Helmholtz-Zentrum Dresden-Rossendorf, Institute of Safety Research. Annual report 2010

International Nuclear Information System (INIS)

Gerbeth, Gunter; Schaefer, Frank

2011-01-01

The Institute of Safety Research (ISR) was over the past 20 years one of the six Research Institutes of Forschungszentrum Dresden-Rossendorf e.V. (FZD), which in 2010 belonged to the Wissenschaftsgemeinschaft Gottfried Wilhelm Leibniz. Together with the Institutes of Radiochemistry and Radiation Physics, ISR implements the research programme ''Nuclear Safety Research'' (NSR), which was during last years one of the three scientific programmes of FZD. NSR involves two main topics, i.e. ''Safety Research for Radioactive Waste Disposal'' and ''Safety Research for Nuclear Reactors''. The research of ISR aims at assessing and enhancing the safety of current and future reactors, the development of advanced simulation tools including their validation against experimental data, and the development of the appropriate measuring techniques for multi-phase flows and liquid metals.

Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

2002-03-01

The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

KfK, Institute for Reactor Components. Results of research and development activities in 1989

International Nuclear Information System (INIS)

1990-03-01

R and D activities at IRB (Institut fuer Reaktorbauelemente - Institute for Reactor Components) are dedicated to thermodynamics and fluid dynamics. Emphasis is on the design of nuclear reactor and fusion reactor components. Environmental engineering was added recently. Most activities are applications-oriented. Fundamental investigations focus on energy research and energy technology. The activities are carried out in the framework of different projects (PKF/nuclear fusion, PSF/nuclear safety, PSU/pollution control). Points of main effort are the development of basic liquid-metal-cooled blanket solutions, investigations on natural convection in reactor tanks, and the cooling properties of future containments for pressurized water reactors in the case of nuclear fusion accidents. (orig./GL) [de

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Helmholtz-Zentrum Dresden-Rossendorf, Institute of Safety Research. Annual report 2010

Energy Technology Data Exchange (ETDEWEB)

Gerbeth, Gunter; Schaefer, Frank (eds.)

2011-07-01

The Institute of Safety Research (ISR) was over the past 20 years one of the six Research Institutes of Forschungszentrum Dresden-Rossendorf e.V. (FZD), which in 2010 belonged to the Wissenschaftsgemeinschaft Gottfried Wilhelm Leibniz. Together with the Institutes of Radiochemistry and Radiation Physics, ISR implements the research programme ''Nuclear Safety Research'' (NSR), which was during last years one of the three scientific programmes of FZD. NSR involves two main topics, i.e. ''Safety Research for Radioactive Waste Disposal'' and ''Safety Research for Nuclear Reactors''. The research of ISR aims at assessing and enhancing the safety of current and future reactors, the development of advanced simulation tools including their validation against experimental data, and the development of the appropriate measuring techniques for multi-phase flows and liquid metals.

Guidelines for the Review of Research Reactor Safety: Revised Edition. Reference Document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

International Nuclear Information System (INIS)

2013-01-01

The Integrated Safety Assessment of Research Reactors (INSARR) is an IAEA safety review service available to Member States with the objective of supporting them in ensuring and enhancing the safety of their research reactors. This service consists of performing a comprehensive peer review and an assessment of the safety of the respective research reactor. The reviews are based on IAEA safety standards and on the provisions of the Code of Conduct on the Safety of Research Reactors. The INSARR can benefit both the operating organizations and the regulatory bodies of the requesting Member States, and can include new research reactors under design or operating research reactors, including those which are under a Project and Supply Agreement with the IAEA. The first IAEA safety evaluation of a research reactor operated by a Member State was completed in October 1959 and involved the Swiss 20 MW DIORIT research reactor. Since then, and in accordance with its programme on research reactor safety, the IAEA has conducted safety review missions in its Member States to enhance the safety of their research reactor facilities through the application of the Code of Conduct on the Safety of Research Reactors and the relevant IAEA safety standards. About 320 missions in 51 Member States were undertaken between 1972 and 2012. The INSARR missions and other limited scope safety review missions are conducted following the guidelines presented in this publication, which is a revision of Guidelines for the Review of Research Reactor Safety (IAEA Services Series No. 1), published in December 1997. This publication details those IAEA safety standards and guidance publications relevant to the safety of research reactors that have been revised or published since 1997. The purpose of this publication is to give guidance on the preparation, implementation, reporting and follow-up of safety review missions. It is also intended to be of assistance to operators and regulators in conducting

Safety of Research Reactors. Safety Requirements

International Nuclear Information System (INIS)

2010-01-01

The main objective of this Safety Requirements publication is to provide a basis for safety and a basis for safety assessment for all stages in the lifetime of a research reactor. Another objective is to establish requirements on aspects relating to regulatory control, the management of safety, site evaluation, design, operation and decommissioning. Technical and administrative requirements for the safety of research reactors are established in accordance with these objectives. This Safety Requirements publication is intended for use by organizations engaged in the site evaluation, design, manufacturing, construction, operation and decommissioning of research reactors as well as by regulatory bodies

Safety device for nuclear reactor

International Nuclear Information System (INIS)

Jacquelin, Roland.

1977-01-01

This invention relates to a safety device for a nuclear reactor, particularly a liquid metal (generally sodium) cooled fast reactor. This safety device includes an absorbing element with a support head connected by a disconnectable connector formed by the armature of an electromagnet at the end of an axially mobile vertical control rod. This connection is so designed that in the event of it becoming disconnected, the absorbing element gravity slides in a passage through the reactor core into an open container [fr

Results of 1989/90 research and development activities at KfK Institute for Reactor Components

International Nuclear Information System (INIS)

1991-03-01

R and D activities at IRB (Institut fuer Reaktorbauelemente - Institute for Reactor Components) are dedicated to thermodynamics and fluid dynamics. Emphasis is on the design of nuclear reactor and fusion reactor components. Environmental engineering was added recently. Most activities are applications-oriented. Fundamental investigations focus on energy research and energy technology. The activities are carried out in the framework of different projects (PKF/nuclear fusion, PSF/nuclear safety, PSU/pollution control). Points of main effort are the development of basic liquid-metal-cooled blanket solutions, investigations on natural convection in reactor ranks, and the cooling properties of future containments for pressurized water reactors in the case of nuclear fusion accidents. (orig./GL) [de

Ageing Management for Research Reactors. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

This Safety Guide was developed under the IAEA programme for safety standards for research reactors, which covers all the important areas of research reactor safety. It supplements and elaborates upon the safety requirements for ageing management of research reactors that are established in paras 6.68-6.70 and 7.109 of the IAEA Safety Requirements publication, Safety of Research Reactors. The safety of a research reactor requires that provisions be made in its design to facilitate ageing management. Throughout the lifetime of a research reactor, including its decommissioning, ageing management of its structures, systems and components (SSCs) important to safety is required, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions. Managing the safety aspects of research reactor ageing requires implementation of an effective programme for the monitoring, prediction, and timely detection and mitigation of degradation of SSCs important to safety, and for maintaining their integrity and functional capability throughout their service lives. Ageing management is defined as engineering, operation, and maintenance strategy and actions to control within acceptable limits the ageing degradation of SSCs. Ageing management includes activities such as repair, refurbishment and replacement of SSCs, which are similar to other activities carried out at a research reactor in maintenance and testing or when a modification project takes place. However, it is important to recognize that effective management of ageing requires the use of a methodology that will detect and evaluate ageing degradation as a consequence of the service conditions, and involves the application of countermeasures for prevention and mitigation of ageing degradation. The objective of this Safety Guide is to provide recommendations on managing ageing of SSCs important to safety at research reactors on the basis of international

Ageing Management for Research Reactors. Specific Safety Guide

International Nuclear Information System (INIS)

2010-01-01

This Safety Guide was developed under the IAEA programme for safety standards for research reactors, which covers all the important areas of research reactor safety. It supplements and elaborates upon the safety requirements for ageing management of research reactors that are established in paras 6.68-6.70 and 7.109 of the IAEA Safety Requirements publication, Safety of Research Reactors. The safety of a research reactor requires that provisions be made in its design to facilitate ageing management. Throughout the lifetime of a research reactor, including its decommissioning, ageing management of its structures, systems and components (SSCs) important to safety is required, to ensure continued adequacy of the safety level, reliable operation of the reactor, and compliance with the operational limits and conditions. Managing the safety aspects of research reactor ageing requires implementation of an effective programme for the monitoring, prediction, and timely detection and mitigation of degradation of SSCs important to safety, and for maintaining their integrity and functional capability throughout their service lives. Ageing management is defined as engineering, operation, and maintenance strategy and actions to control within acceptable limits the ageing degradation of SSCs. Ageing management includes activities such as repair, refurbishment and replacement of SSCs, which are similar to other activities carried out at a research reactor in maintenance and testing or when a modification project takes place. However, it is important to recognize that effective management of ageing requires the use of a methodology that will detect and evaluate ageing degradation as a consequence of the service conditions, and involves the application of countermeasures for prevention and mitigation of ageing degradation. The objective of this Safety Guide is to provide recommendations on managing ageing of SSCs important to safety at research reactors on the basis of international

Safety Analysis for Medium/Small Size Integral Reactor: Evaluation of Safety Characteristics for Small and Medium Integral Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hho jung; Seul, K W; Ahn, S K; Bang, Y S; Park, D G; Kim, B K; Kim, W S; Lee, J H; Kim, W K; Shim, T M; Choi, H S; Ahn, H J; Jung, D W; Kim, G I; Park, Y M; Lee, Y J [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

1997-07-01

The Small and medium integral reactor is developed to be utilized for non-electric areas such as district heating and steam production for desalination and other industrial purposes, and then these applications may typically imply a closeness between the reactor and the user. It requires the reactor to be designed with the adoption of special functional and inherent safety features to ensure and promote a high level of safety and reliability, in comparison with the existing nuclear power plants. The objective of the present study is to establish the bases for the development of regulatory requirements and technical guides to address the special safety characteristics of the small and medium integral reactor. In addition, the study aims to identify and to propose resolutions to the possible safety concerns in the design of the small and medium integral reactor. 34 refs., 20 tabs. (author)

KfK, Institute for Reactor Development. Results of research and development activities in 1990

International Nuclear Information System (INIS)

1991-03-01

R and D activities at IRE (Institut fuer Reaktorentwicklung-Institute for Reactor Development) are dedicated to power engineering and handling processes in the framework of the point-of-main-effort projects 'nuclear fusion', 'solid-state and materials research', 'handling', 'nuclear safety' and 'miscellaneous research'. Nuclear fusion contributions deal with special vacuum system design problems, heavy-duty design and materials selection and safety aspects. Specimens from 30 different carbon-based materials were subjected to comparative tests in a plasma spraying plant. In the framework of solid-state and materials research activities a viscoplastic model for mechanical component analysis was investigated for its compatibility with elementary physical laws under complex loads. The handling project was dedicated to the specific requirements of nuclear fusion with regard to JET and NET uses, the development of system solutions for flexible industrial techniques, and to standardization. A study on the control system of a work station for NET remote handling tasks was completed. In the framework of the nuclear safety project one currently investigates the dynamics of fast reactors under failure conditions, the possible propagation of local cooling failures in the reactor core, and core monitoring problems. (orig./GL) [de

Complementary Safety Assessments for Research Reactors for the French Nuclear Safety Authority

International Nuclear Information System (INIS)

Kassiotis, Christophe; Rigaud, Antoine; Evrard, Lydie

2013-01-01

The 'Autorite de surete nucleaire' (ASN) requested licensees to undertake stress tests, called complementary safety assessments (CSA), of their installations on May 5th 2011, following the accident that occurred in Japan on March 11th 2011. Their mission consisted in providing feedback on the consequences of potential extreme events. In this process, all the French facilities were divided into three categories of decreasing priority, depending on two main factors: on the one hand, their vulnerability to the various phenomena that led to the Fukushima accident, and on the other hand, the amount of radioactive elements that would be dispersed in the event of a failure of the safety functions. On the 79 high-priority facilities, only five of them are research or experimental reactors (including two currently shutdown or in decommissioning) and their operators (the 'Comissariat a l'energie atomique et aux energies alternatives' (CEA) and the 'Institut Laue Langevin') submitted their reports to the ASN on September 15 th 2011. Concerning the lower-priority facilities, including three other facilities (two research reactors operated by the CEA and a facility operated by ITER Organization) the deadline was September 15 th 2012. Finally, the remaining facilities were not asked to submit a report yet, but they will have to do it later, mainly on the occasion of their next periodic safety review. The analyses of the cliff-edge effects, that may occur in extreme situations (exceptional scale event, combination of several disasters...), led to the definition of a hardened safety core concept by the 'Institut de radioprotection et de surete nucleaire' (IRSN). This hardened safety core of structures, equipment and organizational measures must ensure the ultimate protection of the concerned facilities in extreme situations : it is designed to prevent severe accidents (or curb their progression), limit large scale releases for extreme accidents, and enables the operating teams to

Reactor safety

International Nuclear Information System (INIS)

Butz, H.P.; Heuser, F.W.; May, H.

