of computationally expensive simulations, which cannot be efficiently run each cost...... function evaluation of an MDO process. This work describes a method that allows integrating the calculation of the blade load envelopes inside an MDO loop. Ultimate blade load envelopes are calculated for a baseline design and a design obtained after an iteration of an MDO. These envelopes are computed...... for a full standard design load basis (DLB) and a deterministic reduced DLB. Ultimate loads extracted from the two DLBs with the two blade designs each are compared and analyzed. Although the reduced DLB supplies ultimate loads of different magnitude, the shape of the estimated envelopes are similar...

  18. German (GRS) approach to accident analysis (part I). German licensing basis for accident analyses. Applicants accident analyses in second part license for Konvoi-plants. Appendix 1. Assessor accident analyses in second part license for Konvoi-plants. Appendix 2. Reference list of DBA to be considered in the safety status analysis of a PSR. Appendix 3a. Reference list of special very rare and BDB plant conditions to be considered in the safety status analysis of a PSE. Appendix 3b

    International Nuclear Information System (INIS)

    Velkov, K.

    2002-01-01

    LOCA analyses.Appendix 3a: Concerns Level 3, accidents and events to be considered for transients, accidents and events to be considered for losses of coolant (LOCA), accidents and radiologically representative events, as well as PWR-specific spreading impacts to be considered as possible initiating events for transients and SBLOCA. These events are defined for PWR and BWR. Appendix 3b: Level 4, PWR- and BWR-specific special, very rare events,(ATWS and site-specific external civil impacts (certain emergencies)), beyond-design-basis plant conditions. These event and accident as in previous Appendix is defined again for PWR and BWR reactor types

  19. Design and development of self-powered sensors on wireless sensor network for standalone plant critical data management during SBO and beyond design basis events

    International Nuclear Information System (INIS)

    Aparna, J.; Dulera, I.V.; Rama Rao, A.; Vijayan, P.K.

    2015-01-01

    Advanced reactors are designed with an aim of maximum safety, optimized fuel utilization and effective system design. Safety aspects in reactor designs are being viewed for all possible vulnerabilities, and as a result, robust self-regulating passive safety features have been favored in Gen IV and advanced reactor designs. In addition to passive systems, the accidents scenarios at Fukushima indicate the dire need of reliable and stand-alone self-powered sensors, for monitoring plant critical parameters for effective damage control actions. There is a strong need for plant critical data management and situation awareness during the unavailability of all conventional power sources in a nuclear power plant, during extended station blackout (SBO) conditions. These self-powered sensors would assist the operators in managing events like SBO and help in containing any Beyond Design Basis Events (BDBE) conditions, well away from the public domain

  20. Radionuclide release calculations for selected severe accident scenarios. Volume 3. PWR, subatmospheric containment design

    International Nuclear Information System (INIS)

    Denning, R.S.; Gieseke, J.A.; Cybulskis, P.; Lee, K.W.; Jordan, H.; Curtis, L.A.; Kelly, R.F.; Kogan, V.; Schumacher, P.M.

    1986-07-01

    This report presents results of analyses of the enviromental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with a subatmospheric containment design. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the US Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. the Surry plant has been used in this study as the reference plant for a subatmospheric design

  1. Development of Mitigation Strategy for Beyond Design Basis External Events for NRC Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Hak; Lee, Jae Jong; Kim, Myung Ki [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, how to develop FLEX strategy for beyond-design-basis external events for U. S. NRC design certification is examined. The development method of FLEX strategy for U. S. NRC design certification is examined. The applicants should make unit-specific FLEX strategy and establish the minimum coping capabilities consistent with unit-specific evaluation of the potential impacts and responses to BDBEEs. NEI 12-06 outlines the process to define and deploy the diverse and flexible mitigation strategies(FLEX strategy) that will increase defense-in-depth for beyond-design-basis scenarios to address the extended loss of alternating current (ac) power (ELAP) and loss of normal access to the ultimate heat sink (LUHS) occurring simultaneously at all units on a site. The order (EA-12-049) is issued to all reactor licensees, including holders of active, Construction Permit (CP) holders, and Combined License (COL) holders. Applicants for the new reactor design certification should prepare and submit FLEX strategy for NRC staff's review. Site-specific data related with the new reactor can't be determined during the new reactor design certification applications so that the unit-specific FLEX strategy should be developed.

  2. Development of Mitigation Strategy for Beyond Design Basis External Events for NRC Design Certification

    International Nuclear Information System (INIS)

    Kim, Dong Hak; Lee, Jae Jong; Kim, Myung Ki

    2013-01-01

    In this study, how to develop FLEX strategy for beyond-design-basis external events for U. S. NRC design certification is examined. The development method of FLEX strategy for U. S. NRC design certification is examined. The applicants should make unit-specific FLEX strategy and establish the minimum coping capabilities consistent with unit-specific evaluation of the potential impacts and responses to BDBEEs. NEI 12-06 outlines the process to define and deploy the diverse and flexible mitigation strategies(FLEX strategy) that will increase defense-in-depth for beyond-design-basis scenarios to address the extended loss of alternating current (ac) power (ELAP) and loss of normal access to the ultimate heat sink (LUHS) occurring simultaneously at all units on a site. The order (EA-12-049) is issued to all reactor licensees, including holders of active, Construction Permit (CP) holders, and Combined License (COL) holders. Applicants for the new reactor design certification should prepare and submit FLEX strategy for NRC staff's review. Site-specific data related with the new reactor can't be determined during the new reactor design certification applications so that the unit-specific FLEX strategy should be developed

  3. Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Su, S.F.; Cahalan, J.E.; Sevy, R.H.

    1985-01-01

    This paper presents the results of a systematic study of the effectiveness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). The results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs, and safety margins are quantified in sensitivity studies. All analyses were carried out using the SASSYS LMFBR systems analysis code (1)

  4. Design Safety Considerations for Water Cooled Small Modular Reactors Incorporating Lessons Learned from the Fukushima Daiichi Accident

    International Nuclear Information System (INIS)

    2016-03-01

    The global future deployment of advanced nuclear reactors for electricity generation depends primarily on the ability of nuclear industries, utilities and regulatory authorities to further enhance their reliability and economic competitiveness while satisfying stringent safety requirements. The IAEA has a project to help coordinate Member States efforts in the development and deployment of small and medium sized or small modular reactor (SMR) technology. This project aims simultaneously to facilitate SMR technology developers and potential SMR uses, particularly States embarking on a nuclear power programme, in identifying key enabling technologies and enhancing capacity building by resolving issues relevant to deployment, including nuclear reactor safety. The objective of this publication is to explore common practices for Member States, which will be an essential resource for future development and deployment of SMR technology. The accident at the Fukushima Daiichi nuclear power plant was caused by an unprecedented combination of natural events: a strong earthquake, beyond the design basis, followed by a series of tsunamis of heights exceeding the design basis tsunami considered in the flood analysis for the site. Consequently, all the operating nuclear power plants and advanced reactors under development, including SMRs, have been incorporating lessons learned from the accident to assure and enhance the performance of the engineered safety features in coping with such external events

  5. Design Basis for Fibre Reinforced Concrete (FRC) Pavements

    DEFF Research Database (Denmark)

    Bendixen, Søren; Stang, Henrik

    1996-01-01

    -crack opening relationship can beused to descibe the properties of fibre reinforced concrete (FRC) intension and how the stress-crack opening relationship can beapplied in a simple design scheme for pavements. The projectincludes development of design tools, experiments to determine thestress-crack opening...

  6. Selection of design basis event for modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

    2016-06-01

    Japan Atomic Energy Agency (JAEA) has been investigating safety requirements and basic approach of safety guidelines for modular High Temperature Gas-cooled Reactor (HTGR) aiming to increase internarial contribution for nuclear safety by developing an international HTGR safety standard under International Atomic Energy Agency. In this study, we investigate a deterministic approach to select design basis events utilizing information obtained from probabilistic approach. In addition, selections of design basis events are conducted for commercial HTGR designed by JAEA. As a result, an approach for selecting design basis event considering multiple failures of safety systems is established which has not been considered as design basis in the safety guideline for existing nuclear facility. Furthermore, selection of design basis events for commercial HTGR has completed. This report provides an approach and procedure for selecting design basis events of modular HTGR as well as selected events for the commercial HTGR, GTHTR300. (author)

  7. CIF---Design basis for an integrated incineration facility

    International Nuclear Information System (INIS)

    Bennett, G.F.

    1991-01-01

    This paper discusses the evolution of chosen technologies that occurred during the design process of the US Department of Energy (DOE) incineration system designated the Consolidated Incineration Facility (CIF) as the Savannah River Plant, Aiken, South Carolina. The Plant is operated for DOE by the Westinghouse Savannah River Company. The purpose of the incineration system is to treat low level radioactive and/or hazardous liquid and solid wastes by combustion. The objective for the facility is to thermally destroy toxic constituents and volume reduce waste material. Design criteria requires operation be controlled within the limits of RCRA's permit envelope

  8. Main aspects of the design of a support structure of a LMFBR with particular reference to the explosive accident consequences

    International Nuclear Information System (INIS)

    Giuliano, V.; Lazzeri, L.

    1977-01-01

    The aim of this paper is a review of the main aspects of the design of a support structure of a LMFBR tank, with particular reference to the analysis of the non-linear dynamic behaviour of the structure in the plastic range under the effect of an explosive accident within the tank. The structure is composed by a L-shaped flange, which supports the tank, connected by means of nine square beams to a rigid box-type ring, fixed to the concrete. The plug of the tank is connected to the L-shaped figure by means of a group of SS bars. The non-linear dynamic analysis of the explosive accident has been carried out on a lumped mass model, with elastic-plastic elements which simulate main components of the support structure and tank. The impulsive load connected to the explosive accident has been modelled (on the basis of extensive comparative studies carried out) as two triangular pressure impulses acting on the plug and on the botton of the tank. A large amount of results, which describe displacements, velocities and accelerations of the plug, of the tank, and of the support structure, together with the forces and stresses acting on the main structural components are presented and discussed, with particular reference to the influence of the various parameters involved in the analysis. (Auth.)

  9. Design basis for the operational modelling of the atmospheric dispersion

    International Nuclear Information System (INIS)

    Doury, A.

    1987-10-01

    Based on the latest practices at the Institut de Protection et de Surete Nucleaire of the Commissariat a l'Energie Atomique (CEA), we shall first present the basis elements used for a simple and adequate modelling method for assessing hypothetical atmospheric pollution from transient or continuous discharge with any given kinetics under various weather conditions which are not necessarily stationary or uniform, which are likely to occur even with little or no wind. Discharges shall be considered as sequences of instantaneous successive puffs. The parameters deduced experimentally or from observations are functions of the transfer time and cover all time and space scales. The restrictions of use are indicated, especially concerning heavy gases. Finally, simple formulas are proposed for concentrations and depositions so as to be able to make a rapid estimation of the orders of magnitude with almost no computation [fr

  10. Design basis for the operational modelling of the atmospheric dispersion

    International Nuclear Information System (INIS)

    Doury, A.

    1987-11-01

    Based on the latest practices at the Institut de Protection et de Surete Nucleaire of the Commissariat a l'Energie Atomique (CEA), we shall first present the basis elements used for a simple and adequate modelling method for assessing hypothetical atmospheric pollution from transient or continuous discharge with any given kinetics under various weather conditions which are not necessarily stationary or uniform, which are likely to occur even with little or no wind. Discharges shall be considered as sequences of instantaneous successive puffs. The parameters deduced experimentally or from observations are functions of the transfer time and cover all time and space scales. The restrictions of use are indicated, especially concerning heavy gases. Finally, simple formulas are proposed for concentrations and depositions so as to be able to make a rapid estimation of the orders of magnitude with almost no computation [fr

  11. Electrical systems design applications on Japanese PWR plants in light of the Fukushima Daiichi Accident

    International Nuclear Information System (INIS)

    Nomoto, Tsutomu

    2015-01-01

    After the Fukushima Daiichi nuclear power plant (1F-NPP) accident (i.e. Station Blackout), several design enhancements have been incorporated or are under considering to Mitsubishi PWR plants' design of not only operational plants' design but also new plants' design. Especially, there are several important enhancements in the area of the electrical system design. In this presentation, design enhancements related to following electrical systems/equipment are introduced; - Offsite Power System; - Emergency Power Source; - Safety-related Battery; - Alternative AC Power Supply Systems. In addition, relevant design requirements/conditions which are or will be considered in Mitsubishi PWR plants are introduced. (authors)

  12. Assessment of Loads and Performance of a Containment in a Hypothetical Accident (ALPHA). Facility design report

    International Nuclear Information System (INIS)

    Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Moriyama, Kiyofumi; Ito, Hideo; Komori, Keiichi; Sonobe, Hisao; Sugimoto, Jun

    1998-06-01

    In the ALPHA (Assessment of Loads and Performance of Containment in Hypothetical Accident) program, several tests have been performed to quantitatively evaluate loads to and performance of a containment vessel during a severe accident of a light water reactor. The ALPHA program focuses on investigating leak behavior through the containment vessel, fuel-coolant interaction, molten core-concrete interaction and FP aerosol behavior, which are generally recognized as significant phenomena considered to occur in the containment. In designing the experimental facility, it was considered to simulate appropriately the phenomena mentioned above, and to cover experimental conditions not covered by previous works involving high pressure and temperature. Experiments from the viewpoint of accident management were also included in the scope. The present report describes design specifications, dimensions, instrumentation of the ALPHA facility based on the specific test objectives and procedures. (author)

  13. Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burnup condition

    International Nuclear Information System (INIS)

    Kim, H. D.; Kim, D. H.; Park, S. Y.; Park, J. H.

    2005-10-01

    This project was performed by KAERI in the frame of construction of the international cooperative basis on the nuclear energy. This was supported from MOST under the title of 'Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burn up condition'. The current operating NPP are converting the burned fuel to the wasted fuel after burn up of 40 GWD/MTU. But in Korea, burn up of more than 60 GWD/MTU will be expected because of the high fuel efficiency but also cost saving for storing the wasted fuel safely. The domestic research for the purpose of developing the fuel and the cladding that can be used under the high burn up condition up to 100 GWD/MTU is in progress now. But the current computer code adopts the model and the data that are valid only up to the 40 GWD/MTU at most. Therefore the current model could not take into account the phenomena that may cause differences in the fission product release behavior or in the core damage process due to the high burn up operation (more than 40 GWD/MTU). To evaluate the safety of the NPP with the high burn up fuel, the improvement of current severe accident code against the high burn up condition is an important research item. Also it should start without any delay. Therefore, in this study, an expert group was constructed to establish the research basis for the severe accident under high burn up conditions. From this expert group, the research items regarding the high burn up condition were selected and identified through discussion and technical seminars. Based on these selected items, the meeting between IRSN and KAERI to find out the cooperative research items on the severe accident under the high burn up condition was held in the IRSN headquater in Paris. After the meeting, KAERI and IRSN agreed to cooperate with each other on the selected items, and to co-host the international seminar, and to develop the model and to

  14. Design basis reconstitution and configuration management of nuclear power plants

    International Nuclear Information System (INIS)

    Smith, P.R.

    1989-01-01

    The major design requirements of nuclear power plant components, systems, and structures are found in the plant's licensing commitments documented in the Final Safety Analysis Report and in the technical specification commitments of the plant. These specifications consider the original design and its degradation by in-service use. Before a nuclear power plant begins operation, the plant systems, structures, and organizational elements are functionally arranged to operate in a particular way. This functional arrangement is specified by the plant's design requirements and is called its configuration. The paper discusses configuration management and information management for configuration management. The management of large amounts of information and the various information systems associated with nuclear generating facilities is an ever-growing challenge for utilities. Plant operations involve a complex interrelation among data elements, especially in relation to design modifications and operational changes. Consequently, the operation of these data systems is interrelated and, as a result, redundant data items may exist. Thus, in view of the need to control and manage the plant configuration baseline, managers are striving to streamline their information management programs, which usually involves the integration of data-base systems

  15. Simulant Basis for the Standard High Solids Vessel Design

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-01

    This document provides the requirements for a test simulant suitable for demonstrating the mixing requirements for the Single High Solids Vessel Design (SHSVD). This simulant has not been evaluated for other purposes such as gas retention and release or erosion. The objective of this work is to provide an underpinning for the simulant properties based on actual waste characterization.

  16. Cognitive Process as a Basis for Intelligent Retrieval Systems Design.

    Science.gov (United States)

    Chen, Hsinchun; Dhar, Vasant

    1991-01-01

    Two studies of the cognitive processes involved in online document-based information retrieval were conducted. These studies led to the development of five computational models of online document retrieval which were incorporated into the design of an "intelligent" document-based retrieval system. Both the system and the broader implications of…

  17. Ergonomic (human factors) problems in design of NPPs. A review of TMI and Chernobyl accidents

    International Nuclear Information System (INIS)

    Huang Xiangrui; Zheng Fuyu; Gao Jia

    1994-01-01

    The general principle of ergonomic in design of NPPs is given and some causes of TMI and Chernobyl accidents from the view point of human factor engineering are reviewed. The paper also introduces some Ergonomic problems in design, operation and management of earlier NPPs. Some ergonomic principles of man-machine systems design have been described. Some proposals have been suggested for improving human reliability in NPPs

  18. Hot laboratory design on the basis of standardized components

    International Nuclear Information System (INIS)

    Cadrot, J.

    1976-01-01

    The paper describes the principal effects on hot laboratory design brought about over the last 15 years by the use of standardized components developed jointly with the CEA and the industrial associates of AFINE. After a rapid survey of the various advantages of standardization, the author turns to the specific case of a laboratory producing mixed plutonium and uranium oxide fuels, giving a brief description of the glove-boxes and ancillary equipment. He then deals with the design of an isotope production laboratory. The basic component is the DR 200 standard cell, which permits the civil engineering work to be effected on modular principles. Use of a safety-flow pressure regulating valve makes possible pneumatic automation of the production-cell internals. A substantial gain in output is the result. In the next section the paper refers to a pilot facility for irradiated fuel studies, and describes the components used, which require taking into account the high activities and intense radiations encountered in studies of this type. The author then demonstrates the flexibility with which standardized components can be adapted to different uses, thus solving many distinct problems, an example of which is represented by a semi-hot box for handling up to 100g of americium-241. Finally, the paper offers a rapid summary of the effects of standardization at the various stages concerned, from initial design to the commissioning of a hot laboratory. (author)

  19. Exploratory Shaft Seismic Design Basis Working Group report

    International Nuclear Information System (INIS)

    Subramanian, C.V.; King, J.L.; Perkins, D.M.; Mudd, R.W.; Richardson, A.M.; Calovini, J.C.; Van Eeckhout, E.; Emerson, D.O.

    1990-08-01

    This report was prepared for the Yucca Mountain Project (YMP), which is managed by the US Department of Energy. The participants in the YMP are investigating the suitability of a site at Yucca Mountain, Nevada, for construction of a repository for high-level radioactive waste. An exploratory shaft facility (ESF) will be constructed to permit site characterization. The major components of the ESF are two shafts that will be used to provide access to the underground test areas for men, utilities, and ventilation. If a repository is constructed at the site, the exploratory shafts will be converted for use as intake ventilation shafts. In the context of both underground nuclear explosions (conducted at the nearby Nevada Test Site) and earthquakes, the report contains discussions of faulting potential at the site, control motions at depth, material properties of the different rock layers relevant to seismic design, the strain tensor for each of the waveforms along the shaft liners, and the method for combining the different strain components along the shaft liners. The report also describes analytic methods, assumptions used to ensure conservatism, and uncertainties in the data. The analyses show that none of the shafts' structures, systems, or components are important to public radiological safety; therefore, the shafts need only be designed to ensure worker safety, and the report recommends seismic design parameters appropriate for this purpose. 31 refs., 5 figs., 6 tabs

  20. Main aspects of the design of a support structure of a LMFBR with particular reference to the explosive accident consequences

    International Nuclear Information System (INIS)

    Giuliano, V.; Lazzeri, L.

    1977-01-01

    The aim of this paper is a review of the main aspects of the design of a support structure of a LMFBR tank, with particular reference to the analysis of the non-linear dynamic behavior of the structure in the plastic range under the effect of an explosive accident within the tank. The structure is composed by a L-shaped flange, which supports the tank, connected by means of nine square beams to a rigid box-type ring, fixed to the concrete. The plug of the tank is connected to the L-shaped flange by means of a group of SS bars. The non-linear dynamic analysis of the explosive accident has been carried out on a lumped mass model, with elastic-plastic elements which simulate main components of the support structure and tank. The impulsive load connected to the explosive accident has been modelled (on the basis of extensive comparative studies carried out) as two triangular pressure impulses has been the object of a parametric evaluation. The dynamic transient on the support structure during and after the explosive accident for each couple of pressure impulses has been analyzed by means of modified version of the NON SAP code running on a CDC 7600 computer. A large amount of results, which describe displacements, velocities and accelerations of the plug, of the tank, and of the support structure, together with the forces and stresses acting on the main structural components are presented and discussed, with particular reference to the influence of the various parameters involved in the analysis

  1. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    International Nuclear Information System (INIS)

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment

  2. Design and implementation of an identification system in construction site safety for proactive accident prevention.

    Science.gov (United States)

    Yang, Huanjia; Chew, David A S; Wu, Weiwei; Zhou, Zhipeng; Li, Qiming

    2012-09-01

    Identifying accident precursors using real-time identity information has great potential to improve safety performance in construction industry, which is still suffering from day to day records of accident fatality and injury. Based on the requirements analysis for identifying precursor and the discussion of enabling technology solutions for acquiring and sharing real-time automatic identification information on construction site, this paper proposes an identification system design for proactive accident prevention to improve construction site safety. Firstly, a case study is conducted to analyze the automatic identification requirements for identifying accident precursors in construction site. Results show that it mainly consists of three aspects, namely access control, training and inspection information and operation authority. The system is then designed to fulfill these requirements based on ZigBee enabled wireless sensor network (WSN), radio frequency identification (RFID) technology and an integrated ZigBee RFID sensor network structure. At the same time, an information database is also designed and implemented, which includes 15 tables, 54 queries and several reports and forms. In the end, a demonstration system based on the proposed system design is developed as a proof of concept prototype. The contributions of this study include the requirement analysis and technical design of a real-time identity information tracking solution for proactive accident prevention on construction sites. The technical solution proposed in this paper has a significant importance in improving safety performance on construction sites. Moreover, this study can serve as a reference design for future system integrations where more functions, such as environment monitoring and location tracking, can be added. Copyright © 2011 Elsevier Ltd. All rights reserved.

  3. Design basis for the safe disposal of radioactive waste

    International Nuclear Information System (INIS)

    Lewi, J.; Kaluzny, Y.

    1990-01-01

    All radioactive waste disposal sites, regardless of disposal concept, are designed to isolate the radioactive substances contained in such waste for a period at least equal to the time it may remain potentially harmful. Isolation is achieved through the use of containment barriers. This paper summarises the function and limits of different types of barrier used in various disposal systems. For each type of barrier, the paper describes and comments on the site selection criteria and waste packaging requirements applicable in various countries. 13 refs., 1 fig [fr

  4. Radionuclide release calculations for selected severe accident scenarios. PWR, ice condenser design

    Energy Technology Data Exchange (ETDEWEB)

    Denning, R S; Gieseke, J A; Cybulskis, P; Lee, K W; Jordan, H; Curtis, L A; Kelly, R F; Kogan, V; Schumacher, P M

    1986-07-01

    This report presents results of analyses of the environmental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with an ice-condenser containment. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the U.S. Nuclear Regulatory Commission by Sandia National Laboratories. In the SARRP program, risk estimates are being generated for a number of reference plant designs. The Sequoyah Plant has been used in this study as an example of a PWR ice-condenser plant. (author)

  5. Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

    Science.gov (United States)

    Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos

    2012-06-01

    The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High

  6. Improved Design Basis for Laterally Loaded Large Diameter Pile

    DEFF Research Database (Denmark)

    Leth, Caspar Thrane

    of the diameter, depth and soil strength, and increase of each these will give an increase in stiffness. • Cyclic response of a lateral loaded pile is depended on the characteristics of the cyclic load. Behaviour of a monopile is a classic soil-structure interaction problem depending on the pile stiffness....... The target is to improve the use of monopiles as preferred support structure beyond the current limit at a water depth of 30 m. Design of foundations for wind turbines has a large focus on the stiffness of the combined structure, turbine-tower-foundation, which has an influence on the environmental loads...... initial response and a higher ultimate capacity. The initial stiffness of the soil-structure interaction measured in the centrifuge tests, equivalent to initial stiffness of p-y curves, shows a dependency of depth and diameter. Control issues in relation to cyclic tests have resulted in tests...

  7. Guide to radiological accident considerations for siting and design of DOE nonreactor nuclear facilities

    International Nuclear Information System (INIS)

    Elder, J.C.; Graf, J.M.; Dewart, J.M.; Buhl, T.E.; Wenzel, W.J.; Walker, L.J.; Stoker, A.K.

    1986-01-01

    This guide was prepared to provide the experienced safety analyst with accident analysis guidance in greater detail than is possible in Department of Energy (DOE) Orders. The guide addresses analysis of postulated serious accidents considered in the siting and selection of major design features of DOE nuclear facilities. Its scope has been limited to radiological accidents at nonreactor nuclear facilities. The analysis steps addressed in the guide lead to evaluation of radiological dose to exposed persons for comparison with siting guideline doses. Other possible consequences considered are environmental contamination, population dose, and public health effects. Choices of models and parameters leading to estimation of source terms, release fractions, reduction and removal factors, dispersion and dose factors are discussed. Although requirements for risk analysis have not been established, risk estimates are finding increased use in siting of major nuclear facilities, and are discussed in the guide. 3 figs., 9 tabs

  8. Design study on dose evaluation method for employees at severe accident

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka; Irie, Takashi; Kohriyama, Tamio; Kudo, Seiichi; Nishimura, Kazuya

    2002-01-01

    If a severe accident occurs in a pressurized water reactor plant, it is required to estimate dose values of operators engaged in emergency such as accident management, repair of failed parts. However, it might be difficult to measure radiation dose rate during the progress of an accident, because radiation monitors are not always installed in areas where the emergency activities are required. In this study, we analyzed the transport of radioactive materials in case of a severe accident, investigated a method to obtain radiation dose rate in the plant from estimated radioactive sources, made up a prototype analyzing system from this design study, and then evaluated its availability. As a result, we obtained the following: (1) A new dose evaluation method was established to predict the radiation dose rate at any point in the plant during a severe accident scenario. (2) This evaluation of total dose including access route and time for emergency activities is useful for estimating radiation dose limit for these employee actions. (3) The radiation dose rate map is effective for identifying high radiation areas and for choosing a route with lower radiation dose rate. (author)

  9. Design considerations for post accident monitoring system of a research reactor

    International Nuclear Information System (INIS)

    Jang, Gwi Sook; Park, Je Yun; Kim, Young Ki

    2012-01-01

    The Post Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. The PAMS of NPP (Nuclear Power Plant) in KOREA provides the continuous display of the PAM category 1 parameters specified in R.G 1.97, Rev. 03. Recently the PAMS of NPP has been designed according to R.G 1.97, Rev. 04. There is no PAMS at the HANARO in KOREA, but recently RRs (Research Reactors) around the world are going to have PAMS for various multi purposes. We should determine the design considerations for PAMS in a Korean RR based on the design state analysis. Thus, this paper proposes strategies on the design considerations for the PAMS of a Korean RR

  10. 10 CFR 72.94 - Design basis external man-induced events.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Design basis external man-induced events. 72.94 Section 72... WASTE Siting Evaluation Factors § 72.94 Design basis external man-induced events. (a) The region must be examined for both past and present man-made facilities and activities that might endanger the proposed...

  11. 46 CFR 177.310 - Satisfactory service as a design basis.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Satisfactory service as a design basis. 177.310 Section... (UNDER 100 GROSS TONS) CONSTRUCTION AND ARRANGEMENT Hull Structure § 177.310 Satisfactory service as a design basis. When scantlings for the hull, deckhouse, and frames of the vessel differ from those...

  12. 10 CFR 50.67 - Accident source term.

    Science.gov (United States)

    2010-01-01

    ... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The... 10 Energy 1 2010-01-01 2010-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... to January 10, 1997, who seek to revise the current accident source term used in their design basis...