1985-01-01

The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

Study on personnel qualification for non-destructive tests in the field of reactor safety

International Nuclear Information System (INIS)

Trusch, K.; Wuestenberg, H.

1977-01-01

The training system for non-destructive testing is described, and the available and necessary personnel is analyzed; the personnel required for reactor safety problems is treated separately. On this basis, the subjects discussed in the study - available personnel, personnel requirements, training, training requirements, and suggestions for realisation - are treated in a general manner to begin with and afterwards with a view to specific problems of reactor safety. The methods employed are adapted to this situation. To obtain the necessary empirical data, questionnaires were set up and distributed, and experts in selected business companies and institutions were interviewed who work in the field of reactor safety or do same training in non-destructive testing. (orig.) [de

Probabilistic safety assessment for research reactors

International Nuclear Information System (INIS)

1986-12-01

Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

International standardization of safety requirements for fast reactors

International Nuclear Information System (INIS)

2011-06-01

Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

Research for enhancing reactor safety

International Nuclear Information System (INIS)

1989-05-01

Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG) [de

Light water reactor safety

CERN Document Server

Pershagen, B

2013-01-01

This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Nuclear reactors safety issues

International Nuclear Information System (INIS)

Barre, Francois; Seiler, Nathalie

2008-01-01

Full text of publication follows: Since the seventies, economic incentives have led the utilities to drive a permanent evolution of the light water reactor (LWR). The evolution deals with the reactor designs as well as the way to operate them in a more flexible manner. It is for instance related to the fuel technologies and management. On the one hand, the technologies are in continuous evolution, such as the fuel pellets (MOX, Gd fuel, or Cr doped fuels..) as well as advanced cladding materials (M5 TM , MDA or ZIRLO). On the other hand, the fuel management is also subject to continuous evolution in particular in terms of increasing the level of burn-up, the reactor (core) power, the enrichment, as well as the duration of reactor cycles. For instance, in a few years in France, the burn-up has raised beyond the value of 39 GWj/t, initially authorized up to 52 GWj/t for the UO 2 fuel. In the near future, utilities foreseen to reach fuel burn-up of 60 GWj/t for MOX fuel and 70 GWj/t for UO 2 fuel. Furthermore, the future reactor of fourth generation will use new fuels of advanced conception. Furthermore with the objective of improving the safety margins, methods and calculation tools used by the utilities in the elaboration of their safety demonstrations submitted to the Safety Authority, are in movement. The margin evaluation methodologies often consist of a calculation chain of best-estimate multi-field simulations (e.g. various codes being coupled to simulate in a realistic way the evolution of the thermohydraulic, neutronic and mechanic state of the reactor). The statistical methods are more and more sophisticated and the computer codes are integrating ever-complex physical models (e.g. three-dimensional at fine scale). Following this evolution, the Institute of Radioprotection and Nuclear Safety (IRSN), whose one of the roles is to examine the safety records and to rend a technical expertise, considers the necessity of reevaluating the safety issues for advanced

An overview-probabilistic safety analysis for research reactors

International Nuclear Information System (INIS)

Liu Jinlin; Peng Changhong

2015-01-01

For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

Methods and strategies for future reactor safety goals

Science.gov (United States)

Arndt, Steven Andrew

There have been significant discussions over the past few years by the United States Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and others as to the adequacy of the NRC safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. In its review of the new plant licensing policy the ACRS has looked at the safety goals, as has the NRC. A number of issues have been raised including what the Commission had in mind when it drafted the safety goals and QHOs, how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. The NRC safety goals and the QHOs refer to a "nuclear power plant," but do not discuss whether a "plant" refers to only a single unit or all of the units on a site. There has been much discussion on this issue recently due to the development of modular reactors. Additionally, the risk of multiple reactor accidents on the same site has been largely ignored in the probabilistic risk assessments (PRAs) done to date, and in most risk

Experience with safety assessment of digital upgrading of IandC in VVER type reactors

International Nuclear Information System (INIS)

Wach, D.; Mulka, B.; Schnuerer, G.

1997-01-01

The digital upgrading of IandC systems important to safety in WWER type reactors requires a broad expertise in various knowledge fields. The approach of the Institute for safety Technology to the qualification and categorization of safety-critical software systems is highlighted. The role of the Institute in the qualification of the Teleperm XS and the type testing of its components is described. The aspects of the safety assessment of digital IandC systems in WWER type reactors is discussed in some detail. (A.K.)

Transactions of the Twentieth Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1992-10-01

This report contains summaries of papers on reactor safety research to be presented at the 20th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 21--23, 1992. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included

Progress report on research and development in 1991, Institute of Neutron Physics and Reactor Engineering, KfK

International Nuclear Information System (INIS)

1992-03-01

Progress report on research and development in 1991 Institute of Neutron Physics and Reactor Engineering. The Institute of Neutron Physics and Reactor Engineering is concerned with research work in the field of nuclear engineering related to the safety of fast and thermal reactors as well as with specific problems of fusion reactor technology. Under the project of nuclear safety research, the Institute works on concepts designed to drastically improve reactor safety. Apart from that, methods to estimate and minimize the radiological consequences of reactor accidents are developed. Under the fusion technology project, the Institute deals with neutron physics and technological questions of the breeding blanket. Basic research covers technico-physical questions of the interaction between light ion radiation of a high energy density and matter. In addition and to a small extent, questions of employing hydrogen in the transport area are studied. For all these tasks it is indispensable to use up-to-date data processing methods and equipment, from the highest capacity computer to the integrated minicomputer system. (orig./DG) [de

RB research reactor safety report

International Nuclear Information System (INIS)

Sotic, O.; Pesic, M.; Vranic, S.

1979-04-01

This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document

Development of Safety Review Guidance for Research and Training Reactors

International Nuclear Information System (INIS)

Oh, Kju-Myeng; Shin, Dae-Soo; Ahn, Sang-Kyu; Lee, Hoon-Joo

2007-01-01

The KINS already issued the safety review guidance for pressurized LWRs. But the safety review guidance for research and training reactors were not developed. So, the technical standard including safety review guidance for domestic research and training reactors has been applied mutates mutandis to those of nuclear power plants. It is often difficult for the staff to effectively perform the safety review of applications for the permit by the licensee, based on peculiar safety review guidance. The NRC and NSC provide the safety review guidance for test and research reactors and European countries refer to IAEA safety requirements and guides. The safety review guide (SRG) of research and training reactors was developed considering descriptions of the NUREG- 1537 Part 2, previous experiences of safety review and domestic regulations for related facilities. This study provided the safety review guidance for research and training reactors and surveyed the difference of major acceptance criteria or characteristics between the SRG of pressurized light water reactor and research and training reactors

Reactor safety training for decision making

International Nuclear Information System (INIS)

Scott, C.K.

2003-01-01

The purpose of this paper is to describe an approach to reactor safety training for technical staff working at an operating station. The concept being developed is that, when the engineer becomes a registered professional engineer, they have sufficient reactor safety knowledge to perform independent technical work without compromising the safety of the plant. This goal would be achieved with a focused training program while working as an engineer-in-training (four years in NB). (author)

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2004

International Nuclear Information System (INIS)

Weiss, F.P.; Rindelhardt, U.

2005-01-01

The Institute of Safety Research (ISR) is one of the six Research Institutes of Forschungszentrum Rossendorf e.V. (FZR e.V.) which is a member institution of the Wissenschaftsgemeinschaft Gottfried Wilhelm Leibniz (Leibniz Association). Together with the Institute of Radiochemistry, ISR constitutes the research programme ''Safety and Environment'' which is one from three scientific programmes of FZR. In the framework of this research programme, the institute is responsible for the two subprogrammes ''Plant and Reactor Safety'' and ''Thermal Fluid Dynamics'', respectively. We also provide minor contributions to the sub-programme ''Radio-Ecology''. Moreover, with the development of a pulsed photo-neutron source at the radiation source ELBE (Electron linear accelerator for beams of high brilliance and low emittance), we are involved in a networking project carried out by the FZR Institute of Nuclear and Hadron Physics, the Physics Department of TU Dresden, and ISR. (orig.)

Taipower's reload safety evaluation methodology for pressurized water reactors

International Nuclear Information System (INIS)

Huang, Ping-Hue; Yang, Y.S.

1996-01-01

For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

Development of safety analysis technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

RB research reactor Safety Report

International Nuclear Information System (INIS)

Sotic, O.; Pesic, M.; Vranic, S.

1979-04-01

This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

Operational safety evaluation for minor reactor accidents

International Nuclear Information System (INIS)

Wang, O.S.

1981-01-01

The purpose of this paper is to address a concern of applying conservatism in analysing minor reactor incidents. A so-called ''conservative'' safety analysis may exaggerate the system responses and result in a reactor scram tripped by the reactor protective system (RPS). In reality, a minor incident may lead the reactor to a new thermal hydraulic steady-state without scram, and the mitigation or termination of the incident may entirely depend on operator actions. An example on a small steamline break evaluation for a pressurized water reactor recently investigated by the staff at the Washington Public Power Supply System is presented to illustrate this point. A safety evaluation using mainly the safety-related systems to be consistent with the conservative assumptions used in the Safety Analysis Report was conducted. For comparison, a realistic analysis was also performed using both the safety- and control-related systems. The analyses were performed using the RETRAN plant simulation computer code. The ''conservative'' safety analysis predicts that the incident can be turned over by the RPS scram trips without operator intervention. However, the realistic analysis concludes that the reactor will reach a new steady-state at a different plant thermal hydraulic condition. As a result, the termination of the incident at this stage depends entirely on proper operator action. On the basis of this investigation it is concluded that, for minor incidents, ''conservative'' assumptions are not necessary, sometimes not justifiable. A realistic investigation from the operational safety point of view is more appropriate. It is essential to highlight the key transient indications for specific incident recognition in the operator training program

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

International Nuclear Information System (INIS)

Nero, A.V.; Farnaam, M.R.K.

1977-01-01

This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

Safety of nuclear power reactors

International Nuclear Information System (INIS)

MacPherson, H.G.

1982-01-01

Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

RA research reactor in 'Vinca' Institute-approach to the decommissioning

International Nuclear Information System (INIS)

Ljubenov, V.Lj.; Pesic, M.P.; Sotic, O.

2002-01-01

In this paper short overview of decommissioning process of research reactors according to IAEA standards and world practice is given. Basic technical characteristics and details of operational history of the RA research reactor in Vinca Institute of Nuclear Sciences are present. The main nuclear and radiation safety problems related to the RA reactor facility are defined and the outlines of the future decommissioning project are proposed. (author)

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

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

Safety of Research Reactors. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2017-01-01

This Safety Requirements publication establishes requirements for all main areas of safety for research reactors, with particular emphasis on requirements for design and operation. It explains the safety objectives and concepts that form the basis for safety and safety assessment for all stages in the lifetime of a research reactor. Technical and administrative requirements for the safety of new research reactors are established in accordance with these objectives and concepts, and they are to be applied to the extent practicable for existing research reactors. The safety requirements established in this publication for the management of safety and regulatory supervision apply to site evaluation, design, manufacturing, construction, commissioning, operation (including utilization and modification), and planning for decommissioning of research reactors (including critical assemblies and subcritical assemblies). The publication is intended for use by regulatory bodies and other organizations with responsibilities in these areas and in safety analysis, verification and review, and the provision of technical support.

Independent assessment for new nuclear reactor safety

Directory of Open Access Journals (Sweden)

D'Auria Francesco

2017-01-01

Full Text Available A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On the one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs. Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry. The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty approach.

Independent assessment for new nuclear reactor safety

International Nuclear Information System (INIS)

D'Auria, F.; Glaeser, H.; Debrecin, N.

2017-01-01

A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs). Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry). The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty (BEPU) approach. (authors)

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2004

Energy Technology Data Exchange (ETDEWEB)

Weiss, F.P.; Rindelhardt, U. (eds.)

2005-07-01

The Institute of Safety Research (ISR) is one of the six Research Institutes of Forschungszentrum Rossendorf e.V. (FZR e.V.) which is a member institution of the Wissenschaftsgemeinschaft Gottfried Wilhelm Leibniz (Leibniz Association). Together with the Institute of Radiochemistry, ISR constitutes the research programme ''Safety and Environment'' which is one from three scientific programmes of FZR. In the framework of this research programme, the institute is responsible for the two subprogrammes ''Plant and Reactor Safety'' and ''Thermal Fluid Dynamics'', respectively. We also provide minor contributions to the sub-programme ''Radio-Ecology''. Moreover, with the development of a pulsed photo-neutron source at the radiation source ELBE (Electron linear accelerator for beams of high brilliance and low emittance), we are involved in a networking project carried out by the FZR Institute of Nuclear and Hadron Physics, the Physics Department of TU Dresden, and ISR. (orig.)

Review of light water reactor safety through the Three Mile Island accident

International Nuclear Information System (INIS)

Phung, D.L.

1984-05-01

This review of light water reactor safety through the Three Mile Island accident has the purpose of establishing the baseline over which safety achievement post-TMI is assessed, and the need for new reactor designs and business direction is judged. Five major areas of reactor safety pre-TMI are examined: (1) safety philosophy and institutions, (2) reactor design criteria, (3) operational problems, (4) the Rasmussen reactor safety study, and (5) the TMI accident and repercussions. Although nuclear power has made spectacular achievements over the period pre-TMI and although TMI is technically a minor accident, this review concludes that there were basic flaws in the technology and in the manner safety philosophy was conceived and carried out. These flaws included (1) a reactor design that has high core power density, low heat capacity, and low system tolerance to upsets, (2) reactor deployment that had been expedited without extensive operational experience, (3) rules and regulations that had to play catch-up with commercial reactor development, (4) an industry that was fragmented, short-sighted, and tended to rely on the Nuclear Regulatory Commission for safety guidance, (5) information that was not effectively shared, and (6) attention that was inadequate to the human aspects of reactor operation and to public reaction to the specter of a reactor accident, major or minor

Introduction to Safety Analysis Approach for Research Reactors

International Nuclear Information System (INIS)

Park, Suki

2016-01-01

The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2015-01-30

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

2000-01-01

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

Reactivity requirements and safety systems for heavy water reactors

International Nuclear Information System (INIS)

Kati, S.L.; Rustagi, R.S.

1977-01-01

The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

Safety philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

2002-01-01

Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

Energy Technology Data Exchange (ETDEWEB)

Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1996-09-01

An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

Annual report 1986 of Society for Reactor Safety (GRS)

International Nuclear Information System (INIS)

1987-01-01

The overall development of Society for Reactor Safety (GRS) is indicated by its annual volume of achievement. After initially high growth rates that achievement has come to stand at some 62 million marks or so over the last years. GRS today employs some 450 staff, of which 325 are scientific-technical experts representing altogether nearly 4000 man-years of experience in nuclear technology. The report presents essential results from the main points of activity of GRS over 10 years. In accordance with the aim of the institution they primarily cover projects in the sectors safety and radiation protection for nuclear facilities. Since 1985, since the extension of its contract of associates, GRS has also increasingly been concerned with safety questions in non-nuclear technology, concentrating mainly on the transfer of existing methods. (orig./DG) [de

Institute of Energy and Climate Research IEK-6 : nuclear waste management & reactor safety report 2009/2010 ; material science for nuclear waste management

OpenAIRE

Klinkenberg, M.; Neumeier, S.; Bosbach, D. (Editors)

2011-01-01

This is the first issue of a new series of bi-annual reports intended to provide an overview of research activities for the safe management of nuclear waste in the Institute of Energy and Climate Research (IEK-6), Nuclear Waste Management and Reactor Safety devision in Jülich. The report gives a thematic overview of the research in 2009 and 2010 by short papers of five to eight pages. Some papers are discussing the work within different projects with intensive overlap, such as ...

Development of Safety Analysis Technology for Integral Reactor

Energy Technology Data Exchange (ETDEWEB)

Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

2000-03-01

The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

Inherent safety characteristics of innovative reactors

International Nuclear Information System (INIS)

Heil, J.A.

1995-11-01

The added safety value of innovative or third generation reactor designs has been evaluated in order to determine the most suitable candidate for Dutch government funded research and development support. To this end, four innovative reactor concepts, viz. PIUS (Process Inherent Ultimate Safety), PRISM (Power Reactor Innovative Small), HTR-M (High Temperature Reactor Module) and MHTGR (Modular High Temperature Gas-cooled Reactor), have been studied and their passive and inherent safety characteristics have been outlined. Also the outlook for further technological and industrial development has been considered. The results of the study confirm the perspective of the innovative reactors for reduced dependence on active safety provisions and for a further reduced vulnerability to technical failures and human errors. The accident responses to generic accident initiators, viz. reactivity and cooling accidents, and also to reactor specific accidents show that neither active safety systems nor short term operator actions are required for maintaining the reactor core in a controlled and coolable condition. Whether this gives rise to a higher total safety of the innovative reactor designs, compared to evolutionary or advanced reactors, cannot be concluded. Supplementary experimental and analytical analyses of reactor specific accidents are required to be able to assess the safety of these innovative designs in a more quantitative manner. It is believed that the safety case of innovative reactors, which are less dependent on active safety systems, can be communicated with the general public in a more transparent way. Considering the perspective for further technological and industrial development it is not expected that any of the considered innovative reactor concepts will become commercially available within the next one to two decades. However, they could be made available earlier if they would receive sufficient financial backing. Considering the added safety perspectives

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza: Izvestaj o sigurnosti reaktora nulte snage RB

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1962-09-15

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined.

Reactor safety analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

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

Advanced gas cooled reactors - Designing for safety

International Nuclear Information System (INIS)

Keen, Barry A.

1990-01-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

Advanced gas cooled reactors - Designing for safety

Energy Technology Data Exchange (ETDEWEB)

Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

1990-07-01

The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

Commissioning of research reactors. Safety guide

International Nuclear Information System (INIS)

2006-01-01

The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

Causes of extended shutdown state of 'RA' research reactor in Vinca Institute

International Nuclear Information System (INIS)

Pesic, M.; Kolundzija, V.; Ljubenov, V.; Cupac, S.

2001-01-01

This paper describes the causes and reasons for extended shutdown state of RA research reactor in the 'Vinca' Institute of Nuclear Sciences. Technical and legal matters that led to decision to stop RA reactor operation in 1984 and further problems related to maintenance and preparation for continuation of operation are given. Influence of nuclear policy of Yugoslav government and the 'Vinca' Institute at prolongation of the reactor shutdown state, as consequence of changing of nuclear programme in the country and the world are discussed and underlined. An overview of the legislation in the field of nuclear safety and regulatory control of radiation sources and radioactive materials in Yugoslavia is presented. (author)

Nuclear safety cooperation for Soviet designed reactors

International Nuclear Information System (INIS)

Reisman, A.W.; Horak, W.C.

1995-01-01

The nuclear accident at the Chernobyl nuclear power plant in 1986 first alerted the West to the significant safety risks of Soviet designed reactors. Five years later, this concern was reaffirmed when the IAEA, as a result of a review by an international team of nuclear safety experts, announced that it did not believe the Kozloduy nuclear power plants in Bulgaria could be operated safely. To address these safety concerns, the G-7 summit in Munich in July 1992 outlined a five point program to address the safety problems of Soviet Designed Reactors: operational safety improvement; near-term technical improvements to plants based on safety assessment; enhancing regulatory regimes; examination of the scope for replacing less safe plants by the development of alternative energy sources and the more efficient use of energy; and upgrading of the plants of more recent design. As of early 1994, over 20 countries and international organizations have pledged hundreds of millions of dollars in financial assistance to improve safety. This paper summarizes these assistance efforts for Soviet designed reactors, draws lessons learned from these activities, and offers some options for better addressing these concerns

Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

International Nuclear Information System (INIS)

1975-06-01

The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

Japan: The institute for the economy of energy recommends a quick re-start of nuclear reactors

International Nuclear Information System (INIS)

Anon.

2014-01-01

The Japanese Institute for the Economy of the Energy (IEEJ) considers that the sooner the nuclear reactors will re-start, the better the Japanese economy and environment will be. The 48 Japanese reactors were stopped after the Fukushima accident and their restart is linked to the implementation of new measures for reinforcing safety. Until now only 2 reactors Sendai 1 and Sendai 2 have been allowed to re-start. The procedure for the safety assessment of the reactors is slower than expected. A study shows that only 7 reactors may be allowed to re-start before march 2015 and a total of 19 units may be operating in march 2016. In this scenario 2% of the electricity will come from nuclear energy in 2014 and 15% in 2015, natural gas imports will still be necessary for the production of electricity and their global cost is estimated to reach 56 billions euros while Japan's rate of energy independence will drop by 4.6%. (A.C.)

SRP reactor safety evolution

International Nuclear Information System (INIS)

Rankin, D.B.

1984-01-01

The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

Results on safety research for five years (from fiscal year 1996 to 2000). A field of power reactors

International Nuclear Information System (INIS)

2001-10-01

This safety research carried out by the Japan Nuclear Cycle Development Institute (JNC) for five years ranged from 1996 to 2000 fiscal year, was performed according to the safety research basic plan (from 1996 to 2000 fiscal year) established on March, 1996 (revised again on May, 2000). This report was arranged on a field on power reactors (all subjects on fields of advanced conversion reactor and a subject on power reactor in a field of seismic resistant and probability theoretical safety evaluation) by combining its research results for five years ranged from 1996 to 2000 fiscal year with general outlines on the safety research basic plan. Here were shown outlines on the safety research basic plan, aims and subjects on safety research at a field of power reactors, a list of survey sheets on safety research result, and survey sheets on safety research results. The survey sheets containing research field, title, organization, researcher name, researching period, names of cooperative organization, using facilities, research outline, research results, established contents, application, and research trends, are ranged to 5 items on advanced conversion reactor, 29 items on high breeder reactor, 1 item on seismic resistance, and 5 items on probability theoretical safety evaluation. (G.K.)