  13. Case Study for Enhanced Accident Tolerance Design Changes

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Koonce, Tony [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-06-01

    The ability to better characterize and quantify safety margin is important to improved decision making about Light Water Reactor (LWR) design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant systems, structures, and components (SSCs) needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, reliability, and safety, the Risk Informed Safety Margin Characterization (RISMC) Pathway provides methods and tools that enable mitigation options known as risk informed margins management (RIMM) strategies.

  14. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role...... for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented...

  15. Applying Functional Modeling for Accident Management of Nucler Power Plant

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role...... for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented....

  16. Severe accidents at nuclear power plants. Their risk assessment and accident management

    International Nuclear Information System (INIS)

    Abe, Kiyoharu.

    1995-05-01

    This document is to explain the severe accident issues. Severe Accidents are defined as accidents which are far beyond the design basis and result in severe damage of the core. Accidents at Three Mild Island in USA and at Chernobyl in former Soviet Union are examples of severe accidents. The causes and progressions of the accidents as well as the actions taken are described. Probabilistic Safety Assessment (PSA) is a method to estimate the risk of severe accidents at nuclear reactors. The methodology for PSA is briefly described and current status on its application to safety related issues is introduced. The acceptability of the risks which inherently accompany every technology is then discussed. Finally, provision of accident management in Japan is introduced, including the description of accident management measures proposed for BWRs and PWRs. (author)

  17. Criticality accident of nuclear fuel facility. Think back on JCO criticality accident

    International Nuclear Information System (INIS)

    Naito, Keiji

    2003-09-01

    This book is written in order to understand the fundamental knowledge of criticality safety or criticality accident of nuclear fuel facility by the citizens. It consists of four chapters such as critical conditions and criticality accident of nuclear facility, risk of criticality accident, prevention of criticality accident and a measure at an occurrence of criticality accident. A definition of criticality, control of critical conditions, an aspect of accident, a rate of incident, damage, three sufferers, safety control method of criticality, engineering and administrative control, safety design of criticality, investigation of failure of safety control of JCO criticality accident, safety culture are explained. JCO criticality accident was caused with intention of disregarding regulation. It is important that we recognize the correct risk of criticality accident of nuclear fuel facility and prevent disasters. On the basis of them, we should establish safety culture. (S.Y.)

  18. LFR core design for prevention & mitigation of severe accidents

    International Nuclear Information System (INIS)

    Grasso, Giacomo

    2012-01-01

    Conclusions: • Aiming at fully complying Gen-IV safety requirements – even in case of Fukushima-like events –, prevention and mitigation strategies must be stressed in FR design. • The safety of Lead-cooled Fast Reactors can rely on intrinsic features due to the coolant, such as: • the practical impossibility of Lead boiling, hence the unreliability of core (only) voiding for wide safety margins, and the retention of corium; • the high density of lead, for the buoyancy of Control Rods (allowing their safe positioning below the core), and the dispersion of molten core up to the setting up of a “cold melting pot”. • the possibility to adopt wide coolant channels for encouraging natural circulation, without affecting the hardness of the neutron spectrum; • the hard neutron spectrum allows the adiabatic operation of LFRs (which implies minimal criticality swings even through long cycles) with small amounts of Mas (hence with a negligible detriment to the safety features); • an effective reduction of the coolant density effect simply through the shortening of the active height

  19. The role of usability in the evaluation of accidents: human error or design flaw?

    Science.gov (United States)

    Correia, Walter; Soares, Marcelo; Barros, Marina; Campos, Fábio

    2012-01-01

    This article aims to highlight the role of consumer products companies in the heart and the extent of accidents involving these types of products, and as such undesired events take part as an agent in influencing decision making for the purchase of a product that nature on the part of consumers and users. The article demonstrates, by reference, interviews and case studies such as the development of poorly designed products and design errors of design can influence the usage behavior of users, thus leading to accidents, and also negatively affect the next image of a company. The full explanation of these types of questions aims to raise awareness, plan on a reliable usability, users and consumers in general about the safe use of consumer products, and also safeguard their rights before a legal system of consumer protection, even far away by the CDC--Code of Consumer Protection.

  20. Health effects model for nuclear power plant accident consequence analysis. Part I. Introduction, integration, and summary. Part II. Scientific basis for health effects models

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.S.; Moeller, D.W.; Cooper, D.W.

    1985-07-01

    Analysis of the radiological health effects of nuclear power plant accidents requires models for predicting early health effects, cancers and benign thyroid nodules, and genetic effects. Since the publication of the Reactor Safety Study, additional information on radiological health effects has become available. This report summarizes the efforts of a program designed to provide revised health effects models for nuclear power plant accident consequence modeling. The new models for early effects address four causes of mortality and nine categories of morbidity. The models for early effects are based upon two parameter Weibull functions. They permit evaluation of the influence of dose protraction and address the issue of variation in radiosensitivity among the population. The piecewise-linear dose-response models used in the Reactor Safety Study to predict cancers and thyroid nodules have been replaced by linear and linear-quadratic models. The new models reflect the most recently reported results of the follow-up of the survivors of the bombings of Hiroshima and Nagasaki and permit analysis of both morbidity and mortality. The new models for genetic effects allow prediction of genetic risks in each of the first five generations after an accident and include information on the relative severity of various classes of genetic effects. The uncertainty in modeloling radiological health risks is addressed by providing central, upper, and lower estimates of risks. An approach is outlined for summarizing the health consequences of nuclear power plant accidents. 298 refs., 9 figs., 49 tabs.

  1. Health effects model for nuclear power plant accident consequence analysis. Part I. Introduction, integration, and summary. Part II. Scientific basis for health effects models

    International Nuclear Information System (INIS)

    Evans, J.S.; Moeller, D.W.; Cooper, D.W.

    1985-07-01

    Analysis of the radiological health effects of nuclear power plant accidents requires models for predicting early health effects, cancers and benign thyroid nodules, and genetic effects. Since the publication of the Reactor Safety Study, additional information on radiological health effects has become available. This report summarizes the efforts of a program designed to provide revised health effects models for nuclear power plant accident consequence modeling. The new models for early effects address four causes of mortality and nine categories of morbidity. The models for early effects are based upon two parameter Weibull functions. They permit evaluation of the influence of dose protraction and address the issue of variation in radiosensitivity among the population. The piecewise-linear dose-response models used in the Reactor Safety Study to predict cancers and thyroid nodules have been replaced by linear and linear-quadratic models. The new models reflect the most recently reported results of the follow-up of the survivors of the bombings of Hiroshima and Nagasaki and permit analysis of both morbidity and mortality. The new models for genetic effects allow prediction of genetic risks in each of the first five generations after an accident and include information on the relative severity of various classes of genetic effects. The uncertainty in modeloling radiological health risks is addressed by providing central, upper, and lower estimates of risks. An approach is outlined for summarizing the health consequences of nuclear power plant accidents. 298 refs., 9 figs., 49 tabs

  2. Experimental study of ballooning and failure of WWER-1000 fuel cans during maximum design basis accident

    International Nuclear Information System (INIS)

    Karetnikov, G.V.; Bogdanov, A.S.; Semishkin, V.P.; Bezrukov, Yu.A.; Trushin, A.M.; Frizen, E.A.

    2001-01-01

    The processes of ballooning and fracturing in tubular specimens of Eh635 and Eh110 alloy fuel cans are investigated with the use of cinematography. The investigations are carried out under steady-state conditions in the temperature range from 680 to 900 deg C and at pressure drops on the can from 2 to 12 MPa. Time dependences of circumferential strains are plotted for various temperatures of fuel cans at pressure of 2 MPa. It is shown that strain changes are of linear character at an initial portion of the curve and then an accelerated strain development takes place with transition to fracture. Using methods of nonlinear evaluation for time to fracture the approximation dependences are obtained for fuel cans. Experimental data are intended to form the equations of state for fuel can materials and to verify the program TVEL-3 [ru

  3. The effect of severe accident mitigation design on the containment performance for Korean ALWR

    International Nuclear Information System (INIS)

    Na, J. H.; Lee, J. S.; Lim, H. K.; Kim, J. K.

    2001-01-01

    The containment performance analysis for Korean ALWR standard design has been performed to confirm the safety goal and to identify the design features vulnerable to severe accidents for the on-going design. The results in terms of conditional containment failure probability show Korean ALWR design does not have any particular vulnerability given core damage sequences. It shows the conditional containment failure probability for pull power internal event is less than that of design goal. The late containment failure is much less than 4% for given core damages and that of containment bypass is about 2%. New design features of the Korean ALWR such as bydrogen mitigation system (IIMS), cavity flooding system (CFS), and emergency containment spray bakcup system (ECSBS), external reactor vessel cooling (ERVC), etc. are reflected in Korean ALWR design and is reviewed in this paper to give an insight for the design vulnerabilities and input to the development of accident management. These Korean ALWR specific design features showed the containment performance is significantly enhanced compared with the other PWR plants

  4. Steady-state and loss-of-pumping accident analyses of the Savannah River new production reactor representative design

    International Nuclear Information System (INIS)

    Pryor, R.J.; Maloney, K.J.

    1990-10-01

    This document contains the steady-state and loss-of-pumping accident analysis of the representative design for the Savannah River heavy water new production reactor. A description of the reactor system and computer input model, the results of the steady-state analysis, and the results of four loss-of-pumping accident calculations are presented. 5 refs., 37 figs., 4 tabs

  5. Concept, design and equipment of a center for the treatment of radiation accidents at the Staedtisches Krankenhaus Muenchen-Schwabing

    International Nuclear Information System (INIS)

    Bogner, L.; Muehle, P.; Czempiel, H.; Henftling, H.G.

    1987-01-01

    The concept of the treatment center for radiation accidents at the Staedtisches Krankenhaus Muenchen-Schwabing is presented by means of a flow scheme for the treatment of different possible accidents. The resulting design and equipment are discussed in detail. (orig.) [de

  6. Use of probabilistic safety assessment in structuring conceptual design of accident mitigation systems

    Energy Technology Data Exchange (ETDEWEB)

    Nishiura, Hiroshi; Urata, Shigeru; Tsujikura, Yonezo [Kansai Electric Power Co., Inc., Osaka (Japan); Kuroiwa, Katsuya; Fujimoto, Haruo

    2000-07-01

    When there is an opportunity to develop a new safety design, it should be a rational design that serves its intended purpose while giving due consideration to factors such as reliability, economic efficiency, and others. Therefore, we have aimed to establish a methodical conceptual design process for accident mitigation systems as part of the core cooling system. In this consideration, we have proposed a process made up of 4 steps and have confirmed that the PSA method can be used as a tool in this process. (author)

  7. Use of probabilistic safety assessment in structuring conceptual design of accident mitigation systems

    International Nuclear Information System (INIS)

    Nishiura, Hiroshi; Urata, Shigeru; Tsujikura, Yonezo; Kuroiwa, Katsuya; Fujimoto, Haruo

    2000-01-01

    When there is an opportunity to develop a new safety design, it should be a rational design that serves its intended purpose while giving due consideration to factors such as reliability, economic efficiency, and others. Therefore, we have aimed to establish a methodical conceptual design process for accident mitigation systems as part of the core cooling system. In this consideration, we have proposed a process made up of 4 steps and have confirmed that the PSA method can be used as a tool in this process. (author)

  8. Design and Development of Black Box for Analyzing Accidents in Indian Railways

    Directory of Open Access Journals (Sweden)

    Alka DUBEY

    2010-03-01

    Full Text Available Black box also known as engine data recorder (EDR is a device for recording and analyzing train engine status for generation a report and analyzing train accidents. EDR is an application of embedded systems based on sensors, microcontroller, memory, serial interface and display unit. In the present paper designing of EDR with their technical specifications is provided. There railway electric engine WAP-7 is used in present study.

  9. Determination of a Basis for Design of a Yam (Dioscorea Spp ...

    African Journals Online (AJOL)

    Manual separation is both tedious and expensive, so the work reported here was done to determine a suitable basis for the design of a mechanical minisett sorter. Results from this study showed that the minisetts cut from the regions of the parent tuber can be separated on the basis of characteristic dimensions of arc length ...

  10. Review of design criteria for Criticality Accident Alarm System (CAAS) used in Fuel Reprocessing Facility

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Basu, Pew; Sivasubramaniyan, K.; Venkatraman, B.

    2016-01-01

    Though fuel cycle facilities handling fissile materials are designed with careful criticality safety analysis, the criticality accident cannot be ruled out completely. Criticality Accident Alarm System (CAAS) is being installed as part of criticality safety management in fuel cycle facilities. CAAS system being used in India, is ECIL make, ionization chamber based gamma detector, which houses three identical detectors and works on 2/3 logic. As per ISO 7753 and ANSI/ANS-8.3, the CAAS must be designed to be capable of detecting any minimum accident occurs which could be of concern. Based on this, alarm limit used in CAAS is: 4 R/h (fast transient excursion) and 3 mR in 0.5 sec (slow excursion). In case of reprocessing facilities wherein process tanks located in heavy shielding, identification of CAAS installation locations require detailed radiation transport calculations. A study has been taken to estimate the gamma dose rate from thick concrete hot cells in order to determine the locations of CAAS to meet the present design criteria of alarm limit

  11. Discussion on several issues of the accidents management of nuclear power plants in operation

    International Nuclear Information System (INIS)

    Cao Xuewu; Wang Zhe; Zhang Yingzhen

    2009-01-01

    This article discusses several issues of the accident management of nuclear power plants in operation, for example: the necessity, implementation principle of accident management and accident management program etc. For conducting accident management for beyond design basis accidents, this article thinks that the accident management program should be developed and implemented to ensure that the plant and its personnel with responsibilities for accident management are adequately prepared to take effective on-site actions to prevent or mitigate the consequences of severe accident. (authors)

  12. Sizing of type B package tie-downs on the basis of criteria related to hypothetical road transport accident conditions

    International Nuclear Information System (INIS)

    Phalippou, C.

    1986-01-01

    The aim is to guarantee intactness of the type B package containment system under hypothetical road accident conditions. Some experiments performed in France have led to analytical studies taking into account: a) the head-on collision, which is modelised by a uniform deceleration of 35 g, b) the side-on collision, which is modelised by a colliding object 3 times heavier than the package and an impact at 31.9 km/h. In the first case, the adopted criterion is the holding of the package on the vehicle by the strenght of the stowing members (tie-downs and chocks). In the second case, the adopted criterion is the desired breaking of the tie-downs in order to undamage package containment system; therefore it is assumed that no chock is acting against lateral impacts. Analytical and abacus methods have been developed for sizing the strenght of the stowing members in respect with the two above criteria [fr

  13. An Improved Setpoint Determination Methodology for the Plant Protection System Considering Beyond Design Basis Events

    International Nuclear Information System (INIS)

    Lee, C.J.; Baik, K.I.; Baek, S.M.; Park, K.-M.; Lee, S.J.

    2013-06-01

    According to the nuclear regulations and industry standards, the trip setpoint and allowable value for the plant protection system have been determined by considering design basis events. In order to improve the safety of a nuclear power plant, an attempt has been made to develop an improved setpoint determination methodology for the plant protection system trip parameter considering not only a design basis event but also a beyond design basis event. The results of a quantitative evaluation performed for the Advanced Power Reactor 1400 nuclear power plant in Korea are presented herein. The results confirmed that the proposed methodology is able to improve the nuclear power plant's safety by determining more reasonable setpoints that can cover beyond design basis events. (authors)

  14. Site selection and design basis of the National Disposal Facility for LILW. Geological and engineering barriers

    International Nuclear Information System (INIS)

    Boyanov, S.

    2010-01-01

    Content of the presentation: Site selection; Characteristics of the “Radiana” site (location, geological structure, physical and mechanical properties, hydro-geological conditions); Design basis of the Disposal Facility; Migration analysis; Safety assessment approach

  15. Projects of Modifications of design for mitigation of accidents outside the design Bases on nuclear Central PWR Siemens-KWU and Westinghouse; Proyectos de Modificaciones de Sieno para Mitigacion de Accidentes fuera de la Bases de Diseno en Centrales Nucleares PWR Siemens-KWU y Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez Gonzalez, G.; Cano Rodriguez, L. A.; Arguello Tara, A.

    2014-07-01

    Following the accident at the Japanese Fukushima-Daiichi NPP, the different regulators of nuclear power generation have required numerous reports regarding the evaluation and modification of the capacity of the plants to face accidents with severities beyond that established in their Design Bases. Under this new scenario, with multiple new demands and commitments, EA has carried out the required works for the implementation of strategies to mitigate the consequences of beyond Design Basis accidents for utilities owning Siemens-KWU and Westinghouse PWR nuclear power plants. (Author)

  16. The Mixed Waste Management Facility. Design basis integrated operations plan (Title I design)

    International Nuclear Information System (INIS)

    1994-12-01

    The Mixed Waste Management Facility (MWMF) will be a fully integrated, pilotscale facility for the demonstration of low-level, organic-matrix mixed waste treatment technologies. It will provide the bridge from bench-scale demonstrated technologies to the deployment and operation of full-scale treatment facilities. The MWMF is a key element in reducing the risk in deployment of effective and environmentally acceptable treatment processes for organic mixed-waste streams. The MWMF will provide the engineering test data, formal evaluation, and operating experience that will be required for these demonstration systems to become accepted by EPA and deployable in waste treatment facilities. The deployment will also demonstrate how to approach the permitting process with the regulatory agencies and how to operate and maintain the processes in a safe manner. This document describes, at a high level, how the facility will be designed and operated to achieve this mission. It frequently refers the reader to additional documentation that provides more detail in specific areas. Effective evaluation of a technology consists of a variety of informal and formal demonstrations involving individual technology systems or subsystems, integrated technology system combinations, or complete integrated treatment trains. Informal demonstrations will typically be used to gather general operating information and to establish a basis for development of formal demonstration plans. Formal demonstrations consist of a specific series of tests that are used to rigorously demonstrate the operation or performance of a specific system configuration

  17. Models of hemapoietic changes on the basis of systematically collected case histories of radiation accident victims as well as pathophysiologically evaluated patients after chronic radiation exposure

    International Nuclear Information System (INIS)

    Fliedner, T.M.

    2004-01-01

    The research project ''Models of Hematopoietic Changes on the Basis of Systematically Collected Case Histories of Radiation Accident Victims as well as Pathophysiologically Evaluated Patients after Chronic Radiation Exposure'' required the investigation of four major research problem areas. First of all, biomathematical models were improved or newly developed allowing the simulation of the radiation induced response patterns of granulocytes, lymphocytes and blood platelets. The compartment model approach allowed the establishment of the correlation of such blood cell changes to the extent of damage at the level of hemopoietic stem cells distributed throughout the skeleton. The utilization of neural-network techniques resulted in a ''synergetic'' model that enables the medical doctor - using blood cell changes within the first 5-6 days after exposure - to predict the further course of illness and to allow a rational approach to clinical management. Secondly, available information on the clinical consequences of radiation exposure on more than 800 accident victims enabled the team to develop an entirely new concept to recognize and treat such persons. For this approach the biomathematical models were used to identify ''response categories'' (rather than dosimetrically defined ''exposure categories'') with an organ specific grading code of the severity of radiation-induced damage. This grading allowed the semi-quantitative damage assessment of the hemopoiesis, the neurovascular system, the gastrointestinal as well as the cutaneous system. It forms the basis for a ''weighted'' prognosis and for the logistics of radiation accident medical management. In the third project domain, models were developed to understand pathophysiological mechanisms of biological consequences of chronic radiation in human beings (former USSR) as well as in a preclinical dog study (USA). From a large group of patients with the diagnosis of ''Chronic Radiation Sickness'' more than 80 were

  18. Development of Draft Regulatory Guide on Accident Analysis for Nuclear Power Plants with New Safety Design Features

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Woo, Sweng Woong; Hwang, Tae Suk [KINS, Daejeon (Korea, Republic of); Sim, Suk K; Hwang, Min Jeong [Environment and Energy Technology, Daejeon (Korea, Republic of)

    2016-05-15

    The present paper discusses the development process of the draft version of regulatory guide (DRG) on accident analysis of the NPP having the NSFD and its result. Based on the consideration on the lesson learned from the previous licensing review, a draft regulatory guide (DRG) on accident analysis for NPP with new safety design features (NSDF) was developed. New safety design features (NSDF) have been introduced to the new constructing nuclear power plants (NPP) since the early 2000 and the issuance of construction permit of SKN Units 3 and 4. Typical examples of the new safety features includes Fluidic Device (FD) within Safety Injection Tanks (SIT), Passive Auxiliary Feedwater System (PAFS), ECCS Core Barrel Duct (ECBD) which were adopted in APR1400 design and/or APR+ design to improve the safety margin of the plants for the postulated accidents of interest. Also several studies of new concept of the safety system such as Hybrid ECCS design have been reported. General and/or specific guideline of accident analysis considering the NSDF has been requested. Realistic evaluation of the impact of NSDF on accident with uncertainty and separated accident analysis accounting the NSDF impact were specified in the DRG. Per the developmental process, identification of key issues, demonstration of the DRG with specific accident with specific NSDF, and improvement of DGR for the key issues and their resolution will be conducted.

  19. Verification of fire and explosion accident analysis codes (facility design and preliminary results)

    International Nuclear Information System (INIS)

    Gregory, W.S.; Nichols, B.D.; Talbott, D.V.; Smith, P.R.; Fenton, D.L.

    1985-01-01

    For several years, the US Nuclear Regulatory Commission has sponsored the development of methods for improving capabilities to analyze the effects of postulated accidents in nuclear facilities; the accidents of interest are those that could occur during nuclear materials handling. At the Los Alamos National Laboratory, this program has resulted in three computer codes: FIRAC, EXPAC, and TORAC. These codes are designed to predict the effects of fires, explosions, and tornadoes in nuclear facilities. Particular emphasis is placed on the movement of airborne radioactive material through the gaseous effluent treatment system of a nuclear installation. The design, construction, and calibration of an experimental ventilation system to verify the fire and explosion accident analysis codes are described. The facility features a large industrial heater and several aerosol smoke generators that are used to simulate fires. Both injected thermal energy and aerosol mass can be controlled using this equipment. Explosions are simulated with H 2 /O 2 balloons and small explosive charges. Experimental measurements of temperature, energy, aerosol release rates, smoke concentration, and mass accumulation on HEPA filters can be made. Volumetric flow rate and differential pressures also are monitored. The initial experiments involve varying parameters such as thermal and aerosol rate and ventilation flow rate. FIRAC prediction results are presented. 10 figs

  20. Nuclear accident dosimeter designed for use with the Panasonic TLD system

    International Nuclear Information System (INIS)

    Hankins, D.E.

    1985-01-01

    A new design for the nuclear accident dosimeter (NAD) compatible with the Panasonic TLD badge has recently been adopted for use at LLNL. This NAD was tested at the 1984 Oak Ridge National Laboratory Intercomparison of Criticality Accident Dosimeters study. We describe the procedures and constants developed to evaluate the NAD components. These constants were averaged to give reasonable results from bare and moderated spectra. Other procedures to evaluate a person's neutron dose using activation of the blood sodium and hair are described. These latter procedures are used to complement the dose determined using the NAD, or to determine a dose if a NAD had not been worn during exposure. If little is known about the configuration of the fissile material or shielding between the material and the exposed person, a procedure which combines the blood and hair activations gives a good estimate of the dose. (DT) 3 refs., 2 figs., 6 tabs

  1. LMFBR post accident heat removal testing needs and conceptual design of a test facility

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

    1977-03-01

    A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

  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. MDEP AP1000WG Design-Specific Common Position CP-AP1000WG-02. Common position addressing Fukushima Daiichi NPP accident-related issues

    International Nuclear Information System (INIS)

    2016-09-01

    A severe accident involving several units took place in Japan at Fukushima Daiichi nuclear power plant (NPP) in March 2011. The immediate cause of the accident was an earthquake followed by a tsunami coupled with inadequate provisions against the consequences of such events in the design. Opportunities to improve protection against a realistic design basis tsunami had not been taken. As a consequence of the tsunami, safety equipment and the related safety functions were lost at the plant, leading to core damage in three units and subsequently to large radioactive release. Several studies have already been performed to better understand the accident progression and detailed technical studies are still in progress in Japan and elsewhere. In the meantime, on-going studies on the behaviour of nuclear power plants in very severe situations, similar to Fukushima Daiichi, seek to identify potential vulnerabilities in plant design and operation; to suggest reasonably practicable upgrades; or to recommend enhanced regulatory requirements and guidance to address such situations. Likewise, agencies around the world that are responsible for regulating the design, construction and operation of AP1000 R plants are engaged in similar activities. The MDEP AP1000 R Working Group (AP1000 WG) members consist of members from Canada, China, the United Kingdom and the United States. Since the regulatory review of their AP1000 R applications have not been completed by all of these Countries yet, this paper identifies common preliminary approaches to address potential safety improvements for AP1000 R plants as related to lessons learned from the Fukushima Daiichi accident or Fukushima Daiichi-related issues. In seeking common position, regulators will provide input to this paper to reflect their safety conclusions regarding the AP1000 R design and how the design could be enhanced to address Fukushima Daiichi issues. The common preliminary approaches are organized into five sections

  4. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S. E-mail: reyessuarezl@llnl.gov; Latkowski, J.F.; Gomez del Rio, J.; Sanz, J

    2001-05-21

    Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  5. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Science.gov (United States)

    Reyes, S.; Latkowski, J. F.; Gomez del Rio, J.; Sanz, J.

    2001-05-01

    Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  6. Design measures for prevention and mitigation of severe accidents at advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-06-01

    Over 8500 reactor-years of operating experience have been accumulated with the current nuclear energy systems. New generations of nuclear power plants are being developed, building upon this background of experience. During the last decade, requirements for equipment specifically intended to minimize releases of radioactive material to the environment in the event of a core melt accident have been introduced, and designs for new plants include measures for preventing and mitigating a range of severe accident scenarios. The IAEA Technical Committee Meeting on Impact of Severe Accidents on Plant Design and Layout of Advanced Water Cooled Reactors was jointly organized by the Department of Nuclear Energy and the Department of Nuclear Safety to review measures which are being incorporated into advanced water cooled reactor designs for preventing and mitigating severe accidents, the status of experimental and analytical investigations of severe accident phenomena and challenges which support design decisions and accident management procedures, and to understand the impact of explicitly addressing severe accidents on the cost of nuclear power plants. This publication is intended to provide an objective source of information on this topic. It includes 14 papers presented at the Technical Committee meeting held in Vienna between 21-25 October 1996. It also includes a Summary and Findings of the Working Groups. The papers were grouped in three sections. A separate abstract was prepared for each paper

  7. System Design Strategies of Post-Accident Monitoring System for a PGSFR in Korea

    International Nuclear Information System (INIS)

    Jang, Gwi-sook; Jeong, Kwang-il; Keum, Jong-yong; Seong, Seung-hwan

    2013-06-01

    Monitoring systems of a PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) in Korea provide alarms, integrity information in the reactor building, sodium-water reaction information in the steam generator, fuel failure information, and supporting information for maintenance and inspection. In particular, a Post-Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. Some PAM variables can be allocated as more two types. It is important for system designers to confirm the suitability of the selection of PAM variables. In addition, the PAMS is a position 4 display against common cause failures of safety I and C systems. The position 4 display should be independent and diverse from the safety I and C systems. The diversity of safety I and C equipment has led to an increase in the design and verification and validation cost. Thus, this paper proposes the system design strategies on the PAMS design problems of the PGSFR in KOREA. The results will be input into a conceptual system design for the PAMS of the PGSFR in KOREA. (authors)

  8. Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

    2010-01-01

    To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

  9. Licensing aspects in the verification of the SNR 300 design concept against hypothetical accidents

    International Nuclear Information System (INIS)

    Kugler, E.; Wiesner, S.

    1976-01-01

    The German prototype of a fast breeder reactor, the SNR 300, is being built near Kalkar on the Lower Rhine. It is a loop-type fast sodium-cooled reactor, designed and constructed by Interatom, Bensberg. Experiences gained from the first phase of construction are described. The report is restricted to the aspects of the SNR 300 design against a core disruptive accident (CDA) and its consequences and to the difficulties having arisen in the verification of the design concept so far. Some examples of the detailed design are described and discussed from the licensing authority's point of view showing that the difficulties have been typical for a prototype reactor subjected to a regular licensing procedure

  10. Neural network of Gaussian radial basis functions applied to the problem of identification of nuclear accidents in a PWR nuclear power plant

    International Nuclear Information System (INIS)

    Gomes, Carla Regina; Canedo Medeiros, Jose Antonio Carlos

    2015-01-01

    Highlights: • It is presented a new method based on Artificial Neural Network (ANN) developed to deal with accident identification in PWR nuclear power plants. • Obtained results have shown the efficiency of the referred technique. • Results obtained with this method are as good as or even better to similar optimization tools available in the literature. - Abstract: The task of monitoring a nuclear power plant consists on determining, continuously and in real time, the state of the plant’s systems in such a way to give indications of abnormalities to the operators and enable them to recognize anomalies in system behavior. The monitoring is based on readings of a large number of meters and alarm indicators which are located in the main control room of the facility. On the occurrence of a transient or of an accident on the nuclear power plant, even the most experienced operators can be confronted with conflicting indications due to the interactions between the various components of the plant systems; since a disturbance of a system can cause disturbances on another plant system, thus the operator may not be able to distinguish what is cause and what is the effect. This cognitive overload, to which operators are submitted, causes a difficulty in understanding clearly the indication of an abnormality in its initial phase of development and in taking the appropriate and immediate corrective actions to face the system failure. With this in mind, computerized monitoring systems based on artificial intelligence that could help the operators to detect and diagnose these failures have been devised and have been the subject of research. Among the techniques that can be used in such development, radial basis functions (RBFs) neural networks play an important role due to the fact that they are able to provide good approximations to functions of a finite number of real variables. This paper aims to present an application of a neural network of Gaussian radial basis

  11. The design of PSB-VVER experiments relevant to accident management

    International Nuclear Information System (INIS)

    Del Nevo, Alessandro; D'auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    2008-01-01

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes, which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility, operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed. The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility. (author)

  12. The Design of PSB-VVER Experiments Relevant to Accident Management

    Science.gov (United States)

    Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

  13. Radiation accidents

    International Nuclear Information System (INIS)

    Poplavskij, K.K.; Smorodintseva, G.I.

    1978-01-01

    On the basis of a critical analysis of the available data on causes and consequences of radiation accidents (RA), a classification of RA by severity (five groups of accidents) according to biomedical consequences and categories of exposed personnel is proposed. A RA is defined and its main characteristics are described. Methods of RA prevention are proposed, as is a plan of specific measures to deal with RA in accordance with the proposed classification

  14. Evaluation of seismic design by students made after Fukushima Dai-ichi accident

    International Nuclear Information System (INIS)

    Sugiyama, Ken-ichiro

    2012-01-01

    The sense of anxiety for safety of nuclear power plants among people in Japan has not disappeared after Fukushima Dai-ichi accident because of a typical country with frequent earthquakes. The provision of information for seismic design in nuclear power plants prepared for easier comprehension is always required in any kind of study meetings for the social acceptance of nuclear power plants. In the present paper, the effect of the provision of information made an attempt for students in Hokkaido University is reported. (author)

  15. Assessment of value-impact associated with the elimination of postulated pipe ruptures from the design basis for nuclear power plants

    International Nuclear Information System (INIS)

    Holman, G.S.; Chou, C.K.

    1985-01-01

    The US Nuclear Regulatory Commission is proposing to amend the regulations that currently require that the design basis for nuclear power plants include the postulation of dynamic effects from loss of coolant accidents up to and including the double-ended rupture of the largest pipe in the reactor coolant system. Proposed modifications would allow analyses to serve as a sufficient basis for excluding dynamic effects, including but not necessarily limited to pipe whip and jet impingement, associated with specific pipe ruptures. Only dynamic effects would be impacted; current design requirements for containment sizing and discharge capacity of emergency core cooling systems would remain unchanged. This report presents a detailed analysis of value-impact associated with the proposed amendment for PWR reactor coolant loop piping and for BWR recirculation loop piping. The effect of extending application of the proposed rule change to other piping systems is also assessed in a less quantitative manner

  16. Defense-in-depth approach against a beyond design basis event

    Energy Technology Data Exchange (ETDEWEB)

    Hoang, H., E-mail: Hoa.hoang@ge.com [GE Hitachi Nuclear Energy, 1989 Little Orchard St., 95125 San Jose, California (United States)

    2013-10-15

    The US industry, with the approval of the Nuclear Regulatory Commission, is promoting an approach to add diverse and flexible mitigation strategies, or Flex, that will increase the defense-in-depth capability for the nuclear power plants in the event of beyond design basis event, such as at the Fukushima Dai-ichi station. The objective of Flex is to establish and indefinite coping capability to prevent damage to the fuel in the core and spent fuel pool, and to maintain the containment function by utilizing installed equipment, on-site portable equipment and pre-staged off-site resources. This capability will address both an extended loss of all Ac power and a loss of ultimate heat sink which could arise following a design basis event with additional failures, and conditions from a beyond design basis event. (author)

  17. Defense-in-depth approach against a beyond design basis event

    International Nuclear Information System (INIS)

    Hoang, H.

    2013-10-01

    The US industry, with the approval of the Nuclear Regulatory Commission, is promoting an approach to add diverse and flexible mitigation strategies, or Flex, that will increase the defense-in-depth capability for the nuclear power plants in the event of beyond design basis event, such as at the Fukushima Dai-ichi station. The objective of Flex is to establish and indefinite coping capability to prevent damage to the fuel in the core and spent fuel pool, and to maintain the containment function by utilizing installed equipment, on-site portable equipment and pre-staged off-site resources. This capability will address both an extended loss of all Ac power and a loss of ultimate heat sink which could arise following a design basis event with additional failures, and conditions from a beyond design basis event. (author)

  18. Guide to radiological accident considerations for siting and design of DOE nonreactor nuclear facilities

    International Nuclear Information System (INIS)

    Elder, J.; Graf, J.M.

    1984-01-01

    DOE Office of Nuclear Safety has sponsored preparation of a guidance document to aid field offices and contractors in their analyses of consequences of postulated major accidents. The guide addresses the requirements of DOE Orders 5480.1A, Chapter V, and 6430.1, including the general requirement that DOE nuclear facilities be sited, designed, and operated in accordance with standards, codes, and guides consistent with those applied to comparable licensed nuclear facilities. The guide includes both philosophical and technical information in the areas of: siting guidelines doses applied to an offsite reference person; consideration also given to an onsite reference person; physical parameters, models, and assumptions to be applied when calculating doses for comparison to siting criteria; and potential accident consequences other than radiological dose to a reference person which might affect siting and major design features of the facility, such as environmental contamination, population dose, and associated public health effects. Recommendations and/or clarifications are provided where this could be done without adding new requirements. In this regard, the guide is considered a valuable aid to the safety analyst, especially where requirements have been subject to inconsistent interpretation or where analysis methods are in transition, such as use of dose model (ICRP 2 or ICRP 30) or use of probabilistic methods of risk analysis in the siting and design of nuclear facilities

  19. 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

  20. Design Basis Threat (DBT) Approach for the First NPP Security System in Indonesia

    International Nuclear Information System (INIS)

    Ign Djoko Irianto

    2004-01-01

    Design Basis Threat (DBT) is one of the main factors to be taken into account in the design of physical protection system of nuclear facility. In accordance with IAEA's recommendations outlined in INFCIRC/225/Rev.4 (Corrected), DBT is defined as: attributes and characteristics of potential insider and/or external adversaries, who might attempt unauthorized removal of nuclear material or sabotage against the nuclear facilities. There are three types of adversary that must be considered in DBT, such as adversary who comes from the outside (external adversary), adversary who comes from the inside (internal adversary), and adversary who comes from outside and colludes with insiders. Current situation in Indonesia, where many bomb attacks occurred, requires serious attention on DBT in the physical protection design of NPP which is to be built in Indonesia. This paper is intended to describe the methodology on how to create and implement a Design Basis Threat in the design process of NPP physical protection in Indonesia. (author)

  1. Regulatory standpoints on the design-basis capability of safety-related motor-operated valves(MOVs) and power-operated gate valves(POGVs)

    International Nuclear Information System (INIS)

    Kim, W. T.; Kum, O. H.

    1999-01-01

    The weakness in the design-basis capability of Motor-Operated Valves(MOVs) and the susceptibility to Pressure Locking and Thermal Binding phenomena of Power-Operated Gate Valves(POGVs) have been major concerns to be resolved in the nuclear society in and abroad since Three Mile Island accident occurred in the USA in 1979. Through detailed analysis of operating experience and regulatory activities, some MOVs and POGVs have been found to be unreliable in performing their safety functions when they are required to do so under certain conditions, especially under design-basis accident conditions. Further, it is well understood that these safety problems may not be identified by the typical valve in-service testing(IST). USNRC has published three Generic Letters, GL 89-10, GL 95-07, and GL 96-05, requiring nuclear plant licensees to take appropriate actions to resolve the problems mentioned above. Korean nuclear regulatory body has made public an administration measure called 'Regulatory recommendation to verify safety functions of the safety-related MOVs and POGVs' on June 13, 1997, and in this administration measure Korean utility is asked to submit written documents to show how it assure design-basis capability of these valves. The following are among the major concerns being considered from a regulation standpoint. Program scope and implementation priority, dynamic tests under differential pressure conditions, accuracy of diagnostic equipment, torque switch setting and torque bypass percentage, weak link analysis, motor actuator sizing, corrective actions taken to resolve pressure locking and thermal binding susceptibility, and a periodic verification program for the valves once design-basis capability has been verified

  2. EMERALD, Radiation Release and Dose after PWR Accident for Design Analysis and Operation Analysis

    International Nuclear Information System (INIS)

    Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

    1988-01-01

    1 - Description of problem or function: The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place in the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies. 2 - Method of solution - Explicit solutions of first-order linear differential equations are included. In addition, a subroutine is provided which solves a set of simultaneous linear algebraic equations. 3 - Restrictions on the complexity of the problem - Maxima of: 25 isotopes, 7 time periods, 15 volumes or components, 10

  3. Flexural behavior and design of steel-plate composite (SC) walls for accident thermal loading

    Energy Technology Data Exchange (ETDEWEB)

    Booth, Peter N., E-mail: boothpn@purdue.edu [Lyles School of Civil Engineering, Purdue University, West Lafayette, IN (United States); Varma, Amit H., E-mail: ahvarma@purdue.edu [Lyles School of Civil Engineering, Purdue University, West Lafayette, IN (United States); Sener, Kadir C., E-mail: ksener@purdue.edu [Lyles School of Civil Engineering, Purdue University, West Lafayette, IN (United States); Malushte, Sanjeev R. [Bechtel Corp., Frederick, MD (United States)

    2015-12-15

    Modular steel-plate composite (SC) safety-related nuclear power plant structures must be designed to resist accident thermal and mechanical loads. The design accident thermal load represents the condition where high pressure and temperature steam is released as result of a mechanical failure and applied against the surfaces of power plant structural walls. The effect of heating and pressure can have both short and long term effects on the mechanical integrity of SC structures including degradation and cracking of concrete infill, residual stresses, and out-of-plane deformations. The purpose of this research is to study the effects of thermal and mechanical loads on the out-of-plane flexural response of SC walls and to develop simplified equations that can be used to predict behavior. Four experimental beam tests are reported that represent full-scale cross-sections of SC walls subjected to combinations of mechanical and thermal loads. The study determined that thermal loads reduce the out-of-plane flexural stiffness of SC walls. For the ambient condition, the flexural stiffness closely matches a conventional elastic cracked-transformed model, and at elevated temperatures, the stiffness is reduced to a fully-cracked flexural stiffness that only takes into account the stiffness of the steel faceplates. A method is presented for estimating the thermal curvature, ϕ{sub th}, and thermal moment, M{sub th}, resulting from unequal heating of opposing faces of an SC wall. Based on the tests in this study, the application of accident thermal loads did not result in a reduction of the flexural strength of the SC section.

  4. Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, Rajiv; Hibler, Lyle F.; Coleman, Andre M.; Ward, Duane L.

    2011-11-01

    The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.

  5. Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America

    International Nuclear Information System (INIS)

    Prasad, Rajiv; Hibler, Lyle F.; Coleman, Andre M.; Ward, Duane L.

    2011-01-01

    The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.

  6. Study on effective prestressing effects on concrete containment under the design-basis pressure condition

    International Nuclear Information System (INIS)

    Sun Feng; Pan Rong; Wang Lu; Mao Huan; Yang Yu

    2013-01-01

    Prestressing technology is widely used in nuclear power plant containment building, and the durability of containment structure is affected directly by the distribution and loss of prestressing value under design-basis pressure. Containment structure and the distribution of prestressing system are introduced briefly. Furthermore, the calculating process of horizontal prestressing bunch loss near the equipment hatch hole is put forward in details, and the containment structure prestressing loss when 5-year pressure test is obtained. Based above analysis, the finite element model of the prestressed concrete containment structure is built by using ANSYS code, the prestressing effect on concrete containment is analysed. The results show that most of the design pressure is bore by the prestressing system under the design-basis pressure, so the containment structure is safe. These conclusions are consistent with prestressing containment system design concepts, which can provide reference to the engineering staff. (authors)

  7. Accident management insights after the Fukushima Daiichi NPP accident

    International Nuclear Information System (INIS)

    Degueldre, Didier; Viktorov, Alexandre; Tuomainen, Minna; Ducamp, Francois; Chevalier, Sophie; Guigueno, Yves; Tasset, Daniel; Heinrich, Marcus; Schneider, Matthias; Funahashi, Toshihiro; Hotta, Akitoshi; Kajimoto, Mitsuhiro; Chung, Dae-Wook; Kuriene, Laima; Kozlova, Nadezhda; Zivko, Tomi; Aleza, Santiago; Jones, John; McHale, Jack; Nieh, Ho; Pascal, Ghislain; ); Nakoski, John; Neretin, Victor; Nezuka, Takayoshi; )

    2014-01-01

    The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

  8. Reconfigurable Flight Control Design using a Robust Servo LQR and Radial Basis Function Neural Networks

    Science.gov (United States)

    Burken, John J.

    2005-01-01

    This viewgraph presentation reviews the use of a Robust Servo Linear Quadratic Regulator (LQR) and a Radial Basis Function (RBF) Neural Network in reconfigurable flight control designs in adaptation to a aircraft part failure. The method uses a robust LQR servomechanism design with model Reference adaptive control, and RBF neural networks. During the failure the LQR servomechanism behaved well, and using the neural networks improved the tracking.

  9. Chemical data for the calculation of fission product releases in design basis faults in PWRs

    International Nuclear Information System (INIS)

    Ali, S.M.; Bawden, R.J.; Garbett, K.; Deane, A.M.; Large, N.R.

    1982-04-01

    This review considers the chemistry of caesium and iodine and their volatility under the conditions which would exist during a number of design-basis faults. It recommends values which should be used for the distribution of these elements between liquid and gas phases. (author)

  10. Accident selection methodology for TA-55 FSAR

    International Nuclear Information System (INIS)

    Letellier, B.C.; Pan, P.Y.; Sasser, M.K.

    1995-01-01

    In the past, the selection of representative accidents for refined analysis from the numerous scenarios identified in hazards analyses (HAs) has involved significant judgment and has been difficult to defend. As part of upgrading the Final Safety Analysis Report (FSAR) for the TA-55 plutonium facility at the Los Alamos National Laboratory, an accident selection process was developed that is mostly mechanical and reproducible in nature and fulfills the requirements of the Department of Energy (DOE) Standard 3009 and DOE Order 5480.23. Among the objectives specified by this guidance are the requirements that accident screening (1) consider accidents during normal and abnormal operating conditions, (2) consider both design basis and beyond design basis accidents, (3) characterize accidents by category (operational, natural phenomena, etc.) and by type (spill, explosion, fire, etc.), and (4) identify accidents that bound all foreseeable accident types. The accident selection process described here in the context of the TA-55 FSAR is applicable to all types of DOE facilities

  11. Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

    International Nuclear Information System (INIS)

    Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

    1995-01-01

    The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

  12. NDARC-NASA Design and Analysis of Rotorcraft Theoretical Basis and Architecture

    Science.gov (United States)

    Johnson, Wayne

    2010-01-01

    The theoretical basis and architecture of the conceptual design tool NDARC (NASA Design and Analysis of Rotorcraft) are described. The principal tasks of NDARC are to design (or size) a rotorcraft to satisfy specified design conditions and missions, and then analyze the performance of the aircraft for a set of off-design missions and point operating conditions. The aircraft consists of a set of components, including fuselage, rotors, wings, tails, and propulsion. For each component, attributes such as performance, drag, and weight can be calculated. The aircraft attributes are obtained from the sum of the component attributes. NDARC provides a capability to model general rotorcraft configurations, and estimate the performance and attributes of advanced rotor concepts. The software has been implemented with low-fidelity models, typical of the conceptual design environment. Incorporation of higher-fidelity models will be possible, as the architecture of the code accommodates configuration flexibility, a hierarchy of models, and ultimately multidisciplinary design, analysis and optimization.

  13. Radial basis function (RBF) neural network control for mechanical systems design, analysis and Matlab simulation

    CERN Document Server

    Liu, Jinkun

    2013-01-01

    Radial Basis Function (RBF) Neural Network Control for Mechanical Systems is motivated by the need for systematic design approaches to stable adaptive control system design using neural network approximation-based techniques. The main objectives of the book are to introduce the concrete design methods and MATLAB simulation of stable adaptive RBF neural control strategies. In this book, a broad range of implementable neural network control design methods for mechanical systems are presented, such as robot manipulators, inverted pendulums, single link flexible joint robots, motors, etc. Advanced neural network controller design methods and their stability analysis are explored. The book provides readers with the fundamentals of neural network control system design.   This book is intended for the researchers in the fields of neural adaptive control, mechanical systems, Matlab simulation, engineering design, robotics and automation. Jinkun Liu is a professor at Beijing University of Aeronautics and Astronauti...

  14. Control room unfiltered in-leakage limit analysis of design-basis LOCA for Lungmen ABWR plant

    International Nuclear Information System (INIS)

    Tsai Chihming; Chang Chinjang; Yuann Yngruey

    2014-01-01

    In USNRC's Generic Letter 2003-01, 'Control Room Habitability,' it requests utilities provide information to demonstrate that the control room at each of their respective facilities complies with the current licensing and design bases, and applicable regulatory requirements, and that suitable design, maintenance and testing control measures are in place for maintaining this compliance. In particular, each utility is required to perform the control room in-leakage test to demonstrate that the unfiltered in-leakage rate is within that assumed in the licensing analyses. It must be ensured that the control room envelope habitability, in terms of radiation dose, is maintained during normal operations as well as design basis accidents. In view of this, a dose analysis has been performed to establish the control room unfiltered in-leakage limit which can be used as an acceptance criterion for the in-leakage test. The analysis in this study is for Lungmen ABWR plant. The plant has twin units, with each unit having its own control room. The TID-4844 source terms and associated methodology are used. The USNRC RADTRAD v3.03 code is employed for the transport calculation of radioactive materials in different paths, including control room in-leakage path. The radiological criterion on protection of the operators specified in 10 CFR 50, Appendix A, General Design Criterion 19 is followed. It's demonstrated that the performance of Lungmen control room with 500 cfm unfiltered in-leakage air could meet the radiological habitability acceptance criteria in case of radiation hazards. (author)

  15. The investigation of the impacts of major disasters, on the basis of the Van earthquake (October 23, 2011, Turkey), on the profile of the injuries due to occupational accidents.

    Science.gov (United States)

    Hekimoglu, Yavuz; Dursun, Recep; Karadas, Sevdegul; Asirdizer, Mahmut

    2015-10-01

    The purpose of this study is to identify the impacts of major disasters, on the basis of the Van earthquake (October 23, 2011, Turkey), on the profile of the injuries due to occupational accidents. In this study, we evaluated 245 patients of occupational accidents who were admitted to emergency services of Van city hospitals in the 1-year periods including pre-earthquake and post-earthquake. We determined that there was a 63.4% (P accidents in the post-earthquake period compared to the pre-earthquake period. Also, injuries due to occupational accidents increased 211% (P accidents. In this study, the impact of disasters such as earthquakes on the accidents at work was evaluated as we have not seen in literature. This study emphasizes that governments should make regulations and process relating to the post-disaster business before the emergence of disaster by taking into account factors that may increase their work-related accidents. Copyright © 2015 Elsevier Ltd and Faculty of Forensic and Legal Medicine. All rights reserved.

  16. 77 FR 64564 - Implementation of Regulatory Guide 1.221 on Design-Basis Hurricane and Hurricane Missiles

    Science.gov (United States)

    2012-10-22

    ...-Basis Hurricane and Hurricane Missiles AGENCY: Nuclear Regulatory Commission. ACTION: Proposed interim...-ISG-024, ``Implementation of Regulatory Guide 1.221 on Design-Basis Hurricane and Hurricane Missiles....221, ``Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants.'' DATES: Submit...

  17. Design Basis Provisions for New and Existing Nuclear Power Plants and Nuclear Fuel Cycle Facilities in India

    International Nuclear Information System (INIS)

    Soni, R.S.

    2013-01-01

    India has 3-Stage Nuclear Power Program. • Various facilities under design, construction or operation. • Design Basis Knowledge Management (DBKM) is an important and challenging task. • Design Basis Knowledge contributes towards: - Safe operation of running plants; - Design and construction of new facilities; - Addresses issues related to future decommissioning activities

  18. Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.

  19. TRAC analysis of design basis events for the accelerator production of tritium target/blanket

    International Nuclear Information System (INIS)

    Lin, J.C.; Elson, J.

    1997-01-01

    A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers

  20. Approach to accident management in RBMK-1500

    International Nuclear Information System (INIS)

    Kaliatka, A.; Urbonavicius, E.; Uspuras, E.

    2008-01-01

    In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK. Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed. This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account

  1. Consequence analysis of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    Wahba, N.N.; Kim, Y.T.; Lie, S.G.

    1997-01-01

    The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

  2. NPP Design Basis Handover and Knowledge Preservation from Subcontractors, Vendors and EPC

    International Nuclear Information System (INIS)

    Freeland, Kent

    2013-01-01

    Using PLM-based Workflow for Configuration Management (CM) in the Nuclear Power Industry Advantages – some work to do! • NPP’s must adapt to using PLM-based solutions to support CM and to synchronize design changes to asset or product changes, and reduce “slipstreaming”. In the NPP world, this often appears as events that circumvent CM – for example, non-approved parts substitutions and “temporary” plant modifications that are never removed. • PLM serves as the method for unifying the application of requirements to design changes, processes and workflow. In NPP’s, requirements are generally considered only relevant to designs – not process and workflow. • PLM supports Configuration Management and Design Basis in Regulator Action Tracking for NPP’s, and application of PLM-based CM to regulator action and compliance systems. This is a poorly-understood application of CM in NPP’s, yet these elements control large parts of the NPP design basis. • Suppliers, EPC’s and Technology Vendors must also understand the role of CM, SE and PLM in construction of new standards-driven NPP designs (like EPR and Westinghouse AP-1000 NPP designs), as well as understanding the role and handling of Knowledge Systems

  3. The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1986-01-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)

  4. Development of a design basis tornado and structural design criteria for Lawrence Livermore Laboratory's Site 300

    International Nuclear Information System (INIS)

    McDonald, J.R.; Minor, J.E.; Mehta, K.C.

    1975-11-01

    Criteria are prescribed and guidance is provided for professional personnel who are involved with the evaluation of existing buildings and facilities at Site 300 near Livermore, California to resist the possible effects of extreme winds and tornadoes. The development of parameters for the effects of tornadoes and extreme winds and guidelines for evaluation and design of structures are presented. The investigations conducted are summarized and the techniques used for arriving at the combined tornado and extreme wind risk model are discussed. The guidelines for structural design methods for calculating pressure distributions on walls and roofs of structures and methods for accommodating impact loads from missiles are also presented

  5. Calculation and experimental study of the RBMK-1500 reactor emergency cooling at maximum designed accident

    International Nuclear Information System (INIS)

    Cherkashov, Yu.M.; Vasilevskij, V.P.; Labazov, V.H.; Loninov, A.Ya.; Molochnikov, Yu.S.; Novosel'skij, O.Yu.; Podlazov, L.N.; Pavlov, V.B.; Pushkarev, V.I.

    1981-01-01

    The analysis of thermohydraulic and neutron-physical processes occurring in the RBMK-1500 reactor during the reactor emergency cooling system triggering (RECS) after the maximum designed accident (MDA) is conducted. The MDA means hypothetical instant hilliotine break of the main circulating pump head collector. During the whole cooling down period the RECS should provide the temperature level of the fuel elements not exceeding 1200 deg C and the channel pipe temperature - 600 deg C. The principal flowsheet of the balloon type RECS is described. Calculations of the valve fast response effect on the RECS productivity are carried out. It is concluded that the chosen balloon RECS provides reliable temperature modes of fuel elements naand channel pipes under the MDA conditions. At the same time a momentary splash of neutron power by the value not more than 10% can take place [ru

  6. Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

    International Nuclear Information System (INIS)

    Sevon, T.

    2005-11-01

    In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

  7. Development and comparision of techniques for estimating design basis flood flows for nuclear power plants

    International Nuclear Information System (INIS)

    1980-05-01

    Estimation of the design basis flood for Nuclear Power Plants can be carried out using either deterministic or stochastic techniques. Stochastic techniques, while widely used for the solution of a variety of hydrological and other problems, have not been used to date (1980) in connection with the estimation of design basis flood for NPP siting. This study compares the two techniques against one specific river site (Galt on the Grand River, Ontario). The study concludes that both techniques lead to comparable results , but that stochastic techniques have the advantage of extracting maximum information from available data and presenting the results (flood flow) as a continuous function of probability together with estimation of confidence limits. (author)

  8. Ground motion following selection of SRS design basis earthquake and associated deterministic approach

    International Nuclear Information System (INIS)

    1991-03-01

    This report summarizes the results of a deterministic assessment of earthquake ground motions at the Savannah River Site (SRS). The purpose of this study is to assist the Environmental Sciences Section of the Savannah River Laboratory in reevaluating the design basis earthquake (DBE) ground motion at SRS during approaches defined in Appendix A to 10 CFR Part 100. This work is in support of the Seismic Engineering Section's Seismic Qualification Program for reactor restart

  9. Interregional Knowledge Management Workshop on Life Cycle Management of Design Basis Information. Issues, Challenges, Approaches

    International Nuclear Information System (INIS)

    Šula, Radek

    2013-01-01

    Introduction and objectives: • It is evident that the design basis area is from the point of view of knowledge sharing extremely complicated. • Time is changing and puts on us ever greater demands. • We have to analyze the near and remote surroundings and have to simplified the problem of knowledge sharing in that area. • I believe that it is graspable task for knowledge management and I will try to outline some possible context and approaches

  10. The Swedish Utilities joint approach to form common basis for design requirements for the future

    International Nuclear Information System (INIS)

    Hansson, B.

    1998-01-01

    The Owners of the Swedish Nuclear Power Plants have decided to form a document that should state the design principals and requirement for cost-effective and continuous development of the reactor safety in the future. The development of this document will be a part of the modernization and development of the Swedish Nuclear Power Plants. The basis for this document is an evaluation of Swedish and International standards and regulations as IAEA/INSAG, US-regulations, EUR etc. (author)

  11. Applicability of simplified human reliability analysis methods for severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Boring, R.; St Germain, S. [Idaho National Lab., Idaho Falls, Idaho (United States); Banaseanu, G.; Chatri, H.; Akl, Y. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2016-03-15

    Most contemporary human reliability analysis (HRA) methods were created to analyse design-basis accidents at nuclear power plants. As part of a comprehensive expansion of risk assessments at many plants internationally, HRAs will begin considering severe accident scenarios. Severe accidents, while extremely rare, constitute high consequence events that significantly challenge successful operations and recovery. Challenges during severe accidents include degraded and hazardous operating conditions at the plant, the shift in control from the main control room to the technical support center, the unavailability of plant instrumentation, and the need to use different types of operating procedures. Such shifts in operations may also test key assumptions in existing HRA methods. This paper discusses key differences between design basis and severe accidents, reviews efforts to date to create customized HRA methods suitable for severe accidents, and recommends practices for adapting existing HRA methods that are already being used for HRAs at the plants. (author)

  12. Hydrogen-management in beyond design accident conditions in NPP Neckar 2

    International Nuclear Information System (INIS)

    Zaiss, W.