Transactions of the nineteenth water reactor safety information meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1991-10-01

This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately

Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

International Nuclear Information System (INIS)

1986-11-01

Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

Small nuclear reactor safety design requirements for autonomous operation

International Nuclear Information System (INIS)

Kozier, K.S.; Kupca, S.

1997-01-01

Small nuclear power reactors offer compelling safety advantages in terms of the limited consequences that can arise from major accident events and the enhanced ability to use reliable, passive means to eliminate their occurrence by design. Accordingly, for some small reactor designs featuring a high degree of safety autonomy, it may be-possible to delineate a ''safety envelope'' for a given set of reactor circumstances within which safe reactor operation can be guaranteed without outside intervention for time periods of practical significance (i.e., days or weeks). The capability to operate a small reactor without the need for highly skilled technical staff permanently present, but with continuous remote monitoring, would aid the economic case for small reactors, simplify their use in remote regions and enhance safety by limiting the potential for accidents initiated by inappropriate operator action. This paper considers some of the technical design options and issues associated with the use of small power reactors in an autonomous mode for limited periods. The focus is on systems that are suitable for a variety of applications, producing steam for electricity generation, district heating, water desalination and/or marine propulsion. Near-term prospects at low power levels favour the use of pressurized, light-water-cooled reactor designs, among which those having an integral core arrangement appear to offer cost and passive-safety advantages. Small integral pressurized water reactors have been studied in many countries, including the test operation of prototype systems. (author)

Relevant safety issues in designing the HTR-10 reactor

International Nuclear Information System (INIS)

Sun Yuliang; Xu Yuanghui

2001-01-01

The HTR-10 is a 10 MWth pebble bed high temperature gas cooled reactor being constructed as a research facility at the Institute of Nuclear Energy Technology. This paper discusses design issues of the HTR-10 which are related to safety. It addresses the safety criteria used in the development and assessment of the design, the safety important systems, and the safety classification of components. It also summarises the results of safety analysis, including the approach used for the radioactive source term, as well as the approach to containment design. (author)

Fusion reactor safety

International Nuclear Information System (INIS)

1987-12-01

Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

New generation main control room of enhanced safety NPP with MKER reactor

International Nuclear Information System (INIS)

Golovanev, V.E.; Gorelov, A.I.; Proshin, V.A.

1994-01-01

Russia is planning to begin the gradual substitution RMBK NPP units, whose resources were worked out itself, to NPP units with a 800 MW multiloop boiling water power reactor (MKER-800) enhanced safety at next ten-year period. Main drawbacks of RBMK Reactor were completely removed in design of MKER-800 reactor. Moreover some special decisions were made to give MKER-800 self-safety properties. The proposed design of the MKER-800 enhanced safety reactor is not only fully free from the drawbacks of the RBMK reactors, but also show a number of advantages of channel-type reactors. This Paper presents some preliminary proposals of MCR Design, that developed Research and Development Institute of Power Energy (RDIPE). 6 refs, 2 figs

Reactor safety method

International Nuclear Information System (INIS)

Vachon, L.J.

1980-01-01

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

Discussion on safety analysis approach for sodium fast reactors

International Nuclear Information System (INIS)

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

2012-01-01

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

Guidelines for Self-assessment of Research Reactor Safety

International Nuclear Information System (INIS)

2018-01-01

Self-assessment is an organization’s internal process to review its current status, processes and performance against predefined criteria and thereby to provide key elements for the organization’s continual development and improvement. Self-assessment helps the organization to think through what it is expected to do, how it is performing in relation to these expectations, and what it needs to do to improve performance, fulfil the expectations and achieve better compliance with the predefined criteria. This publication provides guidelines for a research reactor operating organization to perform a self-assessment of the safety management and the safety of the facility and to identify gaps between the current situation and the IAEA safety requirements for research reactors. These guidelines also provide a methodology for Member States, regulatory bodies and operating organizations to perform a self-assessment of their application of the provisions of the Code of Conduct on the Safety of Research Reactors. This publication also addresses planning, implementation and follow-up of actions to enhance safety and strengthen application of the Code. The guidelines are applicable to all types of research reactor and critical and subcritical assemblies, at all stages in their lifetimes, and to States, regulatory bodies and operating organizations throughout all phases of research reactor programmes. Research reactor operating organizations can use these guidelines at any time to support self-assessments conducted in accordance with the organization’s integrated management system. These guidelines also serve as a tool for an organization to prepare to receive an IAEA Integrated Safety Assessment of Research Reactors (INSARR) mission. An important result of this is the opportunity for an operating organization to identify focus areas and make safety improvements in advance of an INSARR mission, thereby increasing the effectiveness of the mission and efficiency of the

Proceedings of the seminar on nuclear safety research and the workshop on reactor safety research

International Nuclear Information System (INIS)

2001-07-01

The seminar on the nuclear safety research was held on November 20, 2000 according to the start of new five year safety research plan (FY2001-2005: established by Nuclear Safety Commission) with 79 participants. In the seminar, Commissioner Dr. Kanagawa gave the outline of the next five year safety research plan. Following this presentation, progresses and future scopes of safety researches in the fields of reactor facility, fuel cycle facility, radioactive waste and environmental impact on radiation at Japan Atomic Energy Research Institute (JAERI) were reported. After the seminar, the workshop on reactor safety research was held on November 21-22, 2000 with 141 participants. In the workshop, four sessions titled safety of efficient and economic utilization of nuclear fuel, safety related to long-term utilization of power reactors, research on common safety-related issues and toward further improvement of nuclear safety were organized and, outcomes and future perspectives in these wide research R and D in the related area at other organizations including NUPEC, JAPEIC and Kansai Electric Power Co. was presented in each session. This report compiles outlines of the presentations and used materials in the seminar and the workshop to form the proceedings for the both meetings. (author)

Progress report concerning safety research for nuclear reactor facilities

International Nuclear Information System (INIS)

1978-01-01

Examination and evaluation of safety research results for nuclear reactor facilities have been performed, as more than a year has elapsed since the plan had been initiated in April, 1976, by the special sub-committee for the safety of nuclear reactor facilities. The research is carried out by being divided roughly into 7 items, and seems to be steadily proceeding, though it does not yet reach the target. The above 7 items include researches for (1) criticality accident, (2) loss of coolant accident, (3) safety for light water reactor fuel, (4) construction safety for reactor facilities, (5) reduction of release of radioactive material, (6) safety evaluation based on the probability theory for reactor facilities, and (7) aseismatic measures for reactor facilities. With discussions on the progress and the results of the research this time, research on the behaviour on fuel in abnormal transients including in-core and out-core experiments has been added to the third item, deleting the power-cooling mismatch experiment in Nuclear Safety Research Reactor of JAERI. Also it has been decided to add two research to the seventh item, namely measured data collection, classification and analysis, and probability assessment of failures due to an earthquake. For these 7 items, the report describes the concrete contents of research to be performed in fiscal years of 1977 and 1978, by discussing on most rational and suitable contents conceivable at present. (Wakatsuki, Y.)

Safety of VVER-440 reactors

CERN Document Server

Slugen, Vladimir

2011-01-01

Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

Karlsruhe Nuclear Research Center, Institute of Neutron Physics and Reactor Engineering. Progress report on research and development work in 1993

International Nuclear Information System (INIS)

1994-03-01

The Institute of Neutron Physics and Reactor Engineering is concerned with research work in the field of nuclear engineering related to the safety of thermal reactors as well as with specific problems of fusion reactor technology. Under the project of nuclear safety research, the Institute works on concepts designed to drastically improve reactor safety. Apart from that, methods to estimate and minimize the radiological consequences of reactor accidents are developed. Under the fusion technology project, the Institute deals with neutron physics and technological questions of the breeding blanket. Basic research covers technico-physical questions of the interaction between light ion radiation of a high energy density and matter. In addition and to a small extent, questions of employing hydrogen in the transport area are studied. (orig.) [de

Transactions of the Twenty-First Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Monteleone, S.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session

Transactions of the Twenty-First Water Reactor Safety Information Meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

Fire safety requirements for electrical cables towards nuclear reactor safety

International Nuclear Information System (INIS)

Raju, M.R.

2002-01-01

Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

International Nuclear Information System (INIS)

2009-01-01

In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

The reactor safety study of experimental multi-purpose VHTR design

International Nuclear Information System (INIS)

Yasuno, T.; Mitake, S.; Ezaki, M.; Suzuki, K.

1981-01-01

Over the past years, the design works of the Experimental Very High Temperature Reactor (VHTR) plant have been conducted at Japan Atomic Energy Research Institute. The conceptual design has been completed and the more detailed design works and the safety analysis of the experimental VHTR plant are continued. The purposes of design studies are to show the feasibility of the experimental VHTR program, to specify the characteristics and functions of the plant components, to point out the R and D items necessary for the experimental VHTR plant construction, and to analyze the feature of the plant safety. In this paper the summary of system design and safety features of the experimental reactor are indicated. Main issues are the safety philosophy for the design basis accident, the accidents assumed and the engineered safety systems adopted in the design works

Reactor safety-a New Approach

International Nuclear Information System (INIS)

Machiels, A.J.; Marston, T.U.; Taylor, J.J.

1993-01-01

Since 1982, the U.S. utilities have been leading to an industry-wide effort to establish a technical foundation for the design of the next generation of light water reactors (LWRs) in the United States. Since 1985, the utility initiative has been effected through a major technical program managed by the Electric Power Research Institute (EPRI): the Advanced Light Water Reactor (ALWR) Program. In addition to the U.S. utility leadership and sponsorship, the ALWR Program has also greatly benefitted from the participation and sponsorship of numerous international utility companies and from the close cooperation with the U.S. Department of Energy (DOE). One of the main goals of the ALWR Program has been to develop a comprehensive set of design requirements for the advanced LWRs. The Utility Requirement Document (URD) defines the technical basis for improved and standardized future LWR designs. The URD covers the entire plant up to the grid interface. Therefore, it is the basis for an integrated plant design, i.e., nuclear steam supply system and balance of plant. It emphasizes those areas which are most important to the objective of achieving an advanced LWR that is excellent with respect to safety, performance, constructibility, and economics. There are numerous basic design policies underlying the ALWR URD. Of particular interest is the treatment of reactor safety

The action of the project coordinator with respect to reactor safety

International Nuclear Information System (INIS)

Leclercq, Jacques

1981-01-01

Before describing the various actions of the project coordinator (EDF) entrusted with the building of nuclear power stations, with respect to reactor safety in France, the definition of reactor safety and the various participants are mentioned first. These participants are: the Government Departments and the Experts involved (the Department of Nuclear Safety of the 'Institut de Protection et de Surete Nucleaire' forming the first technical support) and the applicant, namely the EDF. The reactor safety actions of the project coordinator are defined as from the following components: 1 - The targets laid down with respect to safety, the final objective being the protection of workers and the public against the potential dangers of the installations, principally against radiation. 2 - The safety methodology at the design stage of the power station: 'barrier' method, defence method in depth at three levels, lines of assurance method, and probabilistic method. 3 - Safety actions at the construction stage within the context of an assurance of quality programe. 4 - Safety at the trials, commissioning and operating stage, with the backing of the 'Groupe Operationnel de Demarrage (G.O.D.)' and the 'Commission d'Essais sur Site (C.E.S.)'. An initial balance sheet of the reactor safety actions for the PWR units built by the EDF is presented [fr

PX–An Innovative Safety Concept for an Unmanned Reactor

Directory of Open Access Journals (Sweden)

Sung-Jae Yi

2016-02-01

Full Text Available An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

Safety reviews of next-generation light-water reactors

International Nuclear Information System (INIS)

Kudrick, J.A.; Wilson, J.N.