    1999-01-01

    Neckar 2 is a 1340 MWE 4-loop pressurized water reactor (PWR) of Siemens KONVOI type, located in the south of Germany. It was first connected to the grid in January 1989. Commercial operation started in April 1989. Task assignment: In Germany it was recommended by the Reactor Safety Commission (RSK) on December 17, 1997, to reequip passive autocatalytic recombiners for the controlling of the hydrogen problem. The removal of the hydrogen is an essential part which guarantees the integrity of the containment. The implementation of the recombiners is a further step for the decrease of the nuclear rest risk. The RSK confirmed, that the implementation of the passive autocatalytic recombiners is a safety measure for the controlled removal of the hydrogen in beyond design accident conditions. Assumption : Failure of the whole residual heat removal system (RHRS) and non sufficient effect of the systems which have been installed for beyond design accident conditions. Effect on the reactor coolant system (RCS): The reactor core will be damaged by non sufficient cooling with the output of hydrogen because all the specified emergency actions have failed. The overheating of the core is responsible for the production of hydrogen by the reaction of zirconium of the fuel-rod cladding with the water vapour. In case of nuclear superheating it would be possible that the reactor vessel would start smelting. The interacting between the core and the concrete, together with the armouring of the biological shield would also produce hydrogen. The hydrogen would escape together with the water vapour out of the leak and would spread out into the whole containment. Results : the number and the position of the different sized recombiners were determined on engineering judgement. the following 4 scenarios are representatively. The 4 scenarios were analyzed for in beyond design accident conditions with the MELCOR-Code: No. 1: Loss of main feedwater supply with primary feed and bleed. No. 2

  13. The WWER fuel element safety research under the design and heavy accident imitation on the 'PARAMETR' stand

    International Nuclear Information System (INIS)

    Deniskin, V.P.; Nalivaev, V.I.; Parshin, N. Ya.; Fedik, I.I.

    2000-01-01

    Analysis of fuel element behavior in the course of the design and heavy accidents is the component of reactor facility safety prevention. Many tasks of fuel element behavior research may be solved with the help of thermophysical stands. One of such stands implemented in 1991 was thermophysical stand 'PARAMETER'.Several experiments on model assemblies chiefly imitating both heavy accident and design basic accident have already been conducted in 'PARAMETER' stand. There were obtained data about fuel claddings seal failure and deformation condition. In particular it was defined that seal failure of all fuel claddings occurs on stage of fuel element warming, in temperature range (770-900) degree celsius and almost does not depend on inner pressure level

  14. Design and Modeling of RF Power Amplifiers with Radial Basis Function Artificial Neural Networks

    OpenAIRE

    Ali Reza Zirak; Sobhan Roshani

    2016-01-01

    A radial basis function (RBF) artificial neural network model for a designed high efficiency radio frequency class-F power amplifier (PA) is presented in this paper. The presented amplifier is designed at 1.8 GHz operating frequency with 12 dB of gain and 36 dBm of 1dB output compression point. The obtained power added efficiency (PAE) for the presented PA is 76% under 26 dBm input power. The proposed RBF model uses input and DC power of the PA as inputs variables and considers output power a...

  15. Applying probabilistic methods for assessments and calculations for accident prevention

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The guidelines for the prevention of accidents require plant design-specific and radioecological calculations to be made in order to show that maximum acceptable expsoure values will not be exceeded in case of an accident. For this purpose, main parameters affecting the accident scenario have to be determined by probabilistic methods. This offers the advantage that parameters can be quantified on the basis of unambigious and realistic criteria, and final results can be defined in terms of conservativity. (DG) [de

  16. The design of nuclear power plants in the Federal Republic of Germany as regards aircraft accidents

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    On the basis of investigation results of aircraft crashes for the Federal territory and site assessment data of the Ministry of the Interior, air traffic does not present a notable hazard for a nuclear power plant. As a precautionary measure for reducing the remaining risk, design requirements for LWRs were developed which, independently of existing aircraft types, contain mainly an abstract predetermination of an idealized impact load - time diagramme. (HP) [de

  17. Task to Training Matrix Design for Decommissioning Engineer on the basis of Systematic Approach to Training Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Jeong Keun [KHNP, Ulsan (Korea, Republic of)

    2016-10-15

    In nuclear history, before Chernobyl Accident, Three Mile Island (TMI) Accident was the severest accident. For this reason, to resolve the disclosed or potential possibilities of nuclear accident, more than one hundred countermeasures were proposed by United States Nuclear Regulatory Commission (USNRC). Among various recommendations by USNRC, one suggestion was related to training aspect. It was Systematic Approach to Training (SAT) and this event was the initiation of SAT methodology in the world. In Korea, upcoming June 2017, Kori Unit-1 NPP is scheduled to be shut down and it will experience NPP decommissioning for the first time. Present study aims to establish concrete training foundation for NPP decommissioning engineers based on Systematic Approach to Training (SAT) methodology, in particular, Task to Training Matrix (TTM). The objective of this paper is to organize TTM on the basis of SAT for NPP decommissioning engineer. For this reason, eighteen tasks are yielded through Job and Task Analysis (JTA) process. After that, for the settlement of Task to Training Matrix (TTM), various data are determined such as element, condition, standard, knowledge and skill, learning objective and training setting. When it comes to training in nuclear industry, SAT methodology has been the unwavering principle in Korea since NPPs export to UAE.

  18. Current plans to characterize the design basis ground motion at the Yucca Mountain, Nevada Site

    International Nuclear Information System (INIS)

    Simecka, W.B.; Grant, T.A.; Voegele, M.D.; Cline, K.M.

    1992-01-01

    A site at Yucca Mountain Nevada is currently being studied to assess its suitability as a potential host site for the nation's first commercial high level waste repository. The DOE has proposed a new methodology for determining design-basis ground motions that uses both deterministic and probabilistic methods. The role of the deterministic approach is primary. It provides the level of detail needed by design engineers in the characterization of ground motions. The probabilistic approach provides a logical structured procedure for integrating the range of possible earthquakes that contribute to the ground motion hazard at the site. In addition, probabilistic methods will be used as needed to provide input for the assessment of long-term repository performance. This paper discusses the local tectonic environment, potential seismic sources and their associated displacements and ground motions. It also discusses the approach to assessing the design basis earthquake for the surface and underground facilities, as well as selected examples of the use of this type of information in design activities

  19. Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

    2016-02-15

    Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

  20. Proposal of the concept of selection of accidents that release large amounts of radioactive substances in the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Ono, Masato; Honda, Yuki; Takada, Shoji; Sawa, Kazuhiro

    2015-01-01

    In Position, construction and equipment of testing and research reactor to be subjected to the use standards for rules Article 53 (prevention of expansion of the accident to release a large amount of radioactive material) generation the frequency is a lower accident than design basis accident, when what is likely to release a large amount of radioactive material or radiation from the facility has occurred, and take the necessary measures in order to prevent the spread of the accident. There is provided a lower accident than frequency design basis accidents, for those that may release a large amount of radioactive material or radiation. (author)

  1. Off-gas and air cleaning systems for accident conditions in nuclear power plants

    International Nuclear Information System (INIS)

    1993-01-01

    This report surveys the design principles and strategies for mitigating the consequences of abnormal events in nuclear power plants by the use of air cleaning systems. Equipment intended for use in design basis accident and severe accident conditions is reviewed, with reference to designs used in IAEA Member States. 93 refs, 48 figs, 23 tabs

  2. Grid fault and design-basis for wind turbines - Final report

    DEFF Research Database (Denmark)

    Hansen, Anca Daniela; Cutululis, Nicolaos Antonio; Markou, Helen

    , have been performed and compared for two cases, i.e. one when the turbine is immediately disconnected from the grid when a grid fault occurs and one when the turbine is equipped with a fault ride-through controller and therefore it is able to remain connected to the grid during the grid fault......This is the final report of a Danish research project “Grid fault and design-basis for wind turbines”. The objective of this project has been to assess and analyze the consequences of the new grid connection requirements for the fatigue and ultimate structural loads of wind turbines....... The fulfillment of the grid connection requirements poses challenges for the design of both the electrical system and the mechanical structure of wind turbines. The development of wind turbine models and novel control strategies to fulfill the TSO’s requirements are of vital importance in this design. Dynamic...

  3. Breckinridge Project, initial effort. Report XI, Volume V. Critical review of the design basis. [Critical review

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-01-01

    Report XI, Technical Audit, is a compendium of research material used during the Initial Effort in making engineering comparisons and decisions. Volumes 4 and 5 of Report XI present those studies which provide a Critical Review of the Design Basis. The Critical Review Report, prepared by Intercontinental Econergy Associates, Inc., summarizes findings from an extensive review of the data base for the H-Coal process design. Volume 4 presents this review and assessment, and includes supporting material; specifically, Design Data Tabulation (Appendix A), Process Flow Sheets (Appendix B), and References (Appendix C). Volume 5 is a continuation of the references of Appendix C. Studies of a proprietary nature are noted and referenced, but are not included in these volumes. They are included in the Limited Access versions of these reports and may be reviewed by properly cleared personnel in the offices of Ashland Synthetic Fuels, Inc.

  4. Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  5. Accident management-defence in depth in Indian PHWRS

    International Nuclear Information System (INIS)

    Jagannad, V.B.L.; Reddy, V.V.; Hajela, Sameer; Bhatia, C.M.; Nair, Suma

    2015-01-01

    Defence in Depth (DiD) is the established safety principle for the design of Nuclear Power Plants (NPPs). Accident at Fukushima Dai-ichi had highlighted the importance of provisions at Level-4 and 5 of DiD. Post Fukushima accident, on-site measures have been strengthened for Indian Nuclear Power Plants. On procedural front, Accident Management Guidelines have been introduced to handle events more severe than design basis accidents. This paper elaborates enhancement of Defence in Depth provisions for Indian Nuclear Power Plants. (author)

  6. Acoustical features of two Mayan monuments at Chichen Itza: Accident or design?

    Science.gov (United States)

    Lubman, David

    2002-11-01

    Chichen Itza dominated the early postclassic Maya world, ca. 900-1200 C.E. Two of its colossal monuments, the Great Ball Court and the temple of Kukulkan, reflect the sophisticated, hybrid culture of a Mexicanized Maya civilization. The architecture seems intended for ceremony and ritual drama. Deducing ritual practices will advance the understanding of a lost civilization, but what took place there is largely unknown. Perhaps acoustical science can add value. Unexpected and unusual acoustical features can be interpreted as intriguing clues or irrelevant accidents. Acoustical advocates believe that, when combined with an understanding of the Maya worldview, acoustical features can provide unique insights into how the Maya designed and used theater spaces. At Chichen Itza's monuments, sound reinforcement features improve rulers and priests ability to address large crowds, and Ball Court whispering galleries permit speech communication over unexpectedly large distances. Handclaps at Kukulkan stimulate chirps that mimic a revered bird (''Kukul''), thus reinforcing cultic beliefs. A ball striking playing field wall stimulates flutter echoes at the Great Ball Court; their strength and duration arguably had dramatic, mythic, and practical significance. Interpretations of the possible mythic, magic, and political significance of sound phenomena at these Maya monuments strongly suggests intentional design.

  7. Calculation of hydrogen and oxygen uptake in fuel rod cladding during severe accidents using the integral diffusion method -- Preliminary design report

    International Nuclear Information System (INIS)

    Siefken, L.J.

    1999-01-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents.''

  8. Generic evaluation of small break loss-of-coolant accident behavior in Babcock and Wilcox designed 177-FA operating plants

    International Nuclear Information System (INIS)

    1980-01-01

    Slow system depressurization resulting from small break loss-of-coolant accidents (LOCAs) in the reactor coolant system have not, until recently, received detailed analytical study comparable to that devoted to large breaks. Following the TMI-2 accident, the staff had a series of meetings with Babcock and Wilcox (B and W) and the B and W licensees. The staff requested that B and W and the licensees: (1) systematically evaluate plant response for small break loss-of-coolant accidents; (2) address each of the concerns documented in the Michelson report; (3) validate the computer codes used against the TMI-2 accident; (4) extend the break spectrum analysis to very small breaks, giving special consideration to failure of pressurizer valves to close; (5) analyze degraded conditions where AFW is not available; (6) prepare design changes aimed at reducing the probability of loss-of-coolant accidents produced by the failure of a PORV to close; and (7) develop revised emergency procedures for small breaks. This report describes the review of the generic analyses performed by B and W based on the requests stated above

  9. Preliminary worst-case accident analysis to support the conceptual design of a potential repository in tuff

    International Nuclear Information System (INIS)

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

    1984-01-01

    The Nevada Waste Storage Investigations (NNWSI) Project is conducting investigations to determine suitability of a site at Yucca Mountain for development as a high-level waste repository. In support of conceptual design, a preliminary analysis has been performed to identify events that could cause radiological releases from the surface facilities during the operations period. Accidental releases were modeled short-duration release plumes, dispersed under averaged climatic conditions, using the AIRDOS-EPA code. consequences of these accidents, in 50-yr integrated dose commitments to operations personnel, to the minimally exposed member of the public, and to the general population in the surrounding area were calculated. risk to the general public from each event was also assessed. All postulated accidents result in doses to pers of the public that are lower than the 0.5 rem/accident limit set by the NRC in 10 CFR 60. For those accidents that do not involve both fire and breach of waste canisters, doses to operations personnel are behind the NRC limit for routine operations of 5 rem/yr set in 10 CFR 20. Accidents that involve fire and breach of waste canisters may cause doses to some operations personnel that are in excess of this limit

  10. Preliminary worst-case accident analysis to support the conceptual design of a potential repository in tuff

    International Nuclear Information System (INIS)

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

    1984-01-01

    The Nevada Nuclear Waste Storage Investigations (NNWSI) Project is conducting investigations to determine the suitability of a site at Yucca Mountain for development as a high level waste repository. In support of the conceptual design, a preliminary analysis has been performed to identify events that could cause radiological releases from the surface facilities during the operations period. Accidental releases were modeled as short-duration release plumes, dispersed under averaged climatic conditions, using the AIRDOS-EPA code. The consequences of these accidents, in 50-yr integrated dose commitments to operations personnel, to the maximally exposed member of the public, and to the general population in the surrounding area were calculated. The risk to the general public from each event was also assessed. All postulated accidents result in doses to members of the public that are lower than the 0.5 rem/accident limit set by the NRC in 10 CFR 60. For those accidents that do not involve both fire and breach of waste canisters, doses to operations personnel are within the NRC limit for routine operations of 5 rem/yr set in 10 CFR 20. Accidents that involve fire and breach of waste canisters may cause doses to some operations personnel that are in excess of this limit. 18 references, 1 figure, 3 tables

  11. Accident tolerant fuels for LWRs: A perspective

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J., E-mail: zinklesj@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States); Terrani, K.A.; Gehin, J.C.; Ott, L.J.; Snead, L.L. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

    2014-05-01

    The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

  12. Accident tolerant fuels for LWRs: A perspective

    International Nuclear Information System (INIS)

    Zinkle, S.J.; Terrani, K.A.; Gehin, J.C.; Ott, L.J.; Snead, L.L.

    2014-01-01

    The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms

  13. Development of Probabilistic Design Basis Earthquake (DBE) Parameters for Moderate and High Hazard Facilities at INEEL

    International Nuclear Information System (INIS)

    Payne, S. M.; Gorman, V. W.; Jensen, S. A.; Nitzel, M. E.; Russell, M. J.; Smith, R. P.

    2000-01-01

    Design Basis Earthquake (DBE) horizontal and vertical response spectra are developed for moderate and high hazard facilities or Performance Categories (PC) 3 and 4, respectively, at the Idaho National Engineering and Environmental Laboratory (INEEL). The probabilistic DBE response spectra will replace the deterministic DBE response spectra currently in the U.S. Department of Energy Idaho Operations Office (DOE-ID) Architectural Engineering Standards that govern seismic design criteria for several facility areas at the INEEL. Probabilistic DBE response spectra are recommended to DOE Naval Reactors for use at the Naval Reactor Facility at INEEL. The site-specific Uniform Hazard Spectra (UHS) developed by URS Greiner Woodward Clyde Federal Services are used as the basis for developing the DBE response spectra. In 1999, the UHS for all INEEL facility areas were recomputed using more appropriate attenuation relationships for the Basin and Range province. The revised UHS have lower ground motions than those produced in the 1996 INEEL site-wide probabilistic ground motion study. The DBE response spectra were developed by incorporating smoothed broadened regions of the peak accelerations, velocities, and displacements defined by the site-specific UHS. Portions of the DBE response spectra were adjusted to ensure conservatism for the structural design process

  14. A basis for standardized seismic design (SSD) for nuclear power plants/critical facilities

    International Nuclear Information System (INIS)

    O'Hara, T.F.; Jacobson, J.P.; Bellini, F.X.

    1991-01-01

    US Nuclear Power Plants (NPP's) are designed, engineered and constructed to stringent standards. Their seismic adequacy is assured by compliance with regulatory standards and demonstrated by both probabilistic risk assessments (PRAs) and seismic margin studies. However, present seismic siting criteria requires improvement. Proposed changes to siting criteria discussed here will provide a predictable licensing process and a stable regulatory environment. Two recent state-of-the-art studies evaluate the seismic design for all eastern US (EUS) NPP'S: a Lawrence Livermore National Labs study (LLNL, 1989) funded by the NRC and similar research by the Electric Power Research Institute (EPRI, 1989) supported by the utilities. Both confirm that Appendix A 10CFR Part 100 has not provided consistent seismic design levels for all sites. Standardized Seismic Design (SSD) uses a probabilistic framework to accommodate alternative deterministic interpretations. It uses seismic hazard input from EPRI or LLNL to produce consistent bases for future seismic design. SSD combines deterministic and probabilistic insights to provide a comprehensive approach for determining a future site's acceptable seismic design basis

  15. Probabilistic evaluation of concrete containment capacity for beyond design basis internal pressures

    International Nuclear Information System (INIS)

    Tang, H.T.; Dameron, R.A.; Rashid, Y.R.

    1995-01-01

    For beyond design basis internal pressure loading, experimental studies have demonstrated that the most probable failure mode governing the ultimate functional capacity of concrete containments is leak rather than break. Based on leak rates measured in experiments, a prediction formula for leak rate as functions of containment liner size and internal pressure has been postulated. The determination of liner tear is cast in a probabilistic framework. In calculating leakage, particular attention is paid to the evaluation of leakage versus rupture and the loading rates that may be required to leapfrog over a leakage mode. (orig.)

  16. Methodology for assessing the effectiveness of countermeasures in rural settlements in the long term after the Chernobyl accident on the multi-attribute analysis basis

    International Nuclear Information System (INIS)

    Panov, A.V.; Fesenko, S.V.; Aleksakhin, R.M.

    2005-01-01

    The effectiveness of countermeasures in rural settlements affected by the Chernobyl accident was assessed based on a multi-attribute approach, using radiological, economic and socio-psychological parameters. (authors)

  17. Containment integrity analysis under accidents

    International Nuclear Information System (INIS)

    Lin Chengge; Zhao Ruichang; Liu Zhitao

    2010-01-01

    Containment integrity analyses for current nuclear power plants (NPPs) mainly focus on the internal pressure caused by design basis accidents (DBAs). In addition to the analyses of containment pressure response caused by DBAs, the behavior of containment during severe accidents (SAs) are also evaluated for AP1000 NPP. Since the conservatism remains in the assumptions,boundary conditions and codes, margin of the results of containment integrity analyses may be overestimated. Along with the improvements of the knowledge to the phenomena and process of relevant accidents, the margin overrated can be appropriately reduced by using the best estimate codes combined with the uncertainty methods, which could be beneficial to the containment design and construction of large passive plants (LPP) in China. (authors)

  18. Designing Raster Cells as the Basis for Developing Personal Graphic Language

    Directory of Open Access Journals (Sweden)

    Jana Z. Vujić

    2011-05-01

    Full Text Available Continuous work in creating new designer solutions points towards the need to create personal routines as personalcommunication in the relation comprising design, algorithms, and original computer graphics. This paper showsprocedures for developing a control language for creating graphic designs with individual raster elements (screeningelement obtaint by halftoning. Personal commands should set routines in a language understood by the printer andthe designer. The PostScript basis is used because we mix vector and pixel graphics in the same program stream, aswell as different colour systems, and our own raster forms. The printing raster is set with the target of special designmulti-use, and this includes the field of security graphics and art computer reproduction. Each raster form assumesmodifications, creating their raster family. The raster cell content is transformed with PostScript, allowing the settingof basic values, angle and liniature for each pixel separately. Raster cells are mixed in multi-colour graphics to thelevel of individual designs with variable values of parameters determining them.

  19. Why there is a need to revise the Design Basis Threat concept

    International Nuclear Information System (INIS)

    Kondratov, S.; Steinhausler, F.

    2006-01-01

    The terrorist attacks in the USA on 11 September 2001 necessitated a review of the proven concept of the Design Basis Threat (DBT) for nuclear installations. It can be assumed that revised and upgraded DBT will result in costly technical solutions. Since infrastructure deficits and financial limitations in many countries have already limited the practical application of the DBT, the revised threat assessment is likely to worsen the current unsatisfactory situation. Therefore, a new realism in the use of the DBT concept is proposed based on a three-level approach. This will enable countries to tailor the design of their physical protection systems in accordance with their means by implementing either a minimum required security level protecting only against the most probable threat, or an intermediate protection level reflecting the newly introduced AHARA (As High As Reasonably Achievable) principle, or the optimum protection level based on an externally reviewed, fully comprehensive DBT. (author)

  20. DESIGNING ALGORITHMS FOR SOLVING PHYSICS PROBLEMS ON THE BASIS OF MIVAR APPROACH

    Directory of Open Access Journals (Sweden)

    Dmitry Alekseevich Chuvikov

    2017-05-01

    Full Text Available The paper considers the process of designing algorithms for solving physics problems on the basis of mivar approach. The work also describes general principles of mivar theory. The concepts of parameter, relation and class in mivar space are considered. There are descriptions of properties which every object in Wi!Mi model should have. An experiment in testing capabilities of the Wi!Mi software has been carried out, thus the model has been designed which solves physics problems from year 8 school course in Russia. To conduct the experiment a new version of Wi!Mi 2.1 software has been used. The physics model deals with the following areas: thermal phenomena, electric and electromagnetic phenomena, optical phenomena.

  1. Recriticality, a Key Phenomenon to Investigate in Core Disruptive Accident Scenarios of Current and Future Fast Reactor Designs

    International Nuclear Information System (INIS)

    Maschek, W.; Rineiski, A.; Flad, M.; Kriventsev, V.; Gabrielli, F.; Morita, K.

    2012-01-01

    Final comments and conclusions: • Modern plants, should have performed better under Fukushima type event. • In future fast reactor systems significantly higher active and passive safety features are installed, which should cope with events like Fukushima. • One important lesson: put a focus on rare initiators, accident routes and consequences that are neither expected nor have been observed, events that are categorized under ‘black swans’. • Importance of severe accident research demonstrated - both analytically and experimentally for assessing and interpreting accident scenarios and developments. Precondition for developing preventive & mitigative safety measures. Passive safety measures are in the focus of advanced design options and must work under conditions of multiple loads and aggravating events. • Fast reactor systems behavior as the SFR under severe accident conditions: – In fast spectrum systems as the SFR the core is not in its neutronically most reactive configuration and SFRs may be loaded with MAs for waste management; – Recriticalities have a high probability because of the higher enrichment levels; – Short time scales have to be envisioned for core melt-down; – Decay heat levels might be significantly higher, if MA bearing fuel is involved. • Improve design by measures for prevention and/or mitigation of recriticalities; – High reliability of simulations required for proof; • Assessment of fuel relocated on peripheral structures; • Preventive/mitigating measures should not replace containment measures

  2. Conceptual Design of Portable Filtered Air Suction Systems For Prevention of Released Radioactive Gas under Severe Accidents of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Beom W.; Choi, Su Y.; Yim, Man S.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    It becomes evident that severe accidents may occur by unexpected disasters such as tsunami, heavy flood, or terror. Once radioactive material is released from NPP through severe accidents, there are no ways to prevent the released radioactive gas spreading in the air. As a remedy for this problem, the idea on the portable filtered air suction system (PoFASS) for the prevention of released radioactive gas under severe accidents was proposed. In this paper, the conceptual design of a PoFASS focusing on the number of robot fingers and robot arm rods are proposed. In order to design a flexible robot suction nozzle, mathematical models for the gaps which represent the lifted heights of extensible covers for given convex shapes of pipes and for the covered areas are developed. In addition, the system requirements for the design of the robot arms of PoFASS are proposed, which determine the accessible range of leakage points of released radioactive gas. In this paper, the conceptual designs of the flexible robot suction nozzle and robot arm have been conducted. As a result, the minimum number of robot fingers and robot arm rods are defined to be four and three, respectively. For further works, extensible cover designs on the flexible robot suction nozzle and the application of the PoFASS to the inside of NPP should be studied because the radioactive gas may be released from connection pipes between the containment building and auxiliary buildings.

  3. Use of knowledge and experience gained from the Fukushima Daiichi Nuclear Power Station accident to establish the technical basis for strategic off-site response

    International Nuclear Information System (INIS)

    Miyahara, Kaname; Saito, Kimiaki; Iijima, Kazuki; McKinley, Ian; Hardie, Susan

    2015-03-01

    This report provides a concise overview of knowledge and experience gained from the activities for environmental remediation after the Fukushima Daiichi (1F) accident. It is specifically tailored for international use, to establish or refine the technical basis for strategic, off-site response to nuclear incidents. It reflects JAEA's key role in the research associated with both remediation of contaminated areas and also the natural contamination migration processes in non-remediated areas, in collaboration with other Japanese and international organisations and research institutes. Environmental monitoring and mapping to define boundary conditions in terms of the distribution of radioactivity and resultant doses, guides the resultant response. Radiation protection considerations set constraints, with approaches developed to estimate doses to different critical groups and set appropriate dose reduction targets. Decontamination activities, with special emphasis on associated waste management, provide experience in evaluation of the effectiveness of decontamination and the pros and cons of different approaches / technologies. The assessment of the natural behaviour of contaminant radionuclides and their mobility in the environment is now focused almost entirely on radiocaesium. Here, the impact of natural mobility in terms of self-cleaning / re-concentration in cleaned areas is discussed, along with possible actions to modify such transport or manage potential areas of radiocaesium accumulation. Many of the conditions in Fukushima are similar to those following past contamination events in other countries, where natural self-cleaning alone has allowed recovery to such an extent that the original incident is now largely forgotten. Decontamination efforts in Japan will certainly accelerate this process. On-going remediation work is based on a good technical understanding of the movement of radiocaesium in the environment and this understanding is being translated into

  4. Design of turning hydraulic engines for manipulators of mobile machines on the basis of multicriterial optimization

    Directory of Open Access Journals (Sweden)

    Lagerev I.A.

    2016-12-01

    Full Text Available In this paper the mathematical models of the main types of turning hydraulic engines, which at the present time widely used in the construction of handling systems of domestic and foreign mobile transport-technological machines wide functionality. They allow to take into consideration the most significant from the viewpoint of ensuring high technical-economic indicators of hydraulic efficiency criteria – minimum mass (weight, their volume and losses of power. On the basis of these mathematical models the problem of multicriterial constrained optimization of the constructive sizes of turning hydraulic engines are subject to complex constructive, strength and deformation limits. It allows you to de-velop the hydraulic engines in an optimized design which is required for the purpose of designing a comprehensive measure takes into account efficiency criteria. The multicriterial optimization problem is universal in nature, so when designing a turning hydraulic engines allows for one-, two - and three-criteria optimization without making any changes in the solution algorithm. This is a significant advantage for the development of universal software for the automation of design of mobile transport-technological machines.

  5. CARNSORE: Hypothetical reactor accident study

    International Nuclear Information System (INIS)

    Walmod-Larsen, O.; Jensen, N.O.; Kristensen, L.; Meide, A.; Nedergaard, K.L.; Nielsen, F.; Lundtang Petersen, E.; Petersen, T.; Thykier-Nielsen, S.

    1984-06-01

    Two types of design-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are described and their consequences assessed. The PLUCON 2 model was used to calculate the consequences which are presented in terms of individual and collective doses, as well as early and late health consequences. The site proposed for the nucelar power station is Carnsore Point, County Wexford, south-east Ireland. The release fractions for the accidents described are those given in WASH-1400. The analyses are based on the resident population as given in the 1979 census and on 20 years of data from the meteorological stations at Rosslare Harbour, 8.5 km north of the site. The consequences of one of the hypothetical core-melt accidents are described in detail in a meteorological parametric study. Likewise the consequences of the worst conceivable combination of situations are described. Finally, the release fraction in one accident is varied and the consequences of a proposed, more probable ''Class 9 accident'' are presented. (author)

  6. Analysis of severe accidents in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    2008-06-01

    Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

  7. Establishing design basis threats for the physical protection of nuclear materials and facilities

    International Nuclear Information System (INIS)

    Chetvergov, S.

    2001-01-01

    In the area of nuclear energy utilization, the Republic of Kazakhstan follows the standards of international legislation and is a participant of the Nuclear Weapons Non-proliferation Treaty as a country that does not have nuclear weapons. In the framework of this treaty, Kazakhstan provides for the measures to ensure the regime of nonproliferation. The Republic signed the Agreement with the IAEA on the guarantee that was ratified by the Presidential Decree in 1995. Now the Government of the RK is considering the Convention on Physical Protection of Nuclear Materials. Kazakhstan legislation in the area of nuclear energy utilization is represented by a set of laws: the main of them is the Law of the Republic of Kazakhstan 'On the utilization of atomic energy', dated April 14, 1997. According to the Law, the issues of physical protection are regulated by interdepartmental guideline documents. Nuclear science and industry of RK include: Enterprises on uranium mining and processing; Ulba metallurgical plant, manufacturing fuel pellets of uranium dioxide for heat release assemblies of RBMK and WWR reactor types, with the enrichment on U235 1.6-4.4%; Power plant in Aktau for heat and power supply and water desalination, based on fast breeder reactor BN-350; Research reactors of National Nuclear Center: WWR-K - water-water reactor, with 10 MW power, uses highly enriched uranium (up to 36% of U-235); IVG.1M - water-water heterogeneous reactor of vessel type on thermal neutrons, maximum power is 35 MW; IGR - impulse homogeneous graphite reactor on thermal neutrons, with graphite reflector; RA - high temperature gas cooled reactor on thermal neutrons, 0.5 MW power. The establishment of design basis threats for nuclear objects in the Republic of Kazakhstan is an urgent problem because of the developing military-political situation in the region. It is necessary to specify important elements affecting the specific features of the design basis threat: military operations of

  8. Strategies for the prevention and mitigation of severe accidents

    International Nuclear Information System (INIS)

    Ader, C.; Heusener, G.; Snell, V.G.

    1999-01-01

    The currently operating nuclear power plants have, in general, achieved a high level of safety, as a result of design philosophies that have emphasized concepts such as defense-in-depth. This type of an approach has resulted in plants that have robust designs and strong containments. These designs were later found to have capabilities to protect the public from severe accidents (accidents more severe than traditional design basis in which substantial damage is done to the reactor core). In spite of this high level of safety, it has also been recognized that future plants need to be designed to achieve an enhanced level of safety, in particular with respect to severe accidents. This has led both regulatory authorities and utilities to develop guidance and/or requirements to guide plant designers in achieving improved severe accident performance through prevention and mitigation. The considerable research programs initiated after the TMI-2 accident have provided a large body of technical data, analytical methods, and the expertise necessary to provide for an understanding of a range of severe accident phenomena. This understanding of the ways severe accidents can progress and challenge containments, combined with the wide use of probabilistic safety assessments, have provided designers of evolutionary water cooled reactors opportunities to develop designs that minimize the challenges to the plant and to the public from severe accidents, including the development of accident management strategies intended to further reduce the risk of severe accidents. This paper describes some of the recent progress made in the understanding of severe accidents and related safety assessment methodology and how this knowledge has supported the incorporation of features into representative evolutionary designs that will prevent or mitigate many of the severe accident challenges present in current plants. (author)

  9. A calculational methodology for comparing the accident, occupational, and waste-disposal hazards of fusion reactor designs

    International Nuclear Information System (INIS)

    Fetter, S.

    1985-01-01

    A methodology has been developed for calculating indices of three classes of radiological hazards: reactor accidents, occupational exposures, and waste-disposal hazards. Radionuclide inventories, biological hazard potentials (BHP), and various dose-related indices are calculated. In the case of reactor accidents, the critical, 50-year and chronic dose are computed, as well as the number of early deaths and illnesses and late cancer fatalities. For occupational exposure, the contact dose rate is calculated for several times after reactor shutdown. In the case of waste-disposal hazards, the intruder dose and the intruder hazard potential (IHP) are calculated. Sample calculations for the MARS reactor design show the usefulness of the methodology in exploring design improvements

  10. Basis of the tubesheet heat exchanger design rules used in the French pressure vessel code

    International Nuclear Information System (INIS)

    Osweiller, F.

    1990-01-01

    For about 40 years most tubesheet heat exchangers have been designed according to the standards of TEMA. Partly due to their simplicity, these rules do not assure a safe heat-exchangers design in all cases. This is the main reason why new tubesheet design rules were developed in 1981 in France for the French pressure vessel code CODAP. For fixed tubesheet heat exchangers the new rules account for the elastic rotational restraint of the shell and channel at the outer edge of the tubesheet. For floating-head and U- tube exchangers an approach was selected with some modifications. In both cases the tubesheet is replaced by an equivalent solid plate with adequate effective elastic constants, and the tube bundle is simulated by an elastic foundation. The elastic restraint at the edge of the tubesheet due the shell and channel is accounted for in different ways in the two types of heat exchangers. The purpose of the paper is to present the main basis of these rules and to compare them to TEMA rules

  11. Technical meeting on progress in managing, and limiting the consequences of events exceeding the design basis

    International Nuclear Information System (INIS)

    Fabian, H.

    2004-01-01

    The Technical Groups on 'Reactor Safety' and 'Thermodynamics and Fluid Dynamics' of the Kerntechnische Gesellschaft e.V. organized a joint technical meeting on 'Progress in Managing, and Limiting the Consequences of, Events Exceeding the Design Basis' at the FTU Training Center of the Karlsruhe Research Center. The topic chosen, the papers presented, the presenters, and the non-technical part of the program met with lively interest on the part of institutions in the nuclear field. These were the objectives of the technical meeting: - Establishing a forum for communicating relevant topics. - In-depth discussion of the main topic, i.e. the advanced development of reactor safety, research in the field, and its application, in twenty selected papers presented by speakers from different institutions. - Presentation of topical work in a nuclear technology institution, the Karlsruhe Research Center. (orig.) [de

  12. Facts learnt from the Hanshin-Awaji disaster and consideration on design basis earthquake

    International Nuclear Information System (INIS)

    Shibata, Heki

    1997-01-01

    This paper will deal with how to establish the concept of the design basis earthquake for critical industrial facilities such as nuclear power plants in consideration of disasters induced by the 1995 Hyogoken-Nanbu Earthquake (Southern Hyogo-prefecture Earthquake-1995), so-called Kobe earthquake. The author once discussed various DBEs at 7 WCEE. At that time, the author assumed that the strongest effective PGA would be 0.7 G, and compared to the values of accelerations to a structure obtained by various codes in Japan and other countries. The maximum PGA observed by an instrument at the Southern Hyogo-pref. Earthquake-1995 exceeded the previous assumption of the author, even though the evaluation results of the previous paper had been pessimistic. According to the experience of Kobe event, the author will point out the necessity of the third earthquake S s adding to S 1 and S 2 , previous DBEs. (author)

  13. Facts learnt from the Hanshin-Awaji disaster and consideration on design basis earthquake

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Heki [Yokohama National Univ. (Japan). Faculty of Engineering

    1997-03-01

    This paper will deal with how to establish the concept of the design basis earthquake for critical industrial facilities such as nuclear power plants in consideration of disasters induced by the 1995 Hyogoken-Nanbu Earthquake (Southern Hyogo-prefecture Earthquake-1995), so-called Kobe earthquake. The author once discussed various DBEs at 7 WCEE. At that time, the author assumed that the strongest effective PGA would be 0.7 G, and compared to the values of accelerations to a structure obtained by various codes in Japan and other countries. The maximum PGA observed by an instrument at the Southern Hyogo-pref. Earthquake-1995 exceeded the previous assumption of the author, even though the evaluation results of the previous paper had been pessimistic. According to the experience of Kobe event, the author will point out the necessity of the third earthquake S{sub s} adding to S{sub 1} and S{sub 2}, previous DBEs. (author)

  14. Design Basis Knowledge Management for New Build Projects & Ageing Plants - A Perspective

    International Nuclear Information System (INIS)

    Weightman, Mike

    2013-01-01

    Summary: • KM for Design Basis of New and Ageing nuclear facilities is at a crossroads; • Needs leadership, vision, cultural change and resources; • Outcome of this workshop is vital; • Information is not knowledge; • Knowledge includes the WHAT, the HOW, the WHY, the Environment and, importantly, Application; • In general, Industry and Regulators are behind the curve; • Develop and apply the principles rigorously; • Keep it simple - focus first on Leadership, values (e.g. questioning attitude), culture, and prioritise – risk informed; • KM is a complex organic creature and needs to be nurtured, fed, learn, grow, evolve in response to a changing environment, and discharge what is not needed to prosper

  15. Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

    International Nuclear Information System (INIS)

    Fedotov, A.

    2003-01-01

    The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

  16. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in combustion engineering designed operating plants

    International Nuclear Information System (INIS)

    1980-01-01

    The purpose of this report is to summarize the results of a generic evaluation of feedwater transients, small break loss-of-coolant accidents (LOCAs), and other TMI-2-related events in the Combustion Engineering (CE)-designed operating plants and to establish or confirm the bases for their continued operation. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas

  17. Development of two-dimensional hot pool model and analysis of the ULOHS accident in KALIMER design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Jeong, K. S.; Hahn, H. D

    2000-10-01

    In the new version of HP2D program, the variation model of the hot pool sodium level is added so that the temperature and velocity profiles can be predicted more accurately than old version. To verify and validate the developed new version model, comparison of the MONJU experimental data with the predicted one is performed and analyzed. And also the ULOHS(Unprotected Loss of Heat Sink) accident in the KALIMER design is performed and analyzed.

  18. Development of two-dimensional hot pool model and analysis of the ULOHS accident in KALIMER design

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Jeong, K. S.; Hahn, H. D.

    2000-10-01

    In the new version of HP2D program, the variation model of the hot pool sodium level is added so that the temperature and velocity profiles can be predicted more accurately than old version. To verify and validate the developed new version model, comparison of the MONJU experimental data with the predicted one is performed and analyzed. And also the ULOHS(Unprotected Loss of Heat Sink) accident in the KALIMER design is performed and analyzed

  19. Design basis of off-site emergency response plans for fuel cycle installations

    International Nuclear Information System (INIS)

    Rzepka, J.P.; Dubiau, Ph.; Jouve, A.C.; Charles, T.; Mercier, J.P.

    1995-01-01

    In France, the term 'off-site emergency response plan' refers to all the arrangements which should be made by the government authorities to protect the population in the event of an accident affecting the installations of the site considered. The outline of the method of defining typical accidents, evaluation of 'source-terms' and health consequences is presented. Two applications to installations from the front-end and from the back-end of the fuel cycle are discussed. (K.A.). 1 tab

  20. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    International Nuclear Information System (INIS)

    Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn; Morrow, Charles; Osborn, Douglas; Gauntt, Randall O.

    2015-01-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  1. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Chisom Shawn [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, Douglas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  2. Modeling of Li-Ion Battery Packs as Basis for Design of Battery Thermal Management Systems

    DEFF Research Database (Denmark)

    Coman, Paul Tiberiu

    . The shortcomings of safety were reflected in the recent accidents, where fires and explosions were reported in cell phones, electric cars, laptops, e-hovers and even airplanes. The goal of this thesis is to generate knowledge, understanding and methods to ensure safety in Li-ion cells and packs. For achieving...

  3. Why there is a need to revise the design basis threat concept

    International Nuclear Information System (INIS)

    Kondratov, S.; Steinhaeusler, F.

    2005-01-01

    Full text: The coordinated terrorist attacks in the United States on September 11, 2001, necessitated the review of the proven concept of the Design Basis Threat (DBT) for nuclear installations. It is safe to assume that revised and upgraded DBT will result in costly technical solutions. Since infrastructure deficits and financial limitations in many countries have already limited the practical application of the DBT in many instances, the revised threat assessment is likely to worsen the current dissatisfactory situation further. Therefore, a new realism in the use of the DBT concept is proposed, based on a three-level approach. This will enable countries to tailor the design of their physical protection systems in accordance with their means by implementing either a minimum required security level protecting only against the most probable threat, or aiming for an intermediate protection level reflecting the newly introduced AHARA - as high as reasonably achievable - principle, or providing the optimum protection level based on an externally reviewed, fully comprehensive DBT. (author)

  4. Comparison of deterministic and stochastic techniques for estimation of design basis floods for nuclear power plants

    International Nuclear Information System (INIS)

    Solomon, S.I.; Harvey, K.D.

    1982-12-01

    The IAEA Safety Guide 50-SG-S10A recommends that design basis floods be estimated by deterministic techniques using probable maximum precipitation and a rainfall runoff model to evaluate the corresponding flood. The Guide indicates that stochastic techniques are also acceptable in which case floods of very low probability have to be estimated. The paper compares the results of applying the two techniques in two river basins at a number of locations and concludes that the uncertainty of the results of both techniques is of the same order of magnitude. However, the use of the unit hydrograph as the rainfall runoff model may lead in some cases to nonconservative estimates. A distributed non-linear rainfall runoff model leads to estimates of probable maximum flood flows which are very close to values of flows having a 10 6 - 10 7 years return interval estimated using a conservative and relatively simple stochastic technique. Recommendations on the practical application of Safety Guide 50-SG-10A are made and the extension of the stochastic technique to ungauged sites and other design parameters is discussed

  5. Comparison of deterministic and stochastic techniques for estimation of design basis floods for nuclear power plants

    International Nuclear Information System (INIS)

    Solomon, S.I.; Harvey, K.D.; Asmis, G.J.K.

    1983-01-01

    The IAEA Safety Guide 50-SG-S10A recommends that design basis floods be estimated by deterministic techniques using probable maximum precipitation and a rainfall runoff model to evaluate the corresponding flood. The Guide indicates that stochastic techniques are also acceptable in which case floods of very low probability have to be estimated. The paper compares the results of applying the two techniques in two river basins at a number of locations and concludes that the uncertainty of the results of both techniques is of the same order of magnitude. However, the use of the unit hydrograph as the rain fall runoff model may lead in some cases to non-conservative estimates. A distributed non-linear rainfall runoff model leads to estimates of probable maximum flood flows which are very close to values of flows having a 10 6 to 10 7 years return interval estimated using a conservative and relatively simple stochastic technique. Recommendations on the practical application of Safety Guide 50-SG-10A are made and the extension of the stochastic technique to ungauged sites and other design parameters is discussed

  6. Severe accident analysis and management in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Golshan, Mina

    2013-01-01

    Within the UK regulatory regime, assessment of risks arising from licensee's activities are expected to cover both normal operations and fault conditions. In order to establish the safety case for fault conditions, fault analysis is expected to cover three forms of analysis: design basis analysis (DBA), probabilistic safety assessment (PSA) and severe accident analysis (SAA). DBA should provide a robust demonstration of the fault tolerance of the engineering design and the effectiveness of the safety measures on a conservative basis. PSA looks at a wider range of fault sequences (on a best estimate basis) including those excluded from the DBA. SAA considers significant but unlikely accidents and provides information on their progression and consequences, within the facility, on the site and off site. The assessment of severe accidents is not limited to nuclear power plants and is expected to be carried out for all plant states where the identified dose targets could be exceeded. This paper sets out the UK nuclear regulatory expectation on what constitutes a severe accident, irrespective of the type of facility, and describes characteristics of severe accidents focusing on nuclear fuel cycle facilities. Key rules in assessment of severe accidents as well as the relationship to other fault analysis techniques are discussed. The role of SAA in informing accident management strategies and offsite emergency plans is covered. The paper also presents generic examples of scenarios that could lead to severe accidents in a range of nuclear fuel cycle facilities. (authors)

  7. Design and utilisation of protocols to characterise dynamic PET uptake of two tracers using basis pursuit

    Science.gov (United States)

    Bell, Christopher; Puttick, Simon; Rose, Stephen; Smith, Jye; Thomas, Paul; Dowson, Nicholas

    2017-06-01

    Imaging using more than one biological process using PET could be of great utility, but despite previously proposed approaches to dual-tracer imaging, it is seldom performed. The alternative of performing multiple scans is often infeasible for clinical practice or even in research studies. Dual-tracer PET scanning allows for multiple PET radiotracers to be imaged within the same imaging session. In this paper we describe our approach to utilise the basis pursuit method to aid in the design of dual-tracer PET imaging experiments, and later in separation of the signals. The advantage of this approach is that it does not require a compartment model architecture to be specified or even that both signals are distinguishable in all cases. This means the method for separating dual-tracer signals can be used for many feasible and useful combinations of biology or radiotracer, once an appropriate scanning protocol has been decided upon. Following a demonstration in separating the signals from two consecutively injected radionuclides in a controlled experiment, phantom and list-mode mouse experiments demonstrated the ability to test the feasibility of dual-tracer imaging protocols for multiple injection delays. Increases in variances predicted for kinetic macro-parameters V D and K I in brain and tumoral tissue were obtained when separating the synthetically combined data. These experiments confirmed previous work using other approaches that injections delays of 10-20 min ensured increases in variance were kept minimal for the test tracers used. On this basis, an actual dual-tracer experiment using a 20 min delay was performed using these radio tracers, with the kinetic parameters (V D and K I) extracted for each tracer in agreement with the literature. This study supports previous work that dual-tracer PET imaging can be accomplished provided certain constraints are adhered to. The utilisation of basis pursuit techniques, with its removed need to specify a model

  8. Improvement design study on steam generator of MHR-50/100 aiming higher safety level after water ingress accident

    International Nuclear Information System (INIS)

    Oyama, S.; Minatsuki, I.; Shimizu, K.

    2012-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has been studying on MHI original High Temperature Gas cooled Reactor (HTGR), namely MHR-50/100, for commercialization with supported by JAEA. In the heat transfer system, steam generator (SG) is one of the most important components because it should be imposed a function of heat transfer from reactor power to steam turbine system and maintaining a nuclear grade boundary. Then we especially focused an effort of a design study on the SG having robustness against water ingress accident based on our design experience of PWR, FBR and HTGR. In this study, we carried out a sensitivity analysis from the view point of economic and plant efficiency. As a result, the SG design parameter of helium inlet/outlet temperature of 750 deg. C/300 deg. C, a side-by-side layout and one unit of SG attached to a reactor were selected. In the next, a design improvement of SG was carried out from the view point of securing the level of inherent safety without reliance on active steam dump system during water ingress accident considering the situation of the Fukushima nuclear power plant disaster on March 11, 2011. Finally, according to above basic design requirement to SG, we performed a conceptual design on adapting themes of SG structure improvement. (authors)

  9. Development, Use and Maintenance of the Design Basis Threat. Implementing Guide (Arabic Edition)

    International Nuclear Information System (INIS)

    2009-01-01

    threat to those assets. As described in this publication, an understanding of the threat can lead to a detailed description of potential adversaries (the design basis threat), which, in turn, is the basis of an appropriately designed physical protection system. This direct link gives confidence that protection would be effective against an adversary attack. International experience in using a design basis threat to protect assets of high consequence is largely based on the protection of nuclear material and facilities. Furthermore, the nuclear security documents defining and recommending that physical protection be based upon the threat. The Physical Protection Objectives and Fundamental Principles, the Recommendations on the Physical Protection of Nuclear Facilities and Nuclear Material, and the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended - do so exclusively for the protection of nuclear material and facilities. Given the historical background, and its continuing contemporary relevance, it has been necessary to draw on that nuclear protection experience in developing this publication. However, the general approach can also be applied to protecting other assets that require a high degree of confidence in the effectiveness of their protection, such as high-activity radioactive material. Specialists from France, Germany, Japan, the Russian Federation, Spain, the United Kingdom, and the United States of America assisted the IAEA in preparing this publication. A draft was presented to an open-ended technical meeting in December 2006, and subsequently circulated for comment to all Member States. This publication is consistent with The Physical Protection Objectives and Fundamental Principles; the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended; and the Recommendations on the Physical Protection of Nuclear Facilities and Nuclear Material.

  10. Development, use and maintenance of the design basis threat. Implementing guide

    International Nuclear Information System (INIS)

    2009-01-01

    threat to those assets. As described in this publication, an understanding of the threat can lead to a detailed description of potential adversaries (the design basis threat), which, in turn, is the basis of an appropriately designed physical protection system. This direct link gives confidence that protection would be effective against an adversary attack. International experience in using a design basis threat to protect assets of high consequence is largely based on the protection of nuclear material and facilities. Furthermore, the nuclear security documents defining and recommending that physical protection be based upon the threat - The Physical Protection Objectives and Fundamental Principles (GOV/2001/41/ Attachment), the Recommendations on the Physical Protection of Nuclear Facilities and Nuclear Material (INFCIRC/225/Rev. 4 (corrected)), and the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended (INFCIRC/274) (adopted on 8 July 2005; (GOV/2005/57)) - do so exclusively for the protection of nuclear material and facilities. Given the historical background, and its continuing contemporary relevance, it has been necessary to draw on that nuclear protection experience in developing this publication. However, the general approach can also be applied to protecting other assets that require a high degree of confidence in the effectiveness of their protection, such as high-activity radioactive material. Specialists from France, Germany, Japan, the Russian Federation, Spain, the United Kingdom, and the United States of America assisted the IAEA in preparing this publication. A draft was presented to an open-ended technical meeting in December 2006, and subsequently circulated for comment to all Member States. This publication is consistent with The Physical Protection Objectives and Fundamental Principles; the Convention on Physical Protection of Nuclear Facilities and Nuclear Material as Amended; and the Recommendations on the

  11. Design basis and design features of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia)

    International Nuclear Information System (INIS)

    1994-05-01

    The prime objective of the IAEA Technical Co-operation Project on Evaluation of Safety Aspects of WWER-440 model 213 NPPs is to co-ordinate and to integrate assistance to national organizations in studying selected aspects of safety for the same type of reactors. Consequently, the study integrated the results generated by national activities carried out in the Czech Republic, Hungary, Slovakia and Ukraine and co-ordinated through the IAEA. Valuable assistance in carrying out the tasks was also provided by Bulgaria and Poland. A set of publications is being prepared to present the results of the project. The publications are intended to facilitate the review and utilization of the results of the project. They are also providing assistance in further refinement and/or extension of plant specific safety evaluation of model 213 NPPs. This Technical Document addressing the design basis and safety related design features of WWER-440 model 213 plants is the first of the series to be published. It is hoped that this document will be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, 36 figs, tabs

  12. Assessment of severe accident prevention and mitigation features: PWR, large dry containment design

    International Nuclear Information System (INIS)

    Perkins, K.R.; Hsu, C.J.; Lehner, J.R.; Luckas, W.J.; Cho, N.; Fitzpatrick, R.G.; Pratt, W.T.; Eltawila, F.; Maly, J.A.

    1988-07-01

    Plant features and operator actions which have been found to be important in either preventing or mitigating severe accidents in PWRs with large dry containments have been identified. These features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Zion plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the large dry containment to severe accident containment loads were also identified. In addition, those features of a PWR with a large dry containment, which are important for preventing core damage and are available for mitigating fission-product release to the environment were identified. The report is issued to provide focus to the analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic tributes for assessing those features and actions found to be helpful in reducing the overall risk for Zion and other PWRs with large dry containments. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  13. Assessment of severe accident prevention and mitigation features: PWR, ice-condenser containment design

    International Nuclear Information System (INIS)

    Hsu, C.J.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Cho, N.; Lehner, J.R.; Pratt, W.T.; Eltawila, F.; Maly, J.A.

    1988-07-01

    Plant features and operator actions which have been found to be important in either preventing and mitigating severe accidents in PWRs with ice-condenser containments have been identified. Thus features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Sequoyah plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the ice-condenser containment to sever accident containment loads were also identified. In addition, those features of a PWR with an ice-condenser containment, which are important for preventing core damage and are available for mitigating fission-product release to the environment were identified. This report is issued to provide focus to an analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Sequoyah and other PWRs with ice-condenser containments. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance. 14 tabs

  14. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  15. SRS BEDROCK PROBABILISTIC SEISMIC HAZARD ANALYSIS (PSHA) DESIGN BASIS JUSTIFICATION (U)

    Energy Technology Data Exchange (ETDEWEB)

    (NOEMAIL), R

    2005-12-14

    This represents an assessment of the available Savannah River Site (SRS) hard-rock probabilistic seismic hazard assessments (PSHAs), including PSHAs recently completed, for incorporation in the SRS seismic hazard update. The prior assessment of the SRS seismic design basis (WSRC, 1997) incorporated the results from two PSHAs that were published in 1988 and 1993. Because of the vintage of these studies, an assessment is necessary to establish the value of these PSHAs considering more recently collected data affecting seismic hazards and the availability of more recent PSHAs. This task is consistent with the Department of Energy (DOE) order, DOE O 420.1B and DOE guidance document DOE G 420.1-2. Following DOE guidance, the National Map Hazard was reviewed and incorporated in this assessment. In addition to the National Map hazard, alternative ground motion attenuation models (GMAMs) are used with the National Map source model to produce alternate hazard assessments for the SRS. These hazard assessments are the basis for the updated hard-rock hazard recommendation made in this report. The development and comparison of hazard based on the National Map models and PSHAs completed using alternate GMAMs provides increased confidence in this hazard recommendation. The alternate GMAMs are the EPRI (2004), USGS (2002) and a regional specific model (Silva et al., 2004). Weights of 0.6, 0.3 and 0.1 are recommended for EPRI (2004), USGS (2002) and Silva et al. (2004) respectively. This weighting gives cluster weights of .39, .29, .15, .17 for the 1-corner, 2-corner, hybrid, and Greens-function models, respectively. This assessment is judged to be conservative as compared to WSRC (1997) and incorporates the range of prevailing expert opinion pertinent to the development of seismic hazard at the SRS. The corresponding SRS hard-rock uniform hazard spectra are greater than the design spectra developed in WSRC (1997) that were based on the LLNL (1993) and EPRI (1988) PSHAs. The

  16. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  17. Evaluation of the Main Steam Line Break Accident for the APR+ Standard Design using MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Park, M. H.; Kim, Y. S.; Hwang, Min Jeong; Sim, S. K. [Environment Energy Technology, Daejeon (Korea, Republic of); Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    As a part of licensing evaluation of the APR+ (Advanced Power Reactor +) standard design, Korea Institute of Nuclear Safety(KINS) performed safety evaluation of the APR+ Standard Safety Analysis Report(SSAR). The results of the safety evaluation of the APR+ Main Steam Line Break(MSLB) accident is presented for the most limiting post-trip return-to-power case with the single failure assumption of the Loss Of Offsite Power(LOOP). MARS-KS regulatory safety analysis code has been used to evaluate safety as well as the system behavior during MSLB accident. The MARS-KS analysis results are compared with the results of the MSLB safety analysis presented in the SSAR of the APR+. Independent safety evaluation has been performed using MARS-KS regulatory safety analysis code for the APR+ MSLB accident inside containment for the limiting case of the full power post-trip return-to-power. The results of MARS-KS analysis were compared with the results of the MSLB safety analysis presented in the APR+ SSAR. Due to higher cooldown of the MARS-KS analysis, the MARS-KS analysis results in a higher positive reactivity insertion into the core by the negative moderator and fuel temperature reactivity coefficients than the APR+ SSAR analysis. Both results show no return-to-power during the limiting case of the MSLB inside containment. However, APR+ SSAR moderator temperature reactivity insertion should be evaluated against the MARS-KS moderator density reactivity insertion for is conservatism. This study also clearly shows asymmetric thermal hydraulic behavior during the MSLB accident at intact and affected sides of the downcomer and the core. These asymmetric phenomena should be further investigated for the effects on the system design.