1997-01-01

The Nuclear Regulatory Commission (NRC) is reviewing three applications for design certification under its new licensing process. The U.S. Advanced Boiling Water Reactor (ABWR) and System 80+ designs have received final design approvals. The AP600 design review is continuing. The goals of design certification are to achieve early resolution of safety issues and to provide a more stable and predictable licensing process. NRC also reviewed the Utility Requirements Document (URD) of the Electric Power Research Institute (EPRI) and determined that its guidance does not conflict with NRC requirements. This review led to the identification and resolution of many generic safety issues. The NRC determined that next-generation reactor designs should achieve a higher level of safety for selected technical and severe accident issues. Accordingly, NRC developed new review standards for these designs based on (1) operating experience, including the accident at Three Mile Island, Unit 2; (2) the results of probabilistic risk assessments of current and next-generation reactor designs; (3) early efforts on severe accident rulemaking; and (4) research conducted to address previously identified generic safety issues. The additional standards were used during the individual design reviews and the resolutions are documented in the design certification rules. 12 refs

Safety research needs for Russian-designed reactors

International Nuclear Information System (INIS)

1998-01-01

In June 1995, an OECD Support Group was set up to perform a broad study of the safety research needs of Russian-designed reactors. This Support Group was endorsed by the CSNI. The Support Group, which is composed of senior experts on safety research from several OECD countries and from Russia, prepared this Report. The Group reviewed the safety research performed to support Russian-designed reactors and set down its views on future needs. The review concentrates on the following main topics: Thermal-Hydraulics/Plant Transients for VVERs; Integrity of Equipment and Structures for VVERs; Severe Accidents for VVERs; Operational Safety Issues; Thermal-Hydraulics/Plant Transients for RBMKs; Integrity of Equipment and Structures for RBMKs; Severe Accidents for RBMKs. (K.A.)

Technical and institutional safety features of nuclear power plants in Brazil

International Nuclear Information System (INIS)

Rosa, L.P.

1986-01-01

This work reports technical, political and institutional safety features of nuclear power plants in Brazil. It is mainly concerned with reactor accidents and personnel safety. The three mile Island and Chernobyl accidents are also discussed and taken as examples. (A.C.A.S.)

Safety re-evaluation of the HOR reactor

International Nuclear Information System (INIS)

Verkooijen, A.H.M.; Vries, J.W. de

2001-01-01

The HOR reactor of the Interfaculty Reactor Institute (IRI) of Delft University of Technology has been in operation since 1963. This swimming pool type reactor is the main experimental facility of the institute. It is used for a broad range of both fundamental and applied research. Originally the reactor was intended to serve the Dutch universities. While retaining this inter-university character many of the research themes are carried out with both national and international scientific partners. Gradually the power of the reactor was increased to 3MW maximum operational power. In 1969 a licence was obtained for that power level. Ever since the reactor is operated at a nominal power level of 2 MW. In 1980 a permit was obtained for a new control room and for extension and renewal of the reactor instrumentation. Those modifications were finished in 1982. In the early nineties it became apparent that further modifications and extensions were necessary. As a consequence of the policy of the US government on proliferation the conversion of HEU to LEU was a prerequisite for the availability of fresh fuel and the return of spent fuel to the US. At the same time the development of new instruments for neutron scattering and positron annihilation asked for a new beam hall. Therefore a renewal of the permit was necessary. After extensive studies in which a safety report and an environmental impact statement were made, the application for a new licence was submitted in 1995. The new licence was granted in 1996. However, due to legal actions by individuals. the licence did not become effective until 1997. The beam hall was inaugurated in 1998. In the same year the conversion from HEU to LEU was started. In June 2000 the 'Delft Positron Centre' was officially opened. Here a unique high intensity positron beam serves the 2D-ACAR and Lifetime instruments. The legal procedures lasted until fall 2000 when all objections raised against the new licence were dismissed by the Council of

The IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

According to the research reactor database of IAEA (RRDB), 250 reactors are operating worldwide, 248 have been shut down and 170 have been decommissioned. Among the 248 reactors that do not run, some will resume their activities, others will be dismantled and the rest do not face a clear future. The analysis of reported incidents shows that the ageing process is a major cause of failures, more than two thirds of operating reactors are over 30 years old. It also appears that the lack of adequate regulations or safety standards for research reactors is an important issue concerning reactor safety particularly when reactors are facing re-starting or upgrading or modifications. The IAEA has launched a 4-axis program: 1) to set basic safety regulations and standards for research reactors, 2) to provide IAEA members with an efficient help for the application of these safety regulations to their reactors, 3) to foster international exchange of information on research reactor safety, and 4) to provide IAEA members with a help concerning safety issues linked to malicious acts or sabotage on research reactors

Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

Achievements and future directions in the reactors physics and nuclear safety research

International Nuclear Information System (INIS)

Dumitrache, Ion

2001-01-01

A historical overlook is presented with respect to inception and development of reactor physics research and on the job training in Romania. First these activities were carried out at the Institute for Atomic Physics and Institute for Power Reactors (IRNE) in Bucharest and afterward at the Institute for Nuclear Technologies, later on transformed in the Institute of Nuclear Research at Pitesti. CYBER Computer installed at Pitesti allowed formation in as early as 1971 reactor specialists who worked out computer programs for neutron physics calculations. These specialists were able to assimilate the characteristic of CANDU 6 type reactor as well as the AECL methodology of simulating processes of CANDU reactor physics. At present four programs are under way. These are: 1. The nuclear reactor physics; 2. The nuclear facility safety; 3. Safety analyses for the transport and radioactive waste disposal; 4. Analyses for radiation shielding and biological protection. There are presented results of the work associated to the CANDU type reactor: 1. Adapting and improving the code system for neutron and thermohydraulic calculation for CANDU type reactor, as supplied by AECL; 2. The IRNE manual for CANDU reactor neutron designing; 3. Final sizing of shim rods of Cernavoda NPP Unit 2; 4. Tests and measurements of reactor physics at the Cernavoda NPP Unit 1 commissioning; 5. Simulation and independent analysis of thermosiphoning carried out at Cernavoda NPP Unit 1 commissioning; 6. Static and dynamical response of the detectors in the CANDU reactor core and their time evolution following the burnup in the neutron flux and their ageing effects; 7. PSA studies at Unit 1; 8. Safety analyses for the radioactive waste disposal at Saligny repository. Also, reported are the results of the work associated to the TRIGA reactor, as follows: 1. Flux measurements and neutron computations necessary in the reactor commissioning; 2. Cleaning up controversial issues relating to neutron flux

IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Alcala, F.; Di Meglio, A.F.

1995-01-01

This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

Licensing procedures and safety criteria for research reactors in France

International Nuclear Information System (INIS)

Berry, J.L.; Lerouge, B.

1983-01-01

From the very beginning of the CEA up to now, a great deal of work has been devoted to the development and utilization of research reactors in France for the needs of fundamental and applied research, production of radioisotopes, and training. In recent years, new reactors were commissioned while others were decommissioned. Moreover some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors (Osiris and Isis). This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. At the end, a few considerations are given to the consequences of the Osiris core conversion. Safety of research reactors has been studied in detail and many improvements have been brought due to: implementation of a specific experimental program, and adaptation of safety principles and rules elaborated for power reactors. Research reactors in operation in France have been built within a 22 year period. Meanwhile, safety rules have been improved. Old reactors do not comply with all the new rules but modifications are continuously made: after analysis of incidents, when replacement of equipment has to be carried out, when an important modification (fuel conversion for example) is decided upon

Licensing procedures and safety criteria for research reactors in France

Energy Technology Data Exchange (ETDEWEB)

Berry, J L; Lerouge, B [Centre d' Etudes Nucleaires de Saclay (France)

1983-08-01

From the very beginning of the CEA up to now, a great deal of work has been devoted to the development and utilization of research reactors in France for the needs of fundamental and applied research, production of radioisotopes, and training. In recent years, new reactors were commissioned while others were decommissioned. Moreover some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors (Osiris and Isis). This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. At the end, a few considerations are given to the consequences of the Osiris core conversion. Safety of research reactors has been studied in detail and many improvements have been brought due to: implementation of a specific experimental program, and adaptation of safety principles and rules elaborated for power reactors. Research reactors in operation in France have been built within a 22 year period. Meanwhile, safety rules have been improved. Old reactors do not comply with all the new rules but modifications are continuously made: after analysis of incidents, when replacement of equipment has to be carried out, when an important modification (fuel conversion for example) is decided upon.

ECORA - Evaluation of Computational Methods for Reactor Safety Analysis

International Nuclear Information System (INIS)

Scheuerer, Martina

2002-01-01

There were three motivations behind the ECORA Project: - the shortcomings of 0-D system codes in the simulation of 3-D, local flow and heat transfer phenomena, - increased interest in the application of 3-D CFD software as supplement to system codes, - high safety requirements in the nuclear industry required consistent standards for the use and assessment of CFD software. The purpose of ECORA was therefore: - to establish performance criteria for the assessment of CFD software, - to establish Best Practice Guidelines for application and use of CFD software, with the following objectives: - assessment of CFD applications in reactor safety: flows in containment (PANDA experiments) and flows in primary system (UPTF experiments) - Best Practice Guidelines for reactor safety: starting point (ERCOFTAC Best Practice Guidelines), adaptation to CFD application for nuclear safety, extension to assessment of experimental data - recommendations for improvements of CFD software, - network of European 'Centres of Competence for CFD Applications in Reactor Safety'. Currently, there were twelve partners in the ECORA Project, representing nine European countries. The Project was scheduled to last until September 2004. Ms Scheuerer then described the work programme and project structure, the Best Practice Guidelines for CFD simulations, the procedures for quantifying errors, applications of Best Practice Guidelines, Best Practice Guidelines for experimental data, applications to primary system, UPTF and PANDA data. Her conclusions were the following: - the Project had led to the improvement of the quality of CFD calculations in reactor safety, through: the ECORA Best Practice Guidelines, the assessment of shortcomings and the improvement of mathematical models. - It had also led to higher acceptance of CFD in reactor safety. - The next step was the establishment of European 'Centres of Competence for CFD Applications in reactor Safety'

Safety design analyses of Korea Advanced Liquid Metal Reactor

International Nuclear Information System (INIS)

Suk, S.D.; Park, C.K.

2000-01-01

The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2003

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

The work of the institute is directed to the assessment and enhancement of the safety of technical plants and to the increase of the efficiency and environmental sustainability of those facilities. Subjects of investigation are equally nuclear plants and installations of process industries. To achieve the goals mentioned, the institute is mainly engaged in the scientific fields of thermal fluid dynamics including magneto-hydrodynamics (MHD) and materials sciences. In 2003, the ISR worked on the following main scientific projects. Sub-programme: Plant and Rector Safety. Project: accident analysis of nuclear reactors, safety of materials and components, particle and radiation transport, safety and efficiency of chemical processes. Sub-programme: Thermal Fluid Dynamics. Project: magneto-hydrodynamics, thermal fluid dynamics of multiphase systems. Considerable progress could also be achieved in the CFD simulation of two-phase flows. New approaches for the forces acting on steam bubbles in a water flow could be developed and implemented into the CFX code in close cooperation with the CFX developer ANSYS/CFX. The qualified models allow to simulate the evolution of bubble size specific radial void distribution profiles along the flow path. These theoretical studies and the related experiments at the Rossendorf TOPFLOW test facility represent an important part of the German CFD network that aims at the improvement of thermal hydraulic calculation methods in reactor safety. (orig.)