  18. Design of environment monitoring system to evaluate radionuclide release from subsystem on PWR nuclear power accident

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Sugiyanto; Pande Made Udiyani; Jupiter Sitorus Pane

    2012-01-01

    Nuclear Power Plan (NPP) as a renewable energy source is selected as an alternative, because it has many advantages that is environmentally friendly, fuel supply which is independent of the season, and the price that can compete with other power plants. However, the existence of some public skepticism about nuclear radiation safety, the government must be convinced about the operation of nuclear power plants are safe and secure. Research on the design of environment monitoring system for evaluation of radionuclide release from the reactor subsystems and the environment due to accidents at power reactors has been done. The study was conducted by calculating the distribution of radionuclide release into the reactor subsystem and the environment and also to build the environment radiation monitoring system. Environmental monitoring system consists of a radiation counter, early warning systems, meteorological measurement systems, GPS systems and GIS. Radiation monitoring system used to record the data of radiation, meteorological measurement system used to record data of wind and speed direction, while the GPS system is used to determine position of data measurements. The data is then transmitted to a data acquisition system and then to be transmitted to the control center. Collection and transmission of data is done via SMS formatting using a modem device that is placed in the control center. The control center receives measurement data from various places. In this case the control center has a function as an SMS Gateway. This system can visualize for different measurement locations. Furthermore, radiation data and position data to be integrated with digital maps. System integration is then visualized in a personal computer. To position of measurements directly visualized on the map and also look for the data displayed on a monitor as a red or green circle colour. That colour indicated as a safe limit of radiation monitor. When the cycle colour is red, the system will

  19. Advances in safety countermeasures at the Tomari NPP of Hokkaido Electric Power on the basis of Fukushima Daiichi NPP accident. Fire protection and other advances

    International Nuclear Information System (INIS)

    Shibata, Taku; Dasai, Katsumi

    2014-01-01

    Fire protections for the nuclear power plants have been based on the fire laws and the conventional guide. After Fukushima Daiichi NPP accident, many safety countermeasures - also about Fire Protection - have been discussed in the Japanese authorities. This paper shows our present activities in the Tomari NPP about the fire protections from the view points of Fire Prevention, Fire Detection/Suppression Systems and Fire Protection, and other advances. (author)

  20. Importance of the Design Basis Knowledge Management for New Builds and Existing NPPs

    International Nuclear Information System (INIS)

    Bychkov, Alexander

    2013-01-01

    The Fukushima Daiichi NPP accident in Japan reminds us that sustaining a high level of nuclear safety over the life of a plant is of paramount importance to us all. This and previous major accidents have shaken the confidence of the public and even some governments in the safety of nuclear energy. However, a key message from the St. Petersburg Ministerial Conference this past June was that nuclear power will continue to play an important role in energy security and sustainable development. The IAEA will obviously need to continue to assist Member States in strengthening the safety of NPPs. We will also continue to give technical support to both newcomer countries and to countries with established nuclear energy programmes

  1. Grid fault and design-basis for wind turbines. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, A.D.; Cutululis, N.A.; Markou, H.; Soerensen, Poul; Iov, F.

    2010-01-15

    This is the final report of a Danish research project 'Grid fault and design-basis for wind turbines'. The objective of this project has been to assess and analyze the consequences of the new grid connection requirements for the fatigue and ultimate structural loads of wind turbines. The fulfillment of the grid connection requirements poses challenges for the design of both the electrical system and the mechanical structure of wind turbines. The development of wind turbine models and novel control strategies to fulfill the TSO's requirements are of vital importance in this design. Dynamic models and different fault ride-through control strategies have been developed and assessed in this project for three different wind turbine concepts (active stall wind turbine, variable speed doublyfed induction generator wind turbine, variable speed multipole permanent magnet wind turbine). A computer approach for the quantification of the wind turbines structural loads caused by the fault ride-through grid requirement, has been proposed and exemplified for the case of an active stall wind turbine. This approach relies on the combination of knowledge from complimentary simulation tools, which have expertise in different specialized design areas for wind turbines. In order to quantify the impact of the grid faults and grid requirements fulfillment on wind turbines structural loads and thus on their lifetime, a rainflow and a statistical analysis for fatigue and ultimate structural loads, respectively, have been performed and compared for two cases, i.e. one when the turbine is immediately disconnected from the grid when a grid fault occurs and one when the turbine is equipped with a fault ride-through controller and therefore it is able to remain connected to the grid during the grid fault. Different storm control strategies, that enable variable speed wind turbines to produce power at wind speeds higher than 25m/s and up to 50m/s without substantially increasing

  2. Licensing decisions and safety research related to LMFBR accidents

    International Nuclear Information System (INIS)

    Denise, R.P.; Speis, T.P.; Kelber, C.N.; Curtis, R.T.

    1977-01-01

    The licensing approach which ensures adequate protection of the public health and safety against serious accidents is described. This paper describes the role of core melt and core disruptive accidents in the design, safety research, and licensing processes, using the Clinch River Breeder Reactor (CRBR) as a focal point. Major design attention is placed on the prevention of these accidents so that the probability of core melt accidents is reduced to a sufficiently low level that they are not treated as design basis accidents. Additional requirements are placed upon the design to further reduce residual risk. This licensing process is supported by a confirmatory research program designed to provide an independent basis for licensing judgements. It has as a goal the resolution of generic safety issues prior to the establishment of a commercial LMFBR industry. The program includes accident analysis, experiments in materials interactions, aerosol transport and system integrity and planning for new safety test facilities. The problems are approached in a multi-disciplinary functional manner that identifies key safety issues and centralizes efforts to resolve them. The near term objectives of the program support the licensing of the Clinch River Breeder Reactor (CRBR) and the proposed Prototype Large Breeder Reactor (PLBR). The long term objectives of the program support the licensing of commercial LMFBRs during the late 1980's and beyond. This safety research is designed to provide an independent basis for the licensing judgements which must be made by the Nuclear Regulatory Commission

  3. Ideal flow theory for the double - shearing model as a basis for metal forming design

    Science.gov (United States)

    Alexandrov, S.; Trung, N. T.

    2018-02-01

    In the case of Tresca’ solids (i.e. solids obeying the Tresca yield criterion and its associated flow rule) ideal flows have been defined elsewhere as solenoidal smooth deformations in which an eigenvector field associated everywhere with the greatest principal stress (and strain rate) is fixed in the material. Under such conditions all material elements undergo paths of minimum plastic work, a condition which is often advantageous for metal forming processes. Therefore, the ideal flow theory is used as the basis of a procedure for the preliminary design of such processes. The present paper extends the theory of stationary planar ideal flow to pressure dependent materials obeying the double shearing model and the double slip and rotation model. It is shown that the original problem of plasticity reduces to a purely geometric problem. The corresponding system of equations is hyperbolic. The characteristic relations are integrated in elementary functions. In regions where one family of characteristics is straight, mapping between the principal lines and Cartesian coordinates is determined by linear ordinary differential equations. An illustrative example is provided.

  4. Design of elliptic curve cryptoprocessors over GF(2^163 using the Gaussian normal basis

    Directory of Open Access Journals (Sweden)

    Paulo Cesar Realpe

    2014-05-01

    Full Text Available This paper presents the efficient hardware implementation of cryptoprocessors that carry out the scalar multiplication kP over finite field GF(2163 using two digit-level multipliers. The finite field arithmetic operations were implemented using Gaussian normal basis (GNB representation, and the scalar multiplication kP was implemented using Lopez-Dahab algorithm, 2-NAF halve-and-add algorithm and w-tNAF method for Koblitz curves. The processors were designed using VHDL description, synthesized on the Stratix-IV FPGA using Quartus II 12.0 and verified using SignalTAP II and Matlab. The simulation results show that the cryptoprocessors present a very good performance to carry out the scalar multiplication kP. In this case, the computation times of the multiplication kP using Lopez-Dahab, 2-NAF halve-and-add and 16-tNAF for Koblitz curves were 13.37 µs, 16.90 µs and 5.05 µs, respectively.

  5. Calculation of particulate dispersion in a design-basis tornadic storm from Westinghouse PFDL, Cheswick, Pennsylvania

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1978-07-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Westinghouse Plutonium Fuel Development Laboratory (PFDL) at Cheswick, Pennsylvania. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20-45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

  6. Studies of severe accidents in light water reactors. Containment performance

    International Nuclear Information System (INIS)

    Hayns, M.R.; Phillips, D.W.; Young, R.L.D.

    1987-01-01

    The containment system of a LWR is an obvious component of the plant which performs an important safety function in preventing the release of fission products to the environment in the event of design basis accidents. With over 260 LWRs in service worldwide, and others still under construction, there is a considerable diversity of containment types and combinations of containment safeguards systems. All of these satisfy local regulatory requirements which are principally aimed at the design basis accidents, and these requirements naturally have a considerable uniformity. However, their design diversity becomes more relevant to the performance of the containment in severe accident conditions, and this aspect of containment performance is reviewed in this paper. The ability of the containment to mitigate severe accident consequences introduces the potential for accident management and recovery and this in turn points towards a range of new containment systems and concepts. PSA helps in judging these possibilities and in forming policies and procedures for accident management. It is perhaps in accident management that severe accident containment performance will be most beneficial in the future, and where additional effort in containment analysis will be focused

  7. Analysis of fuel-handling incidents (safety analysis detailed report no. 5). PEC Brasimone reactor design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-15

    The features covered by this report deal with the equipment and cells in which the handling, examination, measurement, conditioning and storage of core elements are carried out. The operations covered range from the receiving of new element shipments to their insertion in the vessel (excluding handling inside the vessel itself, which is covered in report no. 2) and removal of the spent-elements from the vessel, transfer to their final storage and their ultimate loading into containers for transport outside the plant. The incident analysis along the path of the spent fuel was conducted with the same method adopted for other plant systems. It is treated separately here because the operation of the handling system is practically autonomous from reactor operation.

  8. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    2004-01-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  9. Extreme load alleviation using industrial implementation of active trailing edge flaps in a full design load basis

    DEFF Research Database (Denmark)

    Barlas, Athanasios; Pettas, Vasilis; Gertz, Drew Patrick

    2016-01-01

    The application of active trailing edge flaps in an industrial oriented implementation is evaluated in terms of capability of alleviating design extreme loads. A flap system with basic control functionality is implemented and tested in a realistic full Design Load Basis (DLB) for the DTU 10MW...

  10. ALWR utility requirements - A technical basis for updated emergency planning

    International Nuclear Information System (INIS)

    Leaver, David E.W.; DeVine, John C. Jr.; Santucci, Joseph

    2004-01-01

    U.S. utilities, with substantial support from international utilities, are developing a comprehensive set of design requirements in the form of a Utility Requirements Document (URD) as part of an industry wide effort to establish a technical foundation for the next generation of light water reactors. A key aspect of the URD is a set of severe accident-related design requirements which have been developed to provide a technical basis for updated emergency planning for the ALWR. The technical basis includes design criteria for containment performance and offsite dose during severe accident conditions. An ALWR emergency planning concept is being developed which reflects this severe accident capability. The main conclusion from this work is that the likelihood and consequences of a severe accident for an ALWR are fundamentally different from that assumed in the technical basis for existing emergency planning requirements, at least in the U.S. The current technical understanding of severe accident risk is greatly improved compared to that available when the existing U.S. emergency planning requirements were established nearly 15 years ago, and the emerging ALWR designs have superior core damage prevention and severe accident mitigation capability. Thus, it is reasonable and prudent to reflect this design capability in the emergency planning requirements for the ALWR. (author)

  11. Case study on chemical plant accidents for flow-sheet design of the HTTR-IS system

    International Nuclear Information System (INIS)

    Homma, Hiroyuki; Sato, Hiroyuki; Kasahara, Seiji; Hara, Teruo; Kato, Ryoma; Sakaba, Nariaki; Ohashi, Hirofumi

    2007-02-01

    At the present time, we are alarmed by depletion of fossil energy and adverse effect of rapid increase in fossil fuel burning on environment such as climate changes and acid rain, because our lives depend still heavily upon fossil energy. It is thus widely recognized that hydrogen is one of important future energy carriers in which it is used without emission of carbon dioxide greenhouse gas and atmospheric pollutants and that hydrogen demand will increase greatly as fuel cells are developed and applied widely in the near future. To meet massive demand of hydrogen, hydrogen production from water utilizing nuclear, especially by thermochemical water-splitting Iodine-Sulphur (IS) process utilizing heat from High-Temperature Gas-cooled Reactors (HTGRs), offers one of the most attractive zero-emission energy strategies and the only one practical on a substantial scale. However, to establish a technology based for the HTGR hydrogen production by the IS process, we should close several technology gaps through R and D with the High-Temperature Engineering Test Reactor (HTTR), which is the only Japanese HTGR built and operated at the Oarai Research and Development Centre of Japan Atomic Energy Agency (JAEA). We have launched design studies of the IS process hydrogen production system coupled with the HTTR (HTTR-IS system) to demonstrate HTGR hydrogen production. In designing the HTTR-IS system, it is necessary to consider preventive and breakdown maintenance against accidents occurred in the IS process as a chemical plant. This report describes case study on chemical plant accidents relating to the IS process plant and shows a proposal of accident protection measures based on above case study, which is necessary for flow-sheet design of the HTTR-IS system. (author)

  12. Nuclear accidents and epidemiology

    International Nuclear Information System (INIS)

    1987-01-01

    A consultation on epidemiology related to the Chernobyl accident was held in Copenhagen in May 1987 as a basis for concerted action. This was followed by a joint IAEA/WHO workshop in Vienna, which reviewed appropriate methodologies for possible long-term effects of radiation following nuclear accidents. The reports of these two meetings are included in this volume, and cover the subjects: 1) Epidemiology related to the Chernobyl nuclear accident. 2) Appropriate methodologies for studying possible long-term effects of radiation on individuals exposed in a nuclear accident. Figs and tabs

  13. Design parameters and testing techniques for criticality accident detection systems used in various nuclear establishments - a review

    International Nuclear Information System (INIS)

    Janardhanan, S.; Krishnamony, S.; Krishnamurthi, T.N.; Gopalan, C.S.

    1981-01-01

    Accidental criticality excursion is a potential hazard in operations involving fissile material. In this review paper, design criteria for criticality detection systems, associated requirements for reliable functioning of the instrument and recent advances in the field are discussed. Systems based on integrated dose and rate of change of dose rate concepts are explained. A criticality accident simulator using a pneumatically driven 60 Co source for testing the detector is described. The paper also discusses the relative advantages of gamma and neutron sensing devices. (author)

  14. Design parameters and testing techniques for criticality accident detection systems used in various nuclear establishments - a review

    Energy Technology Data Exchange (ETDEWEB)

    Janardhanan, S.; Krishnamony, S.; Krishnamurthi, T.N.; Gopalan, C.S. (Bhabha Atomic Research Centre, Bombay (India). Health Physics Div.)

    Accidental criticality excursion is a potential hazard in operations involving fissile material. In this review paper, design criteria for criticality detection systems, associated requirements for reliable functioning of the instrument and recent advances in the field are discussed. Systems based on integrated dose and rate of change of dose rate concepts are explained. A criticality accident simulator using a pneumatically driven /sup 60/Co source for testing the detector is described. The paper also discusses the relative advantages of gamma and neutron sensing devices.

  15. A contribution to severe accident monitoring: Level measurement of the Incontainment Refueling Water Storage Tank (IRWST), design and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Schumilov, A.; Weber, P.; Esteves, S.

    2012-07-01

    A level measurement sensor for monitoring the water level in the in-containment refueling water storage tank (IRWST) of the EPRTM (generation 3+ pressurized water reactor) during leakage and severe accidents has been developed by AREVA. The development has been accompanied by many functional and material analyses as well as tests to assure the resistivity under extreme conditions, such as high irradiation dose of 5 MGy, increased temperature up to 160 degree centigrade in conjunction with saturated steam conditions. Moreover, the sensor has been designed and experimentally verified to resist the impact of seismic events and airplane crashes as well.

  16. Accident analysis for nuclear power plants

    International Nuclear Information System (INIS)

    2002-01-01

    calculation results. This safety report also discusses various factors that need to be considered to ensure that the accident analysis is of an acceptable quality. The report is intended for use primarily by analyses coordinating, performing or reviewing accident analyses for NPPs, on both the utility and regulatory sides. The report will also be of use as a background document for relevant IAEA activities, such as training courses and workshops. While the main body of the report does not focus exclusively on a single reactor type, the examples provided in the annexes are related mostly to the accident analysis of NPPs with pressurized water reactors. The report: Applies to both NPPs being built and operating plants; deals with internal events in reactors or in their associated process systems; thus the emphasis is on the physical transient behaviour of reactors and their systems, including reactor containment; discusses both best estimate and conservative accident analyses; covers design basis accidents as well as beyond design basis accidents, although the design basis accidents are covered in greater detail; focuses on thermohydraulic aspects of safety analysis; neutronic, structural and radiological aspects are also covered to some extent; covers the course of an accident from the initiating event up to source term estimation. The main body of the report is intended to be as generally applicable as possible to all reactor types

  17. Prevention and investigations of core degradation in case of beyond design accidents of the 2400 MWTH gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Bertrand, F.; Gatin, V.; Bentivoglio, F.; Gueneau, C.

    2011-01-01

    The present paper deals with studies carried out to assess the ability of the core of the Gas Fast Reactor (GFR) to withstand beyond design accidents. The work presented here is aimed at simulating the behaviour of this core by using analytical models whose input parameters are calculated with the CATHARE2 code. Among possible severe accident initiators, the Unprotected Loss Of Coolant Accident (ULOCA of 3 Inches diameter) is investigated in detail in the paper with CATHARE2. Additionally, a simplified pessimistic assessment of the effect of a postulated power excursion that could result from the failure of prevention provisions is presented. (author)

  18. E-learning course: Basis of Harvest and Preservation of Tissues – design and initial experience

    Directory of Open Access Journals (Sweden)

    Pavel Měřička

    2014-05-01

    Full Text Available Background: The design and initial experience with the e-learning course “Basis of Harvest and Preservation of Tissues” used as a support of an elective subject is presented. The aim of the e-learning course was to enable the students to learn the theoretical principles of the subject individually and to present the gained knowledge at the final seminar. Methods: All functions of the course were operated in Moodle, local application of the Charles University in Prague, Faculty of Medicine in Hradec Králové. The course was divided into 3 main topics corresponding with topics of lectures: 1. Principles of tissue and organ donation, 2. Low temperature preservation of cells, tissues and organs, 3. Quality and safety assurance in practice of tissue and procurement establishments. A test consisting of 5 questions selected randomly from the bank of questions followed each topic. If the student answers correctly at least 3 questions he is allowed to pass to the next topic. The fourth topic “Basic processes in the tissue establishment and principles of their validation” was added into the electronic version as a tool for repeating and improving of knowledge. The fifth topic was represented by a database for uploading theses presented by students at the final seminar. The final test consisted of 15 questions (5 ones from each basic topic. It was necessary to answer correctly at least 10 questions to receive a certificate of completing the course. Results: The course was put into operation during the summer term of the academic year 2012/2013. To the date 15 of September the total of 23 students enrolled (17, i.e. all students of the elective subject in the Czech version, 2 students of this subject in the English version, 2 postgraduate students and 2 medical doctors. All enrolled students used the course for on-line learning, downloading, or printing course study materials. All undergraduate students were obliged to use it for preparation

  19. Annual meeting on nuclear technology 1982. Technical meeting: Possibilities and effects of serious reactor accidents

    International Nuclear Information System (INIS)

    1982-01-01

    A critical examination of the forecast of a design basis accident, the view of the Sandia National Laboratory on the probability of a steam explosion after a core meltdown accident is comparison with WASH-1400, the possibilities of interactions with the containment structure and fission product release, as well as the influences for the assessment of risk in Germany taken from the analysis of core meltdown accidents are dealt with in these papers. (DG) [de

  20. Evaluation of heatup and recovery in a loss of feedwater accident with multiple failure

    International Nuclear Information System (INIS)

    Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung

    1991-01-01

    A loss of feedwater accident with multiple failure has been studied in order to identify the potential severity of the accident when compared with the design basis accident in PWR. The PCS heatup and recovery mode in a LOFA with multiple failure was evaluated using the LOFT L9-1/L3-3 experiment. From experimental result, 4 separable subphase were identified and the associated phenomena were also addressed

  1. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Morten; Zhang Xinxin [Harbin Engineering University, Harbin (China)

    2014-08-15

    The paper investigate applications of functional modeling for accident management in complex industrial plant with special reference to nuclear power production. Main applications for information sharing among decision makers and decision support are identified. An overview of Multilevel Flow Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented.

  2. Development of Accident Scenario for Interim Spent Fuel Storage Facility Based on Fukushima Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongjin; Choi, Kwangsoon; Yoon, Hyungjoon; Park, Jungsu [KEPCO-E and C, Yongin (Korea, Republic of)

    2014-05-15

    700 MTU of spent nuclear fuel is discharged from nuclear fleet every year and spent fuel storage is currently 70.9% full. The on-site wet type spent fuel storage pool of each NPP(nuclear power plants) in Korea will shortly exceed its storage limit. Backdrop, the Korean government has rolled out a plan to construct an interim spent fuel storage facility by 2024. However, the type of interim spent fuel storage facility has not been decided yet in detail. The Fukushima accident has resulted in more stringent requirements for nuclear facilities in case of beyond design basis accidents. Therefore, there has been growing demand for developing scenario on interim storage facility to prepare for beyond design basis accidents and conducting dose assessment based on the scenario to verify the safety of each type of storage.

  3. Structural Basis for Escape of Human Astrovirus from Antibody Neutralization: Broad Implications for Rational Vaccine Design

    Energy Technology Data Exchange (ETDEWEB)

    Bogdanoff, Walter A.; Perez, Edmundo I.; López, Tomás; Arias, Carlos F.; DuBois, Rebecca M. (UNAM-Mexico); (UCSC)

    2017-10-25

    ABSTRACT

    Human astroviruses are recognized as a leading cause of viral diarrhea worldwide in children, immunocompromised patients, and the elderly. There are currently no vaccines available to prevent astrovirus infection; however, antibodies developed by healthy individuals during previous infection correlate with protection from reinfection, suggesting that an effective vaccine could be developed. In this study, we investigated the molecular mechanism by which several strains of human astrovirus serotype 2 (HAstV-2) are resistant to the potent HAstV-2-neutralizing monoclonal antibody PL-2 (MAb PL-2). Sequencing of the HAstV-2 capsid genes reveals mutations in the PL-2 epitope within the capsid's spike domain. To understand the molecular basis for resistance from MAb PL-2 neutralization, we determined the 1.35-Å-resolution crystal structure of the capsid spike from one of these HAstV-2 strains. Our structure reveals a dramatic conformational change in a loop within the PL-2 epitope due to a serine-to-proline mutation, locking the loop in a conformation that sterically blocks binding and neutralization by MAb PL-2. We show that mutation to serine permits loop flexibility and recovers MAb PL-2 binding. Importantly, we find that HAstV-2 capsid spike containing a serine in this loop is immunogenic and elicits antibodies that neutralize all HAstV-2 strains. Taken together, our results have broad implications for rational selection of vaccine strains that do not contain prolines in antigenic loops, so as to elicit antibodies against diverse loop conformations.

    IMPORTANCEHuman astroviruses (HAstVs) infect nearly every person in the world during childhood and cause diarrhea, vomiting, and fever. In this study, we investigated how several strains of HAstV are resistant to a virus-neutralizing monoclonal antibody. We determined the crystal structure of the capsid protein spike domain from one of these HAstV strains and found that

  4. Dose evaluation on the basis of {sup 24}Na activity in the human body for the criticality accident at JCO Tokai nuclear fuel processing plant

    Energy Technology Data Exchange (ETDEWEB)

    Momose, T.; Tsujimura, N.; Tasaki, T.; Kanai, K.; Hayashi, N.; Shinohara, K. [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2001-11-01

    {sup 24}Na in the human body, activated by neutrons emitted at the JCO criticality accident, was observed for 62 subjects, where 148 subjects were measured by the whole body counter of JNC Tokai Works. The 148 subjects, including JCO employees and the contractors, residents neighboring the site and emergency service officers, were measured by the whole-body counter. The neutron-energy spectrum around the facility was calculated using neutron transport codes (ANISN and MCNP), and the relation between an amount of activated sodium in human body and neutron dose was evaluated from the calculated neutron energy spectrum and theoretical neutron capture probability by the human body. The maximum {sup 24}Na activity in the body was 7.7 kBq (83 Bq({sup 24}Na)/g({sup 23}Na)) and the relevant effective dose equivalent was 47 mSv. (author)

  5. Dose evaluation on the basis of {sup 24}Na activity in the human body for the criticality accident at JCO Tokai nuclear fuel processing plant

    Energy Technology Data Exchange (ETDEWEB)

    Momose, T.; Tsujimura, N.; Tasaki, T.; Kanai, K.; Hayashi, N.; Shinohara, K. [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2000-07-01

    Sodium-24({sup 24}Na) generated in human body due to neutron activation was measured by whole body counter (WBC) in JNC Tokai works. Total 148 persons (JCO employees and contractor, public member, fire fighters, etc.) were measured and {sup 24}Na was detected in the 62 persons. Neutron energy spectrum around the facility was calculated using ANISN and MCNP code and estimated mean capture probability {xi} of neutron for human body at this accident was around 0.25-0.28 at any distance from the center of the precipitation tank. Effective dose equivalent for the 62 persons were estimated based on the calculated conversion factors from {sup 24}Na specific activity to neutron dose. Maximum {sup 24}Na activity was 7.7 kBq (83 Bq({sup 24}Na)/g({sup 23}Na)) in total body and the evaluated effective dose equivalent was 47 mSv. (author)

  6. Dose evaluation on the basis of 24Na activity in the human body for the criticality accident at JCO Tokai nuclear fuel processing plant

    International Nuclear Information System (INIS)

    Momose, T.; Tsujimura, N.; Tasaki, T.; Kanai, K.; Hayashi, N.; Shinohara, K.

    2001-01-01

    24 Na in the human body, activated by neutrons emitted at the JCO criticality accident, was observed for 62 subjects, where 148 subjects were measured by the whole body counter of JNC Tokai Works. The 148 subjects, including JCO employees and the contractors, residents neighboring the site and emergency service officers, were measured by the whole-body counter. The neutron-energy spectrum around the facility was calculated using neutron transport codes (ANISN and MCNP), and the relation between an amount of activated sodium in human body and neutron dose was evaluated from the calculated neutron energy spectrum and theoretical neutron capture probability by the human body. The maximum 24 Na activity in the body was 7.7 kBq (83 Bq( 24 Na)/g( 23 Na)) and the relevant effective dose equivalent was 47 mSv. (author)

  7. Implementation of an Industrial-Based Case Study as the Basis for a Design Project in an Introduction to Mechanical Design Course

    Science.gov (United States)

    Lackey, Ellen

    2011-01-01

    The purpose of this paper is to discuss the implementation of an industrial-based case study as the basis for a design project for the Spring 2009 Introduction to Mechanical Design Course at the University of Mississippi. Course surveys documented the lack of student exposure in classes to the types of projects typically experienced by engineers…

  8. The need to study of bounding accident in reprocessing plant

    International Nuclear Information System (INIS)

    Segawa, Satoshi; Fujita, Kunio

    2013-01-01

    There is a clear consensus that the severe accident corresponds to the core damage accident for power reactors. On the other hand, for FCFs, there is no clear consensus on what is the accident to assess the safety in the region of beyond design basis, or what is the accident which has very low probability but large consequence. The need to examine a bounding consequence of each type of accident is explained to advance the rationality of safety management and regulation and, as a result, to reinforce the safety of a reprocessing plant. The likelihood of occurrence of an accident causing a bounding consequence should correspond to that of a severe accident at a nuclear power plant. The bounding consequence will be derived using the deterministic method and sound engineering judgment supplemented by the probabilistic method. Once an agreement on such a concept is reached among regulators, operators and related experts it will help to provide a solid basis to ensure the safety of a reprocessing plant independent of that of a nuclear power plant. In this paper, we show a preliminary risk profile of RRP calculated by QSA (Quantitative Safety Assessment) which JNFL developed. The profile shows that bounding consequences of various accidents in a range of occurrence frequency corresponding to a severe accident at a nuclear power plant. And we find that the bounding consequence of high-level liquid waste boiling is the largest among all in this range. Therefore, the risk of this event is shown in this paper as an example. To build a common consensus about bounding accidents among concerned parties will encourage regulatory body to introduce such an idea for more effective regulation with scientific rationality. Additionally the study of bounding accidents can contribute to substantial development for accident management strategy as reprocessing operators. (authors)

  9. The assessment of environmental consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Beattie, J.R.

    1981-01-01

    Thorough measures are taken throughout all stages of design, construction and operation of nuclear power reactors, and therefore no accident producing any significant environmental impact is likely to occur. Nevertheless as a precaution, such accidents have been the subject of intensive scientific predictive studies. After a historical review of theoretical papers on reactor accidents and their imagined environmental impacts and of those accidents that have indeed occurred, this paper gives an outline of fission products or other radioactive substances that may or may not be released by an accident, and of their possible effects after dispersion in the atmosphere. This general introduction is followed by sections describing what are sometimes called 'design basis accidents' for four of the main reactor types (magnox, AGR, PWR and CDFR), the precautions against these accidents and the probable degree of environmental impact likely. The paper concludes with a reference to those very low probability accidents which might have more serious environmental impacts, and proceeds from there to show that both the individual and community risks from such accidents are numerically moderate compared to other risks apparently accepted by society. A brief reflection on the relevance of numerical values and perceived risk concludes the paper. (author)

  10. Analytical method and result of radiation exposure for depressurization accident of HTTR

    International Nuclear Information System (INIS)

    Sawa, K.; Shiozawa, S.; Mikami, H.