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2003

International Nuclear Information System (INIS)

2004-01-01

The work of the institute is directed to the assessment and enhancement of the safety of technical plants and to the increase of the efficiency and environmental sustainability of those facilities. Subjects of investigation are equally nuclear plants and installations of process industries. To achieve the goals mentioned, the institute is mainly engaged in the scientific fields of thermal fluid dynamics including magneto-hydrodynamics (MHD) and materials sciences. In 2003, the ISR worked on the following main scientific projects. Sub-programme: Plant and Rector Safety. Project: accident analysis of nuclear reactors, safety of materials and components, particle and radiation transport, safety and efficiency of chemical processes. Sub-programme: Thermal Fluid Dynamics. Project: magneto-hydrodynamics, thermal fluid dynamics of multiphase systems. Considerable progress could also be achieved in the CFD simulation of two-phase flows. New approaches for the forces acting on steam bubbles in a water flow could be developed and implemented into the CFX code in close cooperation with the CFX developer ANSYS/CFX. The qualified models allow to simulate the evolution of bubble size specific radial void distribution profiles along the flow path. These theoretical studies and the related experiments at the Rossendorf TOPFLOW test facility represent an important part of the German CFD network that aims at the improvement of thermal hydraulic calculation methods in reactor safety. (orig.)

Safety issues at the defense production reactors

International Nuclear Information System (INIS)

1987-01-01

The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors - the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. This report examines the safety objective established by the Department of Energy for the production reactors and the process the Department of its contractors use to implement the objective; focuses on a variety of uncertainties concerning the production reactors, particularly those related to potential vulnerabilities to severe accidents; and identifies ways in which the DOE approach to management of the safety of the production reactors can be improved

Passive safety optimization in liquid-sodium cooled reactors

International Nuclear Information System (INIS)

Cahalan, J. E.; Hahn, D.; Chang, W.-P.; Kwon, Y.-M.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

2004-01-01

This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4)

Safety aspects of designs for future light water reactors (evolutionary reactors)

International Nuclear Information System (INIS)

1993-07-01

The main purpose of this document is to describe the major innovations of proposed designs of future light water reactors, to describe specific safety characteristics and safety analysis methodologies, and to give a general overview of the most important safety aspects related to future reactors. The reactors considered in this report are limited to those intended for fixed station electrical power production, excluding most revolutionary concepts. More in depth discussion is devoted to those designs that are in a more advanced state of completion and have been more extensively described and analysed in the open literature. Other designs will be briefly described, as evidence of the large spectrum of new proposals. Some designs are similar; others implement unique features and require specific discussion (not all aspects of designs with unique features are fully discussed in this document). 131 refs, 22 figs

Some considerations for assurance of reactor safety from experiences in research reactors

International Nuclear Information System (INIS)

Okamoto, Sunao; Nishihara, Hideaki; Shibata, Toshikazu

1981-01-01

For the purpose of assuring reactor safety and strengthening research in the related fields, a multi-disciplinary group was formed among university researchers, including social scientists, with a special allocation of the Grant-in-Aid from the Ministry of Education, Science and Culture. An excerpt from the first year's report (1979 -- 1980) is edited here, which contains an interpretation of Murphy's reliability engineering law, a scope of reactor diagnostic studies to be pursued at universities, and safety measures already implemented or suggested to be implemented in university research reactors. (author)

Considerations in the development of safety requirements for innovative reactors: Application to modular high temperature gas cooled reactors

International Nuclear Information System (INIS)

2003-08-01

Member States of the IAEA have frequently requested this organization to assess, at the conceptual stage, the safety of the design of nuclear reactors that rely on a variety of technologies and are of a high degree of innovation. However, to date, for advanced and innovative reactors and for reactors with characteristics that are different from those of existing light water reactors, widely accepted design standards and rules do not exist. This TECDOC is an outcome of the efforts deployed by the IAEA to develop a general approach for assessing the safety of the design of advanced and innovative reactors, and of all reactors in general including research reactors, with characteristics that differ from those of light water reactors. This publication puts forward a method for safety assessment that is based on the well established and accepted principle of defence in depth. The need to develop a general approach for assessing the safety of the design of reactors that applies to all kinds of advanced reactors was emphasized by the request to the IAEA by South Africa to review the safety of the South African pebble bed modular reactor. This reactor, as other modular high temperature gas cooled reactors (MHTGRs), adopts very specific design features such as the use of coated particle fuel. The characteristics of the fuel deeply affect the design and the safety of the plant, thereby posing several challenges to traditional safety assessment methods and to the application of existing safety requirements that have been developed primarily for water reactors. In this TECDOC, the MHTGR has been selected as a case study to demonstrate the viability of the method proposed. The approach presented is based on an extended interpretation of the concept of defence in depth and its link with the general safety objectives and fundamental safety functions as set out in 'Safety of Nuclear Power Plants: Design', IAEA Safety Standards No. NS-R.1, issued by the IAEA in 2000. The objective

Safety inspections to TRIGA reactors

International Nuclear Information System (INIS)

Byszewski, W.

1988-01-01

The operational safety advisory programme was created to provide useful assistance and advice from an international perspective to research reactor operators and regulators on how to enhance operational safety and radiation protection on their reactors. Safety missions cover not only the operational safety of reactors themselves, but also the safety of associated experimental loops, isotope laboratories and other experimental facilities. Safety missions are also performed on request in other Member States which are interested in receiving impartial advice and assistance in order to enhance the safety of research reactors. The results of the inspections have shown that in some countries there are problems with radiation protection practices and nuclear safety. Very often the Safety Analysis Report is not updated, regulatory supervision needs clarification and improvement, maintenance procedures should be more formalised and records and reports are not maintained properly. In many cases population density around the facility has increased affecting the validity of the original safety analysis

UK experience of safety requirements for thermal reactor stations

International Nuclear Information System (INIS)

Matthews, R.R.; Dale, G.C.; Tweedy, J.N.

1977-01-01

The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to

Learning from Fukushima: Institutional Isomorphism as Constraining and Contributing Nuclear Safety

International Nuclear Information System (INIS)

Ylönen, M.

2016-01-01

This paper is an analysis of the international institutional isomorphic pressures and lessons learned from the Fukushima accident. The recent upgrading of nuclear safety requirements at the international and national level, as well as harmonisation attempts of nuclear reactor safety by the Western European Nuclear Regulators’ Association (WENRA), show serious efforts to improve nuclear safety and implement lessons learned from the Fukushima accident. After Fukushima new requirements for the new nuclear power plants were set, such as preparedness for natural hazards, multiple failure and core melt situations. In addition, improvement of safety culture was emphasised, as well as strengthening of independence of the regulatory body from external pressures, and increasing of independence between different levels of defence in depth safety. However, learning from accidents is often affected by institutional factors, which may both contribute and hamper safety and learning.

Swimming pool reactor reliability and safety analysis

International Nuclear Information System (INIS)

Li Zhaohuan

1997-01-01

A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data

Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Reactor safety assessment system

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

International Nuclear Information System (INIS)

Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

2004-01-01

Full text: The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by (1) implementing safety upgrades, or (2) assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

International Nuclear Information System (INIS)

Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

2004-01-01

The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by implementing safety upgrades, or assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-15

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

Prioritization of R and D programs on probabilistic reactor safety

International Nuclear Information System (INIS)

Husseiny, A.A.

1982-01-01

An interactive computer code based on the multiattribute utility theory has been developed with graphic capabilities to use in selection of probabilistic reactor safety RandD programs. Utility values and proper graphic representation are made through lottery games on the computer terminal. The code is applied to prioritize a set of RandD programs on LWR safety based on attributes including regulatory issues, institutional issues and operation problems. The methodology is described here in detail with its applications. Some of the input includes statistical distributions and subjective judgments on institutional issues. The flexibility of the approach provides a tool for decision makers whether on individual or group level to assess LWR safety priorities and continuously update their strategies

Safety analysis calculations for research and test reactors

Energy Technology Data Exchange (ETDEWEB)

Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

1983-08-01

The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

Safety design philosophy of gas turbine high temperature reactor (GTHTR300)

International Nuclear Information System (INIS)

Katanishi, Shoji; Kunitomi, Kazuhiko

2003-01-01

Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300 based on the experience of the High Temperature Engineering Test Reactor (HTTR) of JAERI which is the first High Temperature Gas-cooled Reactor (HTGR) in Japan. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident induced by a large pipe break is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of the depressurization accident. The safety design philosophies for passive cooling system, reactor shutdown system, and so on were determined. The methodology for the safety evaluation, such as safety criteria and selection of events to be evaluated by using estimation of probability of occurrence, were also discussed and determined. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Syarip, S., E-mail: syarip@batan.go.id [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

2014-10-15

The self-assessment of safety culture and safety management status of Kartini research reactor is a step to foster safety culture and management by identifying good practices and areas for improvement, and also to improve reactor safety in a whole. The method used in this assessment is based on questionnaires provided by the Forum for Nuclear Cooperation in Asia (FNCA), then reviewed by experts. Based on the assessment and evaluation results, it can be concluded that there were several good practices in maintaining the safety status of Kartini reactor such as: reactor operators and radiation protection workers were aware and knowledgeable of the safety standards and policies that apply to their operation, readily accept constructive criticism from their management and from the inspectors of regulatory body that address safety performance. As a proof, for the last four years the number of inspection/audit findings from Regulatory Body (BAPETEN) tended to decrease while the reactor utilization and its operating hour increased. On the other hands there were also some comments and recommendations for improvement of reactor safety culture, such as that there should be more frequent open dialogues between employees and managers, to grow and attain a mutual support to achieve safety goals. (author)

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

Performance Monitoring for Nuclear Safety Related Instrumentation at PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2015-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on performance monitoring for nuclear safety related instrumentation in TRIGA PUSPATI Reactor (RTP) of based on various parameter of reactor safety instrument channel such as log power, linear power, Fuel temperature, coolant temperature will take into consideration. Methodology of performance on estimation and monitoring is to evaluate and analysis of reactor parameters which is important of reactor safety and control. And also to estimate power measurement, differential of log and linear power and fuel temperature during reactor start-up, operation and shutdown .This study also focus on neutron power fluctuation from fission chamber during reactor start-up and operation. This work will present result of performance monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that performance of nuclear safety related instrumentation will improved the reactor control and safety parameter during reactor start-up, operation and shutdown. (author)

Monitoring circuit for reactor safety systems

Science.gov (United States)

Keefe, Donald J.

1976-01-01

The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned.

Safety classification of systems, structures, and components for pool-type research reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-08-15

Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

Self-assessment of application of the Code of Conduct on the safety of research reactors - Mexico

International Nuclear Information System (INIS)

Mamani-Alegria, Y.R.; Salgado-Gonzalez, J.R.; Miranda-Aldaco, J.

2009-01-01

In Mexico, the nuclear regulatory body is the National Commission on Nuclear Safety and Safeguards (CNSNS), and there is one research reactor, a TRIGA MARK III, operated by the National Institute for Nuclear Research (ININ). The main aspects of the Self-assessment of application of The Code of Conduct on the Safety of Research Reactor are given for the case of the TRIGA reactor. Furthermore, in this paper we give a brief description of the legal framework of the licensing process, for nuclear activities in a research reactor, there are also highlights of the major reactor features, the uses of the reactor for isotope production, the management and verification of safety, the radiation protection management program, the emergency planning and the training and qualification of the operation personnel. (author)

Conceptual safety design analysis of Korea advanced liquid metal reactor

International Nuclear Information System (INIS)

Suk, S. D.; Park, C. K.