    1990-01-01

    The Japan Atomic Energy Research Institute (JAERI) is now proceeding with the construction design of the High Temperature Engineering Test Reactor (HTTR). Since the HTTR has some characteristics different from LWRs, analytical method of radiation exposure in accidents provided for LWRs can not be applied directly. This paper describes the analytical method of radiation exposure developed by JAERI for the depressurization accident, which is the severest accident in respect to radiation exposure among the design basis accidents of the HTTR. The result is also described in this paper

  11. Extreme load alleviation using industrial implementation of active trailing edge flaps in a full design load basis

    OpenAIRE

    Barlas, Athanasios; Pettas, Vasilis; Gertz, Drew Patrick; Aagaard Madsen , Helge

    2016-01-01

    The application of active trailing edge flaps in an industrial oriented implementation is evaluated in terms of capability of alleviating design extreme loads. A flap system with basic control functionality is implemented and tested in a realistic full Design Load Basis (DLB) for the DTU 10MW Reference Wind Turbine (RWT) model and for an upscaled rotor version in DTU's aeroelastic code HAWC2. The flap system implementation shows considerable potential in reducing extreme loads in components o...

  12. Use of analytical aids for accident management

    International Nuclear Information System (INIS)

    Ward, L.W.

    1991-01-01

    The use of analytical aids by utility technical support teams can enhance the staff's ability to manage accidents. Since instrumentation is exposed to environments beyond design-basis conditions, instruments may provide ambiguous information or may even fail. While it is most likely that many instruments will remain operable, their ability to provide unambiguous information needed for the management of beyond-design-basis events and severe accidents is questionable. Furthermore, given these limitation in instrumentation, the need to ascertain and confirm current plant status and forecast future behavior to effectively manage accidents at nuclear facilities requires a computational capability to simulate the thermal and hydraulic behavior in the primary, secondary, and containment systems. With the need to extend the current preventive approach in accident management to include mitigative actions, analytical aids could be used to further enhance the current capabilities at nuclear facilities. This need for computational or analytical aids is supported based on a review of the candidate accident management strategies discussed in NUREG/CR-5474. Based on the review of the NUREG/CR-5474 strategies, two major analytical aids are considered necessary to support the implementation and monitoring of many of the strategies in this document. These analytical aids include (1) An analytical aid to provide reactor coolant and secondary system behavior under LOCA conditions. (2) An analytical aid to predict containment pressure and temperature response with a steam, air, and noncondensable gas mixture present

  13. Technical basis document for external events

    International Nuclear Information System (INIS)

    OBERG, B.D.

    2003-01-01

    This document supports the Tank Farms Documented Safety Analysis and presents the technical basis for the FR-equencies of externally initiated accidents. The consequences of externally initiated events are discussed in other documents that correspond to the accident that was caused by the external event. The external events include aircraft crash, vehicle accident, range fire, and rail accident

  14. Technical basis for the ITER final design report, cost review and safety analysis (FDR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement. Refs, figs, tabs

  15. Technical basis for the ITER final design report, cost review and safety analysis (FDR)

    International Nuclear Information System (INIS)

    1998-01-01

    The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement

  16. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided

  17. Severe accidents and operator training - discussion of potential issues

    International Nuclear Information System (INIS)

    Vidard, Michel

    1997-01-01

    R and D programs developed throughout the world allowed significant progress in the understanding of physical phenomena and Severe Accident Management (SAM) programs started in many OECD countries. Basically, the common denominator to all these SAM programs was to provide utility operators with procedures or guidelines allowing to deal with complex situations not formally considered in the Design Basis, including accidents where a significant portion of the core had molten. These SAM procedures or guidelines complement the traditional accident management procedures (event, symptom or physical-state oriented) and should allow operators to deal with a reasonably bounding set of situations. Dealing with operator or crisis team training, it was recognized that training would be beneficial but that training programs were lagging, i.e. though training sessions were either organized or contemplated after implementation of SAM programs, they seemed to be somewhat different from more traditional training sessions on Accident Management. After some explanations on the differences between Design Basis Accidents (DBAs) and Beyond Design Basis Accidents (BDBAs), this paper underlines some potential difficulties for training operators and discuss problems to be addressed by organisms contemplating SAM training sessions consistent with similar activities for less complex events

  18. Multi dimensional analysis of Design Basis Events using MARS-LMR

    International Nuclear Information System (INIS)

    Woo, Seung Min; Chang, Soon Heung

    2012-01-01

    Highlights: ► The one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions. ► The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. ► The difference of the sodium flow pattern due to structure effect in the hot pool and mass flow rates in the core lead the different sodium temperature and temperature history under transient condition. - Abstract: KALIMER-600 (Korea Advanced Liquid Metal Reactor), which is a pool type SFR (Sodium-cooled Fast Reactor), was developed by KAERI (Korea Atomic Energy Research Institute). DBE (Design Basis Events) for KALIMER-600 has been analyzed in the one dimension. In this study, the one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions, such as UIS (Upper Internal Structure), IHX (Intermediate Heat eXchanger), DHX (Decay Heat eXchanger), and pump. The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. First, the results in normal operation condition show the good agreement between the one and multi-dimensional analysis. However, according to the sodium temperatures of the core inlet, outlet, the fuel central line, cladding and PDRC (Passive Decay heat Removal Circuit), the temperatures of the one dimensional analysis are generally higher than the multi-dimensional analysis in conditions except the normal operation state, and the PDRC operation time in the one dimensional analysis is generally longer than

  19. Conceptual design of emergency communication system to cope with severe accidents in NPPs and its performance evaluation

    International Nuclear Information System (INIS)

    Son, Kwang Seop; Kim, Chang Hwoi; Kang, Hyun Gook

    2015-01-01

    Highlights: • The emergency communication system requires the performances of the throughput of 1 Mbps, BER of 10 −6 and network configuration of 1:12 communication. • The emergency communication system consists of the terrestrial communication and satellite communication system. • In the terrestrial communication system, at least two wireless repeaters are needed to secure LOS and the throughput and delay time are 16 Mbps and 16 ms, respectively. • In the satellite communication system, DSSS and FDMA are used and the fade margin range is from 1.3 to 16 dB. - Abstract: The Fukushima accident induced by the great earthquake and tsunami reveals the vulnerability of I and C System. In the severe environment, the normal I and C system did not work properly and results in false information about the internal situation in NPP. Eventually the accident was not properly handled at the early stage. Therefore advanced emergency response system using a wireless channel is necessary to cope with the severe accident. In this paper, we introduce the ERS consisting of the HMS and MCS the ECS linking the HMS with MCS and the performance requirement of the ECS is analyzed. The ECS satisfying the requirement is designed conceptually and the performance of the ECS is evaluated through analysis and simulator. To secure a reliable and diverse configuration, the ECS is configured as the dual system which consists of the terrestrial communication and satellite communication. The terrestrial communication system is designed based on the IEEE 802.11. Analyzed performance results prove that the performance requirement can be sufficiently achieved. But if the scalability of data capacity is considered later, use of the advanced 802.11 standard such as 802.11n and multiple signal paths between the HMS and MCS are necessary. In the satellite communication system, the FDMA is used in the status link and the DSSS is used in the control link. The network supporting various data rates is

  20. The reactor accident in Fukushima Daiichi. The consequence of design deficiencies and inadequate safety engineering; Der Reaktorunfall in Fukushima Daiichi. Folge fehlerhafter Auslegung und unzureichender Sicherheitstechnik

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-03-15

    The reactor accident in Fukushima Daiichi is discussed in the frame of design deficiencies and inadequate safety engineering. The progress of the accident as consequence of the earthquake and the tsunami is described. The radiological situation for the public is supposed to be blow the dose limit of 20 mSv/year. The WHO and UNSCEAR (United Nations Scientific Committee on the Effects of Atomic radiation) did not observe acute radiation injuries. The Japanese authorities have classified the accident to 7 of the INES scale. The German Atomforum e.V. considers the safety engineering of German NPPs to be superior to the Japanese situation due to higher emergency energy supply, extensive measures to reduce the hydrogen accumulation and mitigating measures for the accident management. German NPPS are considered highly robust as the EU stress tests have shown.

  1. Technical basis for the ITER detailed design report, cost review and safety analysis (DDR)

    International Nuclear Information System (INIS)

    1997-01-01

    The ITER Detailed Design Report (DDR), Cost Review and Safety Analysis is the 3rd major milestone representing the progress made in the ITER Engineering Design Activities. With the approval of the Interim Design Report (IDR), it has been possible to freeze the main concepts and system approaches for ITER and to develop the design in more detail for the individual components and sub-systems. This report, although designed to be fully understandable as a separate document, focusses particularly on the main changes since the IDR

  2. Technical basis for the ITER detailed design report, cost review and safety analysis (DDR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The ITER Detailed Design Report (DDR), Cost Review and Safety Analysis is the 3rd major milestone representing the progress made in the ITER Engineering Design Activities. With the approval of the Interim Design Report (IDR), it has been possible to freeze the main concepts and system approaches for ITER and to develop the design in more detail for the individual components and sub-systems. This report, although designed to be fully understandable as a separate document, focusses particularly on the main changes since the IDR. Refs, figs, tabs

  3. An approach to estimating radiological risk of offsite release from a design basis earthquake for the Process Experimental Pilot Plant (PREPP)

    International Nuclear Information System (INIS)

    Lucero, V.; Meale, B.M.; Reny, D.A.; Brown, A.N.

    1990-09-01

    In compliance with Department of Energy (DOE) Order 6430.1A, a seismic analysis was performed on DOE's Process Experimental Pilot Plant (PREPP), a facility for processing low-level and transuranic (TRU) waste. Because no hazard curves were available for the Idaho National Engineering Laboratory (INEL), DOE guidelines were used to estimate the frequency for the specified design-basis earthquake (DBE). A dynamic structural analysis of the building was performed, using the DBE parameters, followed by a probabilistic risk assessment (PRA). For the PRA, a functional organization of the facility equipment was effected so that top events for a representative event tree model could be determined. Building response spectra (calculated from the structural analysis), in conjunction with generic fragility data, were used to generate fragility curves for the PREPP equipment. Using these curves, failure probabilities for each top event were calculated. These probabilities were integrated into the event tree model, and accident sequences and respective probabilities were calculated through quantification. By combining the sequences failure probabilities with a transport analysis of the estimated airborne source term from a DBE, onsite and offsite consequences were calculated. The results of the comprehensive analysis substantiated the ability of the PREPP facility to withstand a DBE with negligible consequence (i.e., estimated release was within personnel and environmental dose guidelines). 57 refs., 19 figs., 20 tabs

  4. An approach to estimating radiological risk of offsite release from a design basis earthquake for the Process Experimental Pilot Plant (PREPP)

    Energy Technology Data Exchange (ETDEWEB)

    Lucero, V.; Meale, B.M.; Reny, D.A.; Brown, A.N.

    1990-09-01

    In compliance with Department of Energy (DOE) Order 6430.1A, a seismic analysis was performed on DOE's Process Experimental Pilot Plant (PREPP), a facility for processing low-level and transuranic (TRU) waste. Because no hazard curves were available for the Idaho National Engineering Laboratory (INEL), DOE guidelines were used to estimate the frequency for the specified design-basis earthquake (DBE). A dynamic structural analysis of the building was performed, using the DBE parameters, followed by a probabilistic risk assessment (PRA). For the PRA, a functional organization of the facility equipment was effected so that top events for a representative event tree model could be determined. Building response spectra (calculated from the structural analysis), in conjunction with generic fragility data, were used to generate fragility curves for the PREPP equipment. Using these curves, failure probabilities for each top event were calculated. These probabilities were integrated into the event tree model, and accident sequences and respective probabilities were calculated through quantification. By combining the sequences failure probabilities with a transport analysis of the estimated airborne source term from a DBE, onsite and offsite consequences were calculated. The results of the comprehensive analysis substantiated the ability of the PREPP facility to withstand a DBE with negligible consequence (i.e., estimated release was within personnel and environmental dose guidelines). 57 refs., 19 figs., 20 tabs.

  5. A model for personal life project design on the basis of vocational guidance

    Directory of Open Access Journals (Sweden)

    Isaac Geovanni Mendoza Cedeño

    2015-09-01

    Full Text Available This article is intended to propose a model for developing vocational education personal life projects and its corresponding theoretical foundations. Therefore, the argument is undertaken on the model developed from philosophical views, epistemological, pedagogical, sociological, and psychological axiological, and a model is provided as a basis for the development of vocational education strategy for personal life project development contributing effectively to the development of responsible autonomy and high school students.

  6. Approach to developing a ground-motion design basis for facilities important to safety at Yucca Mountain

    International Nuclear Information System (INIS)

    King, J.L.

    1990-01-01

    This paper discusses a methodology for developing a ground-motion design basis for prospective facilities at Yucca Mountain that are important to safety. The methodology utilizes a guasi-deterministic construct called the 10,000-year cumulative-slip earthquake that is designed to provide a conservative, robust, and reproducible estimate of ground motion that has a one-in-ten chance of occurring during the preclosure period. This estimate is intended to define a ground-motion level for which the seismic design would ensure minimal disruption to operations engineering analyses to ensure safe performance are included

  7. Development of Pupils Picture Aesthetic Competences on the Basis of IT-didactic Designs of Digital Picture Production

    DEFF Research Database (Denmark)

    Rasmussen, Helle

    : The research method refers to Design Based Research, since the project is based on a design theoretical view of learning. (Cobb et. All 2003, Van den Akker 2006, Collins 2004). Learning is here to be understood as “a sign producing activity in a specific situation within an institutional framing”, which makes...... Education” (English Title), The Danish University of Education Cobb, P. et al. (2003): “Design Experiments in Educational Research” in “Educational Researcher”, vol. 32, no. 1. Collins, Allan et. al. (2004): “Design Research: Theoretical and Metodological Issuses” in “Journal of the Learning Sciences”, Vol...... Competences on the Basis of IT-didactic Designs of Digital Picture Production Proposal information: The topic for this presentation is an ongoing investigation of the connection between the learning outcome of digital picture production and IT-didactic designs, and it refers to a Ph.D.-project in progress...

  8. Source term analyses under severe accidents for KNGR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong

    2001-03-01

    In this study, in-containment source term for LOFW (Loss of Feed Water), which has appeared the most frequent core melt accident, is calculated and compared with NUREG-1465 source term. This study provides not only new source term data using MELCOR1.8.4 and its state-of-the-art models but also evaluating basis of KNGR design and its mitigation capability under severe accidents. As the selected accident is identical with LOFW-S17, which has been analyzed using MAAP by KEPCO with only difference of 2 SITs, mutual comparison of the results is especially expected.

  9. SWR-1000 concept on control of severe accidents

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1998-01-01

    It is essential for the SWR-1000 probabilistic safety concept to consider the results from experiments and reliability system failure within the probabilistic safety analyses for passive systems. Active and passive safety features together reduce the probability of the occurrence of beyond design basis accidents in order to limit their consequences in accordance with the German law. As a reference case we analyzed the most probable core melt accident sequence with a very conservative assumption. An initial event, stuck open of safety and relief valves without the probability of active and passive feeding systems of the pressure vessel, was considered. Other sequences of the loss of coolant accidents lead to lower probability

  10. Design basis for resistance to shock and vibration of radioactive material packages greater than one ton in truck transport (draft standard for trial use and comment)

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This standard specifies minimum design values for shock and vibration in highway transport, by truck or by tractor-trailer combination, for fuel and irradiation experiments when package weight exceeds one ton. Shock values correspond to normal transport over rough roads and to minor accidents such as backing into a loading dock. Vibration values correspond to normal transport; any large-amplitude vibration resulting from rough road conditions or a minor accident is treated as shock. This standard includes recommended methods of application to the design of packaging and tiedown systems

  11. A design basis for the development of advanced CANDU control centres

    International Nuclear Information System (INIS)

    Feher, M.P.; Davey, E.C.; Lupton, L.R.

    1995-01-01

    The basic design for current CANDU control centres was established in the early 1970's. Plants constructed since then have, for the most part, retained the same basic design. Several factors have led to the need to re-examine CANDU control centre design for plants to be built beyond the year 2000. These factors include the changing roles and responsibilities for the operations staff, an improved understanding of operational issues associated with supervisory control, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and marketing pressures. This paper describes the proposed design bases for the development of advanced control centres to be implemented in CANDU plants beyond the year 2000. Four areas have been defined covering design goals, design principles, operational bases, and plant functional bases. (author)

  12. A design basis for the development of advanced CANDU control centres

    Energy Technology Data Exchange (ETDEWEB)

    Feher, M P; Davey, E C; Lupton, L R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The basic design for current CANDU control centres was established in the early 1970`s. Plants constructed since then have, for the most part, retained the same basic design. Several factors have led to the need to re-examine CANDU control centre design for plants to be built beyond the year 2000. These factors include the changing roles and responsibilities for the operations staff, an improved understanding of operational issues associated with supervisory control, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and marketing pressures. This paper describes the proposed design bases for the development of advanced control centres to be implemented in CANDU plants beyond the year 2000. Four areas have been defined covering design goals, design principles, operational bases, and plant functional bases. (author).

  13. MDEP Design-Specific Common Position CP-APR1400WG-01. Common position addressing Fukushima Daiichi nuclear power plant accident

    International Nuclear Information System (INIS)

    2016-05-01

    The MDEP APR1400 Working Group (APR1400WG) members consist of members from Republic of Korea, United Arab Emirates, and the United States. A main objectives of MDEP is to encourage convergence of code, standard and safety goals with exploring the opportunities for harmonization of regulatory practice and cooperation on safety review of APR-1400 specific designs. This common position addressing is aimed at sharing knowledge, information and experience on safety improvement related to lessons learned from the Fukushima Daiichi NPP Accident or Fukushima Daiichi NPP Accident-related issues amongst APR-1400 WG member states to achieve the MEDP goal. Because not all of these Regulators have completed the regulatory review of their APR1400 applications yet, this paper identifies common preliminary approaches to address potential safety improvements for APR1400 plants, as well as common general expectations for new nuclear power plants, as related to lessons learned from the Fukushima Daiichi NPP Accident or Fukushima Daiichi NPP Accident-related issues. While some asymmetry exists among those of three Regulators in terms of design, regulatory practice and licensing milestone sharing information and common understanding on post-Fukushima Daiichi NPP Accident enhancement would be promote resilient design for countering beyond design extreme external event like Fukushima Daiichi NPP nuclear disaster. This common position paper aims at identifying characteristics of post-Fukushima Daiichi NPP Accident enhancements putting in place by each country and setting common position to achieve balanced and harmonized APR-1400 design. After the safety reviews of the APR1400 design applications that are currently in review are completed, the regulators will update this paper to reflect their safety conclusions regarding the APR1400 design and how the design could be enhanced to address Fukushima Daiichi NPP Accident-related issues. The common preliminary approaches are organised into

  14. Use of COSYMA in the design of epidemiological studies in case of radiological accidents

    International Nuclear Information System (INIS)

    Mansoux, H.; Verger, P.; Thomassin, A.

    1996-01-01

    COSYMA is a powerful code for the assessment of a wide range of accident consequences. It has been satisfactory because of the possibility to get access to the code itself and to change options like release duration or outputs. COSYMA is a probabilistic code, but, in this application, it has been used in a deterministic way by extracting the results for one weather sequence only. PWR source terms were constructed using the inventory of the COSYMA users intercomparison exercise and release fractions for French nuclear reactors S1, S2, S3. For the reprocessing plant, two source terms were chosen arbitrarily, based on the actual radioactive materials present in the plant and on the radioactive decay of reprocessed PWR fuel (1g Plutonium and 0.01g Curium releases). For the transportation scenarios, several source terms still need to be defined, according to the materials transported, the modes of transportation, and the types of packaging. The post-accidental scenario has identical main options for all installations and source terms. The ingestion pathway was not treated in this study because of its unsatisfactory modeling in COSYMA. The intervention levels for the various countermeasures were drawn from ICRP 40 recommendations Both low and high levels were used for evacuation, sheltering and stable iodine prophylaxis For long term actions, only the low intervention level was applied. (author)

  15. Shielding design study of the demonstration fast breeder reactor. 2. Shielding design on the basis of the JASPER analysis

    International Nuclear Information System (INIS)

    Suzuoki, Zenro; Tabayashi, Masao; Handa, Hiroyuki; Iida, Masaaki; Takemura, Morio

    2000-01-01

    Conceptual shielding design has been performed for the Demonstration Fast Breeder Reactor (DFBR) to achieve further optimization and reduction of the plant construction cost. The design took into account its implications in overall plant configuration such as reduction of shields in the core, adoption of fission gas plenum in the lower portion of fuel assemblies, and adoption of gas expansion modules. Shielding criteria applied for the design are to secure fast neutron fluence on in-vessel structures as well as responses of the nuclear instrumentation system and to restrict secondary sodium activation. The design utilized the cross sections and the one- and two-dimensional discrete ordinates transport codes, whose verification had been performed by the JASPER experiment analysis. Correction factors yielded by the JASPER analysis were applied to the design calculations to obtain design values with improved accuracy. Design margins, which are defined by the ratios of the design criteria to the design values, were more than two for all shielding issues of interest, showing the adequacy of the shielding design of the DFBR. (author)

  16. THEORETICAL CONTENT FOUNDATION OF PREPARING TEACHERS OF LABOR TRAINING TO TEACHING BASIS OF DESIGN

    Directory of Open Access Journals (Sweden)

    Ihor Savenko

    2016-06-01

    Full Text Available In the article the basic principles of shaping the content of training future teachers of technology education to realize design activity in schools have been considered. Formation of the basic core knowledge, abilities and skills of future teachers should be provided in the fundamental scientific, general and cultural and professional training that is subjected to certain principles of pedagogical design. Design education as a powerful educative potential aimed at designing and providing humanitarian and cultural orientation of a student has been revealed. The design is a valued feature of professional education of future teachers of labor studies and promotes the development of a special type of culture and thinking that directs to educational values and determines the individual educational strategy of professional development.

  17. The detection of criticality accidents

    International Nuclear Information System (INIS)

    Prigent, R.; Renard, C.

    It is necessary to shield the personnel from the radiological consequences of a criticality accident. In the past ten years the study programmes have highlighted fresh data which have led to new thinking on the detection philosophy and as a consequence the design of detection equipment. Concurrently, new recommendations have been drawn up by the Safety Criticality Committee. The new detection equipment was developed by the CEA on the basis of the CRAC and SILENE experiments. Its industrialization was entrusted to the Intertechnique Company and the first network installed dates back to 1976. An examination is made of the problem of accident detection, dealing in turn with detection, the characteristics of the equipment and the installation rules. To clarify the various points discussed, a parallel has been drawn between the equipment existing up to 1975 and the new generation developed since then [fr

  18. Design-based measures to prevent damage in nuclear power plants due to incidents and accidents

    International Nuclear Information System (INIS)

    Feldmann, F.J.

    1983-01-01

    The question 'What are necessary provisions', or rather 'What level of safety is required' is used as an approach to the hitherto undefined legal concept of 'necessary provisions'. The Atomic Energy Act leaves the assessment of the types and extent of risk to the executive organs. So far, attempts to minimize the population risk have been made on the basis of emergency provisions against external influences, site selection and planning of emergency measures. (HP) [de

  19. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core

  20. Development of Emergency Operating Strategies for Beyond Design Basis External Event(BDBEE)s in Korean WH Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Duk-Joo; Lee, Seung-Chan; Sung, Je-Joong; Ha, Sang-Jun [KHNP CRI, Daejeon (Korea, Republic of); Hong, Soon-Joon; Hwang, Su-Hyun; Lee, Byung-Chul; Park, Kang-Min [FNC Tech. Co., Yongin (Korea, Republic of)

    2015-10-15

    Westinghouse developed and connected emergency operating procedures into a set of FLEX Support Guidelines(FSGs). This paper explains that Korean WH(Westinghouse) type nuclear power plants develop emergency operating strategies for ELAP(extended loss of all AC power), which include guidelines to use permanent and portable equipment as necessary to prevent core damage until AC power is restored from a reliable alternate source of AC power. The Korean emergency operating response strategies were developed to cope with a ELAP such as Fukushima event. The strategies include guidelines to prevent fuel damage using the FLEX equipment. Operators should take actions to prepare FLEX equipment within license basis SBO coping time. The loss of all AC power has been analyzed to identify the behavior of major NSSS process variables using RELAP computer code. The accident analysis showed that the plant does not result in fuel damage in 72 hours after an ELAP if operators take actions to cool RCS with opening of SG ADV in 5 gpm seal leak case. In this scenario, because ELAP is in process and all power cannot be used, operator should operate the FLEX equipment in order to actuate active equipment using the EOP fo SBO response. This strategy will prevent entering SAMG because this actions result in core cooling and stay in core exit temperature less than 650 .deg. C. Korean emergency operating guidelines(EOGs) will be developed using this strategies for response to the BDBEE.

  1. Instrumentation for the follow-up of severe accidents

    International Nuclear Information System (INIS)

    Munoz Sanchez, A.; Nino Perote, R.

    2000-01-01

    During severe accidents, it is foreseeable that the instrumentation installed in a plant is subjected to conditions which are more hostile than those for which the instrumentation was designed and qualified. Moreover, new, specific instrumentation is required to monitor variables which have not been considered until now, and to control systems which lessen the consequences of severe accidents. Both existing instrumentation used to monitor critical functions in design basis accident conditions and additional instrumentation which provides the information necessary to control and mitigate the consequences of severe accidents, have to be designed to withstand such conditions, especially in terms of measurements range, functional characteristics and qualification to withstand pressure and temperature loads resulting from steam explosion, hydrogen combustion/explosion and high levels of radiation over long periods of time. (Author)

  2. Accident management

    International Nuclear Information System (INIS)

    Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

    1989-01-01

    The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

  3. Development of a design basis tornado and structural design criteria for the Nevada Test Site, Nevada. Final report

    International Nuclear Information System (INIS)

    McDonald, J.R.; Minor, J.E.; Mehta, K.C.

    1975-06-01

    In order to evaluate the ability of critical facilities at the Nevada Test Site to withstand the possible damaging effects of extreme winds and tornadoes, parameters for the effects of tornadoes and extreme winds and structural design criteria for the design and evaluation of structures were developed. The meteorological investigations conducted are summarized, and techniques used for developing the combined tornado and extreme wind risk model are discussed. The guidelines for structural design include methods for calculating pressure distributions on walls and roofs of structures and methods for accommodating impact loads from wind-driven missiles. Calculations for determining the design loads for an example structure are included

  4. The Concept of Fashion Design on the Basis of Color Coordination Using White LED Lighting

    Science.gov (United States)

    Mizutani, Yumiko; Taguchi, Tsunemasa

    This thesis focuses on the development of fashion design, especially a dress coordinated with White LED Lighting (=LED). As for the design concept a fusion of the advanced science and local culture was aimed for. For such a reason this development is a very experimental one. Here in particular I handled an Imperial Court dinner dress for the last Japanese First Lady, Mrs. Akie Abe who wore it at the Imperial Court dinner for the Indonesian First Couple held on November 2006 to. This dress made by Prof. T. Taguchi and I open up a new field in the dress design.

  5. Unavoidable Accident

    OpenAIRE

    Grady, Mark F.

    2009-01-01

    In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

  6. The neuroscientific basis of successful design how emotions and perceptions matter

    CERN Document Server

    Maiocchi, Marco

    2015-01-01

    The term “design” today encompasses attributes of artifacts that go beyond their intended functions, imbuing them with new meanings. Those meanings are deeply related to the emotions perceived by the users. This book investigates the findings deriving from the neurosciences that are relevant to design. Drawing upon up-to-date neuroscientific knowledge, the authors define what an emotion is, examine the relationship between perceptions and emotions, and discuss the role of metaphoric communication. Particular attention is paid to those elements of perception and metaphoric interpretation that cause the emotions to rise. Consequences for the design process are then considered, and a design process is proposed that takes into account emotional impacts as one of the goals. A solid scientific approach to the subject is maintained throughout, and understanding is facilitated by the inclusion of a rich collection of successful design artifacts, the emotional aspects of which are analyzed.

  7. Design basis and requirements for 241-SY Modular Exhauster concrete pad and retaining wall

    International Nuclear Information System (INIS)

    Kriskovich, J.R.

    1994-01-01

    The purpose of this document is to serve as the design and functional requirements for a concrete pad for the new 241-SY Modular Exhauster and for a retaining wall to be built near the new ventilation systems

  8. Pupil filter design by using a Bessel functions basis at the image plane.

    Science.gov (United States)

    Canales, Vidal F; Cagigal, Manuel P

    2006-10-30

    Many applications can benefit from the use of pupil filters for controlling the light intensity distribution near the focus of an optical system. Most of the design methods for such filters are based on a second-order expansion of the Point Spread Function (PSF). Here, we present a new procedure for designing radially-symmetric pupil filters. It is more precise than previous procedures as it considers the exact expression of the PSF, expanded as a function of first-order Bessel functions. Furthermore, this new method presents other advantages: the height of the side lobes can be easily controlled, it allows the design of amplitude-only, phase-only or hybrid filters, and the coefficients of the PSF expansion can be directly related to filter parameters. Finally, our procedure allows the design of filters with very different behaviours and optimal performance.

  9. Design basis for creep of zirconium alloy components in a fast neutron flux

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Fidleris, V.

    1975-01-01

    The chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behaviour of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

  10. Ergonomic Analysis of Tricycle Sidecar Seats: Basis for Proposed Standard Design

    Directory of Open Access Journals (Sweden)

    Michael C. Godoy

    2015-12-01

    Full Text Available Ergonomics (also called human factors engineering is the study of human characteristics for the appropriate design of the living and work environment. It is applied in various industrial areas which includes transportation.Tricycle being one of the most common means of public transportation in Lipa City has various adaptations to suit the culture, and environment. The purpose of this study is to analyze the variability in design of the tricycles in Lipa City, Philippines and propose a standard ergonomically designed tricycle sidecar seat for a greater population. The study was conducted at 26 tricycle terminals with 232 tricycle samples within Lipa City proper including the public market area where 400 commuters were given questionnaires to determine the risk factors associated with the existing tricycle sidecar seat design. Anthropometric measurements of 100 males and 100 female commuters were obtained together with the sidecar dimensions of 232 tricycles to substantiate the observed variations in design. Using the design for the average and design for the extremes, it was found out that most of the tricycles in Lipa City, Philippines have inappropriate inclined seat and lowered sidecar seat pan height which can result to leg and abdominal pain; narrowed seat pan depth which caused pressure on buttocks and legs; narrowed backrest width which can cause upper and low back pain; low backrest height that can pose upper back pain; which can also result to abdominal pain; inclined backrest and limited vertical clearance which can cause upper back pain and neck pain. The researcher proposed a sidecar seat design standard which can be used by the Land Transportation Office, and Land Transportation Franchising and Regulatory Board to provide ease, comfort, and convenience to the passengers.

  11. Design basis for creep of zirconium alloy components in a fast neutron flux

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Fidleris, V.

    1974-01-01

    The Chalk River Nuclear Laboratory's experience with the creep of zirconium alloys in a neutron flux is described. Fast neutron flux changes the creep behavior of zirconium alloys and new design criteria for in-reactor applications are needed. From experimental results empirical relations describing the effects of neutron flux, stress, temperature, time, and anisotropy on creep rate were established. The relations are applied to the design of pressure tubes. (author)

  12. Design criteria and design basis for the 100-HR-3 and 100-KR-4 pump-and-treat projects

    International Nuclear Information System (INIS)

    McKinley, W.S.; Winters, J.N.

    1996-06-01

    The 100-HR-3 and 100-KR-4 Operable Units are located in the 100 Area at the Hanford Site in Richland, Washington. The document describes the project objectives and design criteria to be used for the 100-HR-3 and 100-KR-4 groundwater pump-and-treat design activities

  13. Designing Smart Artifacts for Adaptive Mediation of Social Viscosity: Triadic Actor-Network Enactments as a Basis for Interaction Design

    Science.gov (United States)

    Salamanca, Juan

    2012-01-01

    With the advent of ubiquitous computing, interaction design has broadened its object of inquiry into how smart computational artifacts inconspicuously act in people's everyday lives. Although user-centered design approaches remains useful for exploring how people cope with interactive systems, they cannot explain how this new breed of…

  14. Preliminary thermal design of a pressurized water reactor containment for handling severe accident consequences

    International Nuclear Information System (INIS)

    Abdullah, A.M.; Karameldin, A.

    1998-01-01

    A one-dimensional mathematical model has been developed for a 4250 MW(th) Advanced Pressurized Water Reactor containment analysis following a severe accident. The cooling process of the composite containment-steel shell and concrete shield- is achievable by natural circulation of atmospheric air. However, for purpose of gettering higher degrees of safety margin, the present study undertakes two objectives: (1) Installment of a diesel engine-driven air blower to force air through the annular space between the steel shell and concrete shield. The engine can be remotely operated to be effective in case of station blackout. (ii) Fixing longitudinally plate fins on the circumference of the inside and outside containment steel shell. These fins increase the heat transfer areas and hence the rate of heat removal from the containment atmosphere. In view of its importance - from the safety viewpoint - the long term behaviour of the containment which is a quasi-steady state problem, is formulated through a system of coupled nonlinear algebraic equations which describe the thermal-hydraulic and thermodynamic behaviour of the double shell containment. The calculated results revealed the following: (i) the passively air cooled containment can remove maximum heat load of 11.5 MW without failure, (ii) the effect of finned surface in the air passage tends to decrease the containment pressure by 20 to 30%, depending on the heat load, (iii) the effect of condensing fins is negligible for the proposed fin dimensions and material. However, by reducing the fin width, increasing their thickness, doubling their number, and using a higher conductive metal than the steel, it is expected that the containment pressure can be further reduced by 10% or more, (iv) the fins' dimensions and their number must be optimized via maximizing the difference or the ratio between the heat removed and pressure drop to get maximum heat flow rate

  15. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    Camous, F.; Chesnel, A.

    1989-12-01

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  16. Physics basis for an axicell design for the end plugs of MFTF-B

    International Nuclear Information System (INIS)

    Baldwin, D.E.; Logan, B.G.

    1982-01-01

    The primary motivation for conversion of MFTF-B to an axicell configuration lies in its engineering promise as a reactor geometry based on circular high-magnetic-field coils. In comparing this configuration to the previous A-cell geometry, we find a number of differences that might significantly affect the physics performance. The purpose of the present document is to examine those features and to assess their impact on the performance of the axicell, as compared to the A-cell configuration, for MFTF-B. In so doing, we address only those issues thought to be affected by the change in geometry and refer to the original report Physics Basis for MFTF-B, for discussion of those issues thought not to be affected. In Sec. 1, we summarize these physics issues. In Sec. 2, we describe operating scenarios in the new configuration. In the Appendices, we discuss those physics issues that require more detailed treatment

  17. Lessons of the radiological accident in Goiania

    International Nuclear Information System (INIS)

    Alves, R.N.; Xavier, A.M.; Heilbron, P.F.L.

    1998-01-01

    On the basis of the lessons teamed from the radiological accident of Goiania, actions are described which a nuclear regulatory body should undertake while responding to an accident of this nature. (author)

  18. Containment pressure monitoring method after severe accident in nuclear power plant

    International Nuclear Information System (INIS)

    Luo Chuanjie; Zhang Shishui

    2011-01-01

    The containment atmosphere monitoring system in nuclear power plant was designed on the basis of design accident. But containment pressure will increase greatly in a severe accident, and pressure instrument in the containment can't satisfy the monitoring requirement. A new method to monitor the pressure change in the containment after a severe accident was considered, through which accident soften methods can be adopted. Under present technical condition, adding a pressure monitoring channel out of containment for post-severe accident is a considerable method. Daya Bay Nuclear Power Plant implemented this modification, by which the containment release time can be delayed during severe accident, and nuclear safety can be increased. After analysis, this method is safe and feasible. (authors)

  19. Regulatory impact of nuclear reactor accident source term assumptions. Technical report

    International Nuclear Information System (INIS)

    Pasedag, W.F.; Blond, R.M.; Jankowski, M.W.

    1981-06-01

    This report addresses the reactor accident source term implications on accident evaluations, regulations and regulatory requirements, engineered safety features, emergency planning, probabilistic risk assessment, and licensing practice. Assessment of the impact of source term modifications and evaluation of the effects in Design Basis Accident analyses, assuming a change of the chemical form of iodine from elemental to cesium iodide, has been provided. Engineered safety features used in current LWR designs are found to be effective for all postulated combinations of iodine source terms under DBA conditions. In terms of potential accident consequences, it is not expected that the difference in chemical form between elemental iodine and cesium iodide would be significant. In order to account for the current information on source terms, a spectrum of accident scenerios is discussed to realistically estimate the source terms resulting from a range of potential accident conditions

  20. Physiological Basis for Prompt Health Effects

    International Nuclear Information System (INIS)

    VINCENT, Andrew

    2006-01-01

    As input to design considerations precluding worker radiological exposure that could lead to an acute health effect from a postulated accident condition, an assessment of the short term health effects was performed. To assure that the impact of the accident scenario on the individual is appropriately considered, both external and internal exposures are included in the evaluation. The focus of this evaluation was to develop a quantitative basis from which to consider the level of exposure postulated in an accident that could lead to a defined physiological impact for short term health effects. This paper does not assess latent health effects of radiological exposure associated with normal operations or emergency response guidelines as these are clearly articulated in existing regulations and ICRP documents. The intent of this paper is to facilitate a dialogue on the appropriate meaning of currently undefined terms such as ''significant'' exposure and ''high-hazard material'' in DSA development

  1. Planning of designing and installation of mechanical elements at the gear speed reducer on the basis of the parameter technology

    Directory of Open Access Journals (Sweden)

    D. Letić

    2013-01-01

    Full Text Available The development and implementation of the computer methods at project managing in the part of the planning of designing and installation of mechanical elements with the fit (assembly block of the gear speed reducer is significant and at present irreplaceable engineering task if it has been realized by the modern parameter technology. There are multifunction uses of this organized group of activities, beginning from the quick changeability of elements still in the phase of designing and constructing, thanks to the characteristics of their associativity, still to the wide basis of standard elements that are incorporated in the very program package. Meanwhile, these activities are not simple, so their realization has to be planned from the stand - point of time, resource and cost of realization. For the very designing and constructing was used AutoCAD Mechanical, and for the design managing Microsoft Project.

  2. Groundwork: Preparing an Effective Basis for Communication and Shared Learning in Design and Technology Education

    Science.gov (United States)

    Looijenga, Annemarie; Klapwijk, Remke; de Vries, Marc J.

    2016-01-01

    In Dutch Design and Technology Education the beginning of a process of learning is usually determined by the teacher. In this paper it is argued that a beginning, determined in interaction with the students, is more profitable as the interaction will lead to joined-up exploring, creating and thinking and an increased motivation to learn.…

  3. Workshops for professionals as basis for students' workshops for sustainable design

    NARCIS (Netherlands)

    Zeiler, W.; Houten, van M.A.; Savanovic, P.; Kim, S.; Chen, L.

    2009-01-01

    The growing complexity due to the increased demand for more sustainability in (Dutch) building practice necessitates developments in other aspects, besides specialized and professional skills. Therefore a new integral approach in building design education has been developed in close cooperation with

  4. Comparison of analyzed design-basis events to actual plant transients

    International Nuclear Information System (INIS)

    Geeting, M.W.; Hightower, N.T. III; Fields, C.C.

    1992-01-01

    Fitness-for-Service Guidelines have recently been developed to provide acceptance criteria and evaluation methods for assessment of the integrity of the Zr-2.5 Nb pressure tubes in operating Canada deuterium uranium (CANDU) reactors. The guidelines provide a methodology for the evaluation of specific conditions in a single tube, such as manufacturing and inservice generated flaws, hydride blisters formed at points of contact between a pressure tube and its calandria tube, and generic degradation of pressure tube properties in service. The guidelines are divided into three sections. The first section describes the requirements that must be met to qualify the tubes ofr continued service. The second section provides the material properties data-base information needed to carry out the assessments. The third section provides the technical basis for the acceptance criteria and evaluation procedures as well as justifications and descriptions of the data bases. The guidelines were issued to CANDU reactor operators for trial use and released to the Atomic Energy Control Board of Canada for review and comment in May 1991

  5. Reducing Production Basis Risk through Rainfall Intensity Frequency (RIF) Indexes: Global Sensitivity Analysis' Implication on Policy Design

    Science.gov (United States)

    Muneepeerakul, Chitsomanus; Huffaker, Ray; Munoz-Carpena, Rafael

    2016-04-01

    The weather index insurance promises financial resilience to farmers struck by harsh weather conditions with swift compensation at affordable premium thanks to its minimal adverse selection and moral hazard. Despite these advantages, the very nature of indexing causes the presence of "production basis risk" that the selected weather indexes and their thresholds do not correspond to actual damages. To reduce basis risk without additional data collection cost, we propose the use of rain intensity and frequency as indexes as it could offer better protection at the lower premium by avoiding basis risk-strike trade-off inherent in the total rainfall index. We present empirical evidences and modeling results that even under the similar cumulative rainfall and temperature environment, yield can significantly differ especially for drought sensitive crops. We further show that deriving the trigger level and payoff function from regression between historical yield and total rainfall data may pose significant basis risk owing to their non-unique relationship in the insured range of rainfall. Lastly, we discuss the design of index insurance in terms of contract specifications based on the results from global sensitivity analysis.

  6. Review of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Connelly, J.W.; Storr, G.J.

    1989-01-01

    Two types of severe reactor accidents - loss of coolant or coolant flow and transient overpower (TOP) accidents - are described and compared. Accidents in research reactors are discussed. The 1961 SL1 accident in the US is used as an illustration as it incorporates the three features usually combined in a severe accident - a design flaw or flaws in the system, a circumvention of safety circuits or procedures, and gross operator error. The SL1 reactor, the reactivity accident and the following fuel-coolant interaction and steam explosion are reviewed. 3 figs

  7. Environment as a basis for the design of advertising structures by forming

    Science.gov (United States)

    Khmelevsky, Y. P.; Seryakov, V. A.; Mamontov, G. Y.; Tsarenko, D. T.

    2017-01-01

    A few different neighbouring styles of architectural forms are quite frequent in the cities of great historical past. As a result, a designer or architect has to solve the complex problem while designing the objects within such environment, i.e. one has to fit them naturally into the existing site development. Often, form making is found to be hard, due to the fact that the existing architectural forms of totally different stylistic execution coexist in the visual proximity. Presently, placement of the advertising bills in urban environment is both an urgent and debatable issue. On the one hand, advertising providers are keen to present their product bigger and brighter, on the other hand, the overall and eye-catching exhibition stands can be disharmonious with the surrounding architectural ensemble of the city. This situation is relevant for every cultural city.

  8. Realistic minimum accident source terms - Evaluation, application, and risk acceptance

    International Nuclear Information System (INIS)

    Angelo, P. L.

    2009-01-01

    The evaluation, application, and risk acceptance for realistic minimum accident source terms can represent a complex and arduous undertaking. This effort poses a very high impact to design, construction cost, operations and maintenance, and integrated safety over the expected facility lifetime. At the 2005 Nuclear Criticality Safety Division (NCSD) Meeting in Knoxville Tenn., two papers were presented mat summarized the Y-12 effort that reduced the number of criticality accident alarm system (CAAS) detectors originally designed for the new Highly Enriched Uranium Materials Facility (HEUMF) from 258 to an eventual as-built number of 60. Part of that effort relied on determining a realistic minimum accident source term specific to the facility. Since that time, the rationale for an alternate minimum accident has been strengthened by an evaluation process that incorporates realism. A recent update to the HEUMF CAAS technical basis highlights the concepts presented here. (authors)

  9. Assessment of Equipment Capability to Perform Reliably under Severe Accident Conditions

    International Nuclear Information System (INIS)

    2017-07-01

    The experience from the last 40 years has shown that severe accidents can subject electrical and instrumentation and control (I&C) equipment to environmental conditions exceeding the equipment’s original design basis assumptions. Severe accident conditions can then cause rapid degradation or damage to various degrees up to complete failure of such equipment. This publication provides the technical basis to consider when assessing the capability of electrical and I&C equipment to perform reliably during a severe accident. It provides examples of calculation tools to determine the environmental parameters as well as examples and methods that Member States can apply to assess equipment reliability.

  10. Identification of important ''PIUS'' design considerations and accident sequences using qualitative plant assessment techniques

    International Nuclear Information System (INIS)

    Higgins, J.; Fullwood, R.; Kroeger, P.; Youngblood, R.

    1992-01-01

    The PIUS (Process Inherent Ultimate Safety) reactor is an advanced design nuclear power plant that uses passive safety features and basic physical processes to address safety concerns. Brookhaven National Laboratory (BNL) performed a detailed study of the PIUS design for the NRC using primarily qualitative engineering analysis techniques. Some quantitative methods were also employed. There are three key initial areas of analysis: FMECA, HAZOP, and deterministic analyses, which are described herein. Once these three analysis methods were completed, the important findings from each of the methods were assembled into thePIUS Interim Table (PIT). This table thus contains a first cut sort of the important design considerations and features of the PIUS reactor. The table also identifies some potential initiating events and systems used for mitigating these initiators. The next stage of the analysis was the construction of event trees for each of the identified initiators. The most significant sequences were then determined qualitatively, using, some quantitative input. Finally, overall insights on the PIUS design developed from the PIT and from the event tree analysis were developed and presented

  11. Creep-Fatigue Life Design with Various Stress and Temperature Conditions on the Basis of Lethargy Coefficient

    International Nuclear Information System (INIS)

    Park, Jung Eun; Yang, Sung Mo; Han, Jae Hee; Yu, Hyo Sun

    2011-01-01

    High temperature and stress are encounted in power plants and vehicle engines. Therefore, determination of the creep-fatigue life of a material is necessary prior to fabricating equipment. In this study, life design was determined on the basis of the lethargy coefficient for different temperatures, stress and rupture times. SP-Creep test data was compared with computed data. The SP-Creep test was performed to obtain the rupture time for X20CrMoV121 steel. The integration life equation was considered for three cases with various load, temperature and load-temperature. First, the lethargy coefficient was calculated by using the obtained rupture stress and the rupture time that were determined by carrying out the SP-Creep test. Next, life was predicted on the basis of the temperature condition. Finally, it was observed that life decreases considerably due to the coupling effect that results when fatigue and creep occur simultaneously

  12. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in GE-designed operating plants and near-term operating license applications

    International Nuclear Information System (INIS)

    1980-01-01

    The results are presented of a generic evaluation of feedwater transients, small-break loss-of-coolant accidents (LOCAs), and other TMI-2-related events for General Electric Company (GE)-designed operating plants and near-term operating license applications to confirm or establish the bases for the continued safe operation of the operating plants. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas. Additional review of the accident is continuing and further information is being obtained and evaluated. Any new information will be reviewed and modifications will be made as appropriate

  13. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-01-01

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. Therefore, empirically based approaches that are used for other regions, such as Western North America, are not appropriate for Eastern North America. Moreover, recent advances in science and technology have now made it possible to combine theoretical and empirical methods to develop new procedures and models for estimating ground motion. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. Specifically considered are magnitudes M from 5 to 8, distances from 0 to 500 km, and frequencies from 1 to 35 Hz. This document, Volume IV, provides Appendix 8.B, Laboratory Investigations of Dynamic Properties of Reference Sites

  14. Design basis and requirements for 241-SY modular exhauster mechanical installation

    International Nuclear Information System (INIS)

    Kriskovich, J.R.

    1994-01-01

    A new ventilation system is being installed to serve as the K-1 primary exhauster. The existing K-1 primary exhauster will then become the backup. This ventilation system services waste tanks 241-SY-101, 102 and 103. The nominal flow rate through the ventilation system is 1,000 cfm. The new ventilation system will contain a moisture eliminator, a heater, a prefilter, two stages of HEPA filtration, an exhaust fan, a stack and stack sampling system. The purpose of this document is to serve as the design and functional requirements for the mechanical installation of the new 241-SY modular exhauster. The mechanical installation will include modifying the existing ductwork (i.e., installing a ''T'' to connect the new exhauster to the existing system), modifying the existing condensate drain lines to accommodate the new lines associated with the new exhauster, a maintenance platform near the stack of the new exhauster, guy wires and guy wire footings to support the stack of the new exhauster, as well as other miscellaneous tasks associated with the mechanical installation design effort

  15. Protein structures in Alzheimer's disease: The basis for rationale therapeutic design.

    Science.gov (United States)

    Montoliu-Gaya, Laia; Villegas, Sandra

    2015-12-15

    Alzheimer's disease (AD) is a neurodegenerative disorder that affects memory, behavior, thinking and emotion. Current therapies to treat AD patients are only capable for temporarily slowing-down the cognitive decline, as they are focused on ameliorating symptoms instead of targeting its underlying causes. The aim of this review is to describe what is known about the protein structures implicated in AD pathogenesis, amyloid cascade members, as well as those structures involved in Aβ clearance. Thus, structural information available for APP, α- β- and γ-secretases, CTFβ and derived Aβ peptides, AICDs, apoE and apoJ, LRP-1 and RAGE, and neprilysin and insulin-degrading enzyme is provided. The recently solved structure for the γ-secretase complex opens the rational design of a new generation of inhibitors, whereas that for Aβ oligomers offers a putative mechanism explaining why monoclonal antibodies targeted to the N-terminus are effective. Then, an overview on therapies targeting some of these molecules presents their benefits and drawbacks. As a general conclusion our knowledge on the protein structures involved in AD has recently substantially advanced, allowing for the rational design of different therapeutic approaches. Hopefully, we are getting closer to finding a strong disease-modifying drug to cure this devastating disease. Copyright © 2015 Elsevier Inc. All rights reserved.

  16. Early Site Permit Demonstration Program: Guidelines for determining design basis ground motions

    International Nuclear Information System (INIS)

    1993-01-01

    This report develops and applies a methodology for estimating strong earthquake ground motion. The motivation was to develop a much needed tool for use in developing the seismic requirements for structural designs. An earthquake's ground motion is a function of the earthquake's magnitude, and the physical properties of the earth through which the seismic waves travel from the earthquake fault to the site of interest. The emphasis of this study is on ground motion estimation in Eastern North America (east of the Rocky Mountains), with particular emphasis on the Eastern United States and southeastern Canada. Eastern North America is a stable continental region, having sparse earthquake activity with rare occurrences of large earthquakes. While large earthquakes are of interest for assessing seismic hazard, little data exists from the region to empirically quantify their effects. Therefore, empirically based approaches that are used for other regions, such as Western North America, are not appropriate for Eastern North America. Moreover, recent advances in science and technology have now made it possible to combine theoretical and empirical methods to develop new procedures and models for estimating ground motion. The focus of the report is on the attributes of ground motion in Eastern North America that are of interest for the design of facilities such as nuclear power plants. Specifically considered are magnitudes M from 5 to 8, distances from 0 to 500 km, and frequencies from 1 to 35 Hz

  17. Cost analysis of a commercial pyroprocess facility on the basis of a conceptual design in Korea

    International Nuclear Information System (INIS)

    Kim, S.K.; Ko, W.I.; Youn, S.R.; Gao, Ruxing

    2015-01-01

    Highlights: • Pyroprocess facility’s direct cost was calculated based on the conceptual design. • The unit cost of pyroprocess was calculated as $781/kgHM. • The unit cost was increased by 3%, considering labor allocation standards. • The operating and maintenance cost was identified as a main cost driver. - Abstract: This study postulated a commercial pyroprocess facility (KAPF+: Korea Advanced Pyroprocess Facility Plus) with a processing capacity of 400 tons/year as a cost object, and utilized an engineering cost estimation method based on a conceptual design to present the results of the total cost and unit cost estimation. According to the calculation results, the total cost and unit cost were calculated with k$779,386 and $781/kgHM, respectively. Moreover, the key cost driver was manifested as the operating and maintenance costs. In particular, equipment replacement cost was identified as an important cost driver. In addition, for an increasingly accurate cost estimation, the calculation results and allocation method of the indirect cost were reanalyzed. Finally the pyroprocess unit cost increased $5 when calculated the indirect cost using the labor time as the allocation standard. Meanwhile, the pyroprocess unit cost increased $22 as a result of allocating the indirect cost using the uniform labor cost as the cost allocation standard. Accordingly, an indirect cost allocation standard was manifested as the factor that exerts a significant effect on the pyroprocess unit cost

  18. The analysis of the radiation induced cancer risks of workers of the nuclear industry - Liquidators of the accident on Chernobyl Atomic Station on the basis of modified poisson regressions

    International Nuclear Information System (INIS)

    Shafransky, I.L.; Tukov, A.R.

    2008-01-01

    Full text: The purpose of work consisted in reception of adequate estimations for additional relative risk in recalculation on 1 Sv in two dose diapason: up to 200 mSv and up to 500 mSv on the basis of materials on prevalence of malignant disease of workers of the nuclear industry - liquidators of the accident on Chernobyl Atomic Station. For this purpose methods of cohort analysis were used. This method realized on the basis of Poisson regression has been used. Estimations ERR on 1 Sv have been calculated as under the traditional scheme with use of module AMFIT (software EPICURE), and under the modified formula offered Paretzke. The received results have shown, that in some cases estimations for the risks, received on the modified formula, are more realistic, in other cases both estimations have close values. Also the lead analysis has shown no correct the procedure of carry of estimations of the risk received on one dose interval, on another a dose interval in a kind of nonlinear dependence of function of risk from a doze. As a whole, it is possible to tell, on an interval up to 200 mSv estimations of risk demand use of more complexes, than regression, models. In a range of dozes up to 500 mSv and even up to 1000 mSv the estimation of risk under the modified formula is more adequate. In a range of small doses application of the traditional approach on the basis of Linear non-threshold concept cannot be statistically justified and correct. (author)

  19. Preventing accidents

    Science.gov (United States)

    2005-08-01

    As the most effective strategy for improving safety is to prevent accidents from occurring at all, the Volpe Center applies a broad range of research techniques and capabilities to determine causes and consequences of accidents and to identify, asses...

  20. Thermal analyses for the rack design with spent fuel pool during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Yeh, C-L.; Chen, Y-S.; Chen, B-Y., E-mail: clinyeh@iner.gov.tw, E-mail: yschen@iner.gov.tw, E-mail: onepicemine@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China); Tseng, Y-S., E-mail: ystseng@mx.nthu.edu.tw [National Tsing Hua Univ., Engineering and System Science, Hsinchu, Taiwan (China); Wei, W-C., E-mail: hn150456@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China)

    2014-07-01

    Alternative fuel arrangements separating the latest fuels discharge from the reactor core are proposed, such as the 1x4 configuration in which the hot assembly is surrounded by 4 assemblies with much lower decay heat. For the rack design in the BWR spent fuel pool design, the lateral flow is eliminated by solid walls. In this study, cooling enhancement of splitting fuel rack is investigated using Computational Fluid Dynamics (CFD). The fuels in the pool are modeled by porous medium. Separating the fuel rack by a distance of 10 cm can lower the peak cladding temperature and the natural convection between the fuels and then earns more response time for the site people to implement necessary mitigation actions. (author)