1999-01-01

The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

Safety of research reactors - A regulator's perspective

International Nuclear Information System (INIS)

Rahman, M.S.

2001-01-01

Due to historical reasons research reactors have received less regulatory attention in the world than nuclear power plants. This has given rise to several safety issues which, if not addressed immediately, may result in an undesirable situation. However, in Pakistan, research reactors and power reactors have received due attention from the regulatory authority. The Pakistan Research Reactor-1 has been under regulatory surveillance since 1965, the year of its commissioning. The second reactor has also undergone all the safety reviews and checks mandated by the licensing procedures. A brief description of the regulatory framework, the several safety reviews carried out have been briefly described in this paper. Significant activities of the regulatory authority have also been described in verifying the safety of research reactors in Pakistan along with the future activities. The views of the Pakistani regulatory authority on the specific issues identified by the IAEA have been presented along with specific recommendations to the IAEA. We are of the opinion that there are more Member States operating nuclear research reactors than nuclear power plants. Therefore, there should be more emphasis on the research reactor safety, which somehow has not been the case. In several recommendations made to the IAEA on the specific safety issues the emphasis has been, in general, to have a similar documentation and approach for maintaining and verifying operational safety at research reactors as is currently available for nuclear power reactors and may be planned for nuclear fuel cycle facilities. (author)

List of reports on reactor safety research from BMFT, CEA, EPRI, JSTA and USNRC

International Nuclear Information System (INIS)

1985-08-01

This list reviews reports from the Federal Republic of Germany, from France, from Japan and from the United States of America concerning special problems in the field of reactor safety research. According to the cooperation of the Federal Ministry for Research and Technology (BMFT) with the Commissariat a l'Energie Atomique (CEA), the Japan Science and Technology Agency (JSTA), the Electric Power Research Institute (EPRI) and the United States Nuclear Regulatory Commission, these reports are available from the Society for Reactor Safety (GRS). The list pursues the following order: Country of origin, problem area concerned, according to the Reactor Safety Research Program of the BMFT, reporting organization. The list of reports appears quarterly. (orig./HP) [de

Overview of research potential of Institute for Nuclear Research

International Nuclear Information System (INIS)

Ciocanescu, Marin

2007-01-01

The main organizations involved in nuclear power production in Romania, under supervision of Presidency, Prime Minister and Parliament are: CNCAN (National Commission for Nuclear Activities Control), Nuclear Agency, Ministry of Economy and Commerce, ANDRAD (Waste Management Agency), SNN (Nuclearelectrica National Society), RAAN (Romanian Authority for Nuclear Activities), ICN (Institute for Nuclear Research - Pitesti), SITON (Center of Design and Engineering for Nuclear Projects- Bucharest); ROMAG-PROD (Heavy Water Plant), CNE-PROD (Cernavoda Nuclear Power Plant - Production Unit), CNE-INVEST (Cernavoda Nuclear Power Plant -Investments Unit), FCN (Nuclear Fuel Factory). The Institute for Nuclear Research, Pitesti INR, Institute for Nuclear Research, Pitesti is endowed with a TRIGA Reactor, Hot Cells, Materials Laboratories, Nuclear Fuel, Nuclear Safety Laboratories, Nuclear Fuel, Nuclear Safety. Waste management. Other research centers and laboratories implied in nuclear activities are: ICIT, National Institute for cryogenics and isotope technologies at Rm Valcea Valcea. with R and D activity devoted to heavy water technologies, IFIN, Institute for nuclear physics and engineering, Bucharest, as well as the educational institutions involved in atomic energy applications and University research, Politechnical University Bucharest, University of Bucharest, University of Pitesti, etc. The INR activity outlined, i.e. the nuclear power research as a scientific and technical support for the Romanian nuclear power programme, mainly dedicated to the existing NPP in the country (CANDU). Focused with priority are: - Nuclear Safety (behavior of plant materials, components, installations during accident conditions and integrity investigations); - Radioactive Waste Management Radioactive; - Radioprotection; Product and services supply for NPP. INR Staff numbers 320 R and D qualified and experienced staff, 240 personnel in devices and prototype workshops and site support

Monitoring circuit for reactor safety systems

International Nuclear Information System (INIS)

Keefe, D.J.

1976-01-01

The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned. 3 claims, 2 figures

Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1989-01-01

Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

Paul Scherrer Institut annual report 1995. Annex IV: PSI nuclear energy and safety

Energy Technology Data Exchange (ETDEWEB)

Birchley, J.; Roesel, R.; Doesburg, R. van [eds.] [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1996-09-01

Nuclear energy research in Switzerland is concentrated at PSI`s Department F4. It is explicitly mentioned in the Institute`s official charter and commands about one fifth of the Institute`s federal resources. Presently, PSI invests approx. 200 py/a in nuclear energy research, one third of this being externally funded; the share of external funding in investment costs totals approx. 50%. This funding is provided by the Swiss utilities and the NAGRA, the Safety Authority (HSK) and the former National Fund for Energy Research (NEFF). PSI`s activities in nuclear research concentrate on three main areas: safety of operating plants, safety features of future reactor concepts and waste management. 7% of personnel are invested in addressing global aspects of energy. (author) figs., tabs., refs.

Safety program considerations for space nuclear reactor systems

International Nuclear Information System (INIS)

Cropp, L.O.

1984-08-01

This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1987-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1988-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicate but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: Nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat. (orig.)

Netherlands Interuniversity Reactor Institut

International Nuclear Information System (INIS)

1978-01-01

This is the annual report of the Interuniversity Reactor Institute in the Netherlands for the Academic Year 1977-78. Activities of the general committee, the daily committee and the scientific advice board are presented. Detailed reports of the scientific studies performed are given under five subjects - radiation physics, reactor physics, radiation chemistry, radiochemistry and radiation hygiene and dosimetry. Summarised reports of the various industrial groups are also presented. Training and education, publications and reports, courses, visits and cooperation with other institutes in the area of scientific research are mentioned. (C.F.)

Safety and regulatory researches on the SMART reactor

International Nuclear Information System (INIS)

Seul, Kwang Won; Kim, Wee Kyong; Chang, Moo Hee

2000-01-01

The 330 MW thermal power of integral pressurized water reactor, named SMART (System integrated Modular Advanced ReacTor), is under development at the Korea Atomic Energy Research Institute (KAERI) for seawater desalination application and electricity generation. The plant is expected to install near the population zone. Thus, the public around the plant should be in depth protected from the possible release of radioactive materials, and also the fresh water should be prevented from radioactivity contamination. Currently, in parallel with the design development, the regulatory research is being conducted to identify and resolve the safety concerns of the nuclear desalination plant. Until now, some general items to be considered in the safety aspects have been identified for the conceptual design of SMART. They include the use of proven technology, application of strengthening defense-in-depth, event categorization and selection, effects of desalination plant, and maintainability of major components. These cooperative researches with regulatory body in the design stage are expected to provide an opportunity to early resolve the safety concerns and eventually the licensing stability of the SMART design. (author)

The human factors and the safety of experimentation reactors

International Nuclear Information System (INIS)

Jeffroy, F.; Delaporte-Normier, M.L.

2007-01-01

Inside IRSN (Institute for Radiological protection and Nuclear Safety), the mission of the Human Factors Group is to assess the way operators of nuclear installations take into account the risks related to human activities. In the last few years, IRSN has been involved in the safety analysis of different installations where Cea develops research programs, in particular experimental reactors. The first part of this article presents the methodology used by IRSN to evaluate how operators take into account risks related to human activities. This methodology is made up of 4 steps: 1) the identification of the human activities that convey a risk for the installation nuclear safety (safety-sensitive activities), for instance in the case of the Masurca reactor, it has been shown that errors made during the manufacturing of fuel tubes can lead to a criticality accident; 2) listing all the dispositions or arrangements taken to make human safety-sensitive activities more reliable; 3) checking the efficiency of such dispositions or arrangements; and 4) assessing the ability of the operators to generate the adequate dispositions or arrangements. The second part highlights the necessity to develop inside these research installations an organisation that facilitates cooperation between experimenters and operators

Annual report 90. Institute for advanced materials

International Nuclear Information System (INIS)

1991-01-01

The Annual Report 1990 of the Institute for Advanced Materials of the JRC highlights the Scientific Technical Achievements and presents in the Annex the Institute's Competence and Facilities available to industry for services and research under contract. The Institute executed in 1990 the R and D programme on advanced materials of the JRC and contributed to the programmes: reactor safety, radio-active waste management, fusion technology and safety, nuclear fuel and actinide research. The supplementary programme: Operation of the High Flux Reactor is presented in condensed form. A full report is published separately

Simplified safety and containment systems for the iris reactor

International Nuclear Information System (INIS)

Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

2001-01-01

The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

Passive and inherent safety technologies for light-water nuclear reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1990-07-01

Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs

Safety philosophy and safety technology of the Soviet RBMK reactors

International Nuclear Information System (INIS)

Zuend, H.; Jarvis, A.S.; Haennis, H.P.; Tikal, J.

1986-01-01

Safety requirements and control in USSR are outlined. Safety criteria and practical application in the case of the RBMK type reactor Chernobyl-4 are discussed. An overview of the Chernobyl-4 reactor accident including its causes is given. Measures to improve the safety of RBMK reactors are described

Philosophy of safety evaluation on fast breeder reactor

International Nuclear Information System (INIS)

1981-01-01

This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

Self Assessment for the Safety of Research Reactor in Indonesia

International Nuclear Information System (INIS)

Melani, Ai; Chang, Soon Heung

2008-01-01

At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. National Nuclear Energy Agency (BATAN) has three research reactor; which are: Reactor Triga Mark II at Bandung, Reactor Kartini at Yogyakarta and Reactor Serbaguna (Multi Purpose Reactor) at Serpong. The Code of Conduct on the Safety of Research Reactors establishes 'best practice' guidelines for the licensing, construction and operation of research reactors. In this paper the author use the requirement in code of conduct to review the safety of research reactor in Indonesia

Safety assessment of research reactors and preparation of the safety analysis report

International Nuclear Information System (INIS)

1994-01-01

This Safety Guide presents guidelines, approved by international consensus, for the preparation, review and assessment of safety documentation for research reactors such as the Safety Analysis Report. While the Guide is most applicable to research reactors in the design and construction stage, it is also recommended for use during relicensing or reassessment of existing reactors

The safety of light water reactors

International Nuclear Information System (INIS)

Pershagen, B.

1986-04-01

The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

Recommended safety objectives, principles and requirements for mini-reactors

International Nuclear Information System (INIS)

1991-05-01

Canadian and international publications containing objectives, principles and requirements for the safety of nuclear facilities in general and nuclear power plants in particular have been reviewed for their relevance to mini-reactors. Most of the individual recommendations, sometimes with minor wording changes, are applicable to mini-reactors. However, some prescriptive requirements for the shutdown, emergency core cooling and containment systems of power reactors are considered inappropriate for mini-reactors. The Advisory Committee on Nuclear Safety favours a generally non-prescriptive approach whereby the applicant for a mini-reactor license is free to propose any means of satisfying the fundamental objectives, but must convince the regulatory agency to that effect. To do so, a probabilistic safety assessment (PSA) would be the favoured procedure. A generic PSA for all mini-reactors of the same design would be acceptable. Notwithstanding this non-prescriptive approach, the ACNS considers that it would be prudent to require the existence of at least one independent shutdown system and two physically independent locations from which the reactor can be shut down and the shutdown condition monitored, and to require provision for an assumed loss of integrity of the primary cooling system's boundary unless convincing arguments to the contrary are presented. The ACNS endorses in general the objectives and fundamental principles proposed by the interorganizational Small Reactor Criteria working group, and intends to review and comment on the documents on specific applications to be issued by that working group

EDF's nuclear safety approach for pressurized water reactors

International Nuclear Information System (INIS)

Tanguy, P.; Kus, J.P.

1987-01-01

The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

EDF'S nuclear safety approach for pressurized water reactors

International Nuclear Information System (INIS)

Tanguy, P.; Kus, J.P.

1988-01-01

The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction - had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's white book on nuclear safety

Safety issues at the defense production reactors

International Nuclear Information System (INIS)

1987-01-01

The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors endash the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. The report is necessarily based on a limited review of the defense production reactors. It does not address whether any of the reactors are ''safe,'' because such an analysis would involve a determination of acceptable risk endash a matter of obvious importance, but one that was beyond the purview of the committee. It also does not address whether the safety of the production reactors is comparable to that of commercial nuclear power stations, because even this narrower question extended beyond the charge to the committee and would have involved detailed analyses that the committee could not undertake

Safety analysis of the present status of the research reactor 'RA' at 'Vinca' Institute

International Nuclear Information System (INIS)

Jovic, V.; Jovic, L.; Zivotic, Z.; Milovanovic, Dj.

1995-01-01

Safety analysis of the nuclear facility which has been out of work for a long time and whose future is not defined at the present moment, can not be connected to the usual, normatively regulated system analysis procedure in both operational and accidental regimes. Therefore, the safety analysis of the present status of the present status of the reactor RA is related to system and components analysis which, in present conditions maintain their nuclear functions operational. In the first place, it refers to components and equipment in which radioactive radiation generation still exists and to installations and equipment maintaining radiation level below permitted limit. in the context of the analysis the following areas are being covered: present status characteristics, accidental events while operating period from 1959. to 1984., nuclear fuels and radioactive waste inventory, basic characteristics and status of safety-related systems and equipment, radiation protection, potential accident analysis at present status of the reactor RA, potential accidental situations due to natural events (earthquakes, water flood) or man-induced events and security. 8 refs

Unique differences in applying safety analyses for a graphite moderated, channel reactor

International Nuclear Information System (INIS)

Moffitt, R.L.

1993-06-01

Unlike its predecessors, the N Reactor at the Hanford Site in Washington State was designed to produce electricity for civilian energy use as well as weapons-grade plutonium. This paper describes the major problems associated with applying safety analysis methodologies developed for commercial light water reactors (LWR) to a unique reactor like the N Reactor. The focus of the discussion is on non-applicable LWR safety standards and computer modeling/analytical variances of standards. The approaches used to resolve these problems to develop safety standards and limits for the N Reactor are described

Research reactor safety - an overview of crucial aspects

International Nuclear Information System (INIS)

Laverie, M.

1998-01-01

Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

The safety of the fast reactor

International Nuclear Information System (INIS)

Matthews, R.R.

1977-01-01

Verbatim of an address by R.R. Matthews, Chief Nuclear Health and Safety Officer, UK Central Electricity Generating Board given on January 15th 1977. The object of this address was to give some opinions on the safety issues of fast reactors as seen from an operational point of view. An outline of the basic responsibilities for nuclear safety is first given, and it is emphasized that the Central Electricity Generating Board has a statutory responsibility for the safe operation of its nuclear plant. The Nuclear Installations Act places absolute responsibility on the operator for ensuring that injury to persons and damage to property do not occur, and the new Health and Safety at Work Act does likewise. In addition the Board has a Nuclear Health and Safety Department that has to ensure that adequate provision for safety is made in the design, construction, and operation of nuclear plant, and safety at operational stations is monitored continuously by inspectors. In addition the requirements of the Nuclear Installations Inspectorate, laid down in the site licence conditions, must be satisfied. All these requirements are here discussed in the light of application to commercial fast reactors. It is considered that the hazards to fast reactor operating personnel are small and little different from those of other types of reactor, and in some respects the fast reactor has advantages, particularly in regard to the use of a Na coolant. The possibility of various types of accident is considered. Radioactive effluent discharge is also considered. The fast reactor as an international problem is discussed, including security matters. The extensive experience gained in operation of the experimental and prototype fast reactors at Dounreay is emphasized. (U.K.)

An intelligent safety system concept for future CANDU reactors

International Nuclear Information System (INIS)

Hinds, H.W.

1980-01-01

A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)

Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

International Nuclear Information System (INIS)

1987-05-01

Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions

Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

International Nuclear Information System (INIS)

Ridha Roslan; Nik Mohd Faiz Khairuddin

2013-01-01

Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)

Activities on safety for the cross-cutting issue of research reactors in the IAEA

International Nuclear Information System (INIS)

Perrotta, J.A.; Boado Magan, H.J.

2003-01-01

IAEA activities in the field of research reactor safety are included in the programme of the Division of Nuclear Installations Safety and implemented by the Engineering Safety Section through its Research Reactor Safety Unit. Following the objectives of the Division, the results of the IAEA missions and the recommendations from International Advisory Groups, the IAEA has conducted in recent years a certain number of activities aiming to enhance the safety of research reactors. The following activities are discussed in this paper: (a) the new Requirements for the Safety of Research Reactors, main features and differences with previous standards (SS-35-S1 and SS-35-S2) and the grading approach for implementation; (b) new documents being developed (safety guides, safety reports and TECDOCs); (c) activities related to the Incident Reporting System for Research Reactor (IRSRR); (d) the new features implemented for the (Integrated Safety Assessment of Research Reactors) INSARR missions; (e) the Code of Conduct on the Safety of Research Reactors developed, following the General Conference Resolution GC(45)/RES/10; and (f) the survey on the safety of research reactors conducted in the year 2002 and the results obtained. (author)

Reactor design and safety approach for a tank-type fast reactor

International Nuclear Information System (INIS)

Davies, S.M.; Yamaki, Hideo; Goodman, L.

1984-06-01

A tank type plant has been designed that offers compactness, high reliability under seismic and thermal transients, and a safety design approach that provides a balance between public safety and plant availability. This report provides a description of the design philosophy and safety features of the reactor

Safety infrastructure for countries establishing their first research reactor

International Nuclear Information System (INIS)

Abou Yehia, H.; Shokr, A.M.

2010-01-01

Establishment of a research reactor is a major project requiring careful planning, preparation, implementation, and investment in time and human resources. The implementation of such a project requires establishment of sustainable infrastructures, including legal and regulatory, safety, technical, and economic. An analysis of the needs for a new research reactor facility should be performed including the development of a utilization plan and evaluation of site availability and suitability. All these elements should be covered by a feasibility study of the project. This paper discusses the elements of such a study with the main focus on the specific activities and steps for developing the necessary safety infrastructure. Progressive involvement of the main organizations in the project, and application of the IAEA Code of Conduct on the Safety of Research Reactors and IAEA Safety Standards in different phases of the project are presented and discussed. (author)

Sensitivity analysis of reactor safety parameters with automated adjoint function generation

International Nuclear Information System (INIS)

Kallfelz, J.M.; Horwedel, J.E.; Worley, B.A.

1992-01-01

A project at the Paul Scherrer Institute (PSI) involves the development of simulation models for the transient analysis of the reactors in Switzerland (STARS). This project, funded in part by the Swiss Federal Nuclear Safety Inspectorate, also involves the calculation and evaluation of certain transients for Swiss light water reactors (LWRs). For best-estimate analyses, a key element in quantifying reactor safety margins is uncertainty evaluation to determine the uncertainty in calculated integral values (responses) caused by modeling, calculational methodology, and input data (parameters). The work reported in this paper is a joint PSI/Oak Ridge National Laboratory (ORNL) application to a core transient analysis code of an ORNL software system for automated sensitivity analysis. The Gradient-Enhanced Software System (GRESS) is a software package that can in principle enhance any code so that it can calculate the sensitivity (derivative) to input parameters of any integral value (response) calculated in the original code. The studies reported are the first application of the GRESS capability to core neutronics and safety codes

Research reactor safety - an overview of crucial aspects

Energy Technology Data Exchange (ETDEWEB)

Laverie, M. [Atomic Energy Commission, Saclay, F-91191 Gif sur Yvette (France)

1998-07-01

Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

Reactor operation safety information document

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

RSAS: a Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Dixon, B.W.; Bray, M.A.

1985-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

Verification of reactor safety codes

International Nuclear Information System (INIS)

Murley, T.E.

1978-01-01

The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

Nuclear reactor safety research in Idaho

International Nuclear Information System (INIS)

Zeile, H.J.

1983-01-01

Detailed information about the performance of nuclear reactor systems, and especially about the nuclear fuel, is vital in determining the consequences of a reactor accident. Fission products released from the fuel during accidents are the ultimate safety concern to the general public living in the vicinity of a nuclear reactor plant. Safety research conducted at the Idaho National Engineering Laboratory (INEL) in support of the U.S. Nuclear Regulatory Commission (NRC) has provided the NRC with detailed data relating to most of the postulated nuclear reactor accidents. Engineers and scientists at the INEL are now in the process of gathering data related to the most severe nuclear reactor accident - the core melt accident. This paper describes the focus of the nuclear reactor safety research at the INEL. The key results expected from the severe core damage safety research program are discussed

Gas-cooled reactor safety and accident analysis

International Nuclear Information System (INIS)

1985-12-01

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

Passive safety design characteristics of the KALIMER-600 burner reactor

International Nuclear Information System (INIS)

Kwon, Young-Min; Jeong, Hae-Yong; Cho, Chung-Ho; Ha, Ki-Seok; Kim, Sang-Ji

2009-01-01

The Korea Atomic Energy Research Institute (KAERI) has recently studied several burner core designs for a transuranics (TRU) transmutation based on the breakeven core geometry of KALIMER-600. The KALIMER-600 is a net electrical rating of 600MWe, sodium-cooled, metallic-fueled, pool-type reactor. For the burner core concept selected for the present analysis, the smearing fractions of the fuel rods in three fuel zones are changed while maintaining the cladding outer diameter and cladding thickness. The resulting fuel slug smearing fractions of the inner, middle, and outer core zones are 36%, 40%, and 48%, respectively. The TRU conversion ratio is 0.57 and the TRU enrichment of the driver fuel is set to 30.0 w/o because of the current practical limitation of the U-TRU-10%Zr metal fuel database. The purpose of this paper is to evaluate the safety performance characteristics provided by the passive safety design features in the KALIMER-600 burner reactor by using a system-wide safety analysis code. The present scoping analysis focuses on an assessment of the enhanced safety design features that provide passive and self-regulating responses to transient conditions and an evaluation of the safety margin during unprotected overpower, unprotected loss of flow, and unprotected loss of heat sink events. The analysis results show that the KALIMER-600 burner reactor provides larger safety margins with respect to the sodium boiling, fuel rod integrity, and structural integrity. The overall inherent safety can be enhanced by accounting for the reactivity feedback mechanisms in the design process. (author)

A study of passive safety conditions for fast reactor core

International Nuclear Information System (INIS)

Shimizu, Akinao

1991-01-01

A study has been made for passive safety conditions of fast reactor cores. Objective of the study is to develop a concept of a core with passive safety as well as a simple safety philosophy. A simple safety philosophy, which is wore easy to explain to the public, is needed to enhance the public acceptance for nuclear reactors. The present paper describes a conceptual plan of the study including the definition of the problem a method of approach and identification of tasks to be solved

Multimegawatt Space Reactor Safety

International Nuclear Information System (INIS)

Stanley, M.L.

1989-01-01

The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed

Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors. Results from the Coordinated Research Project on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors

International Nuclear Information System (INIS)

2014-09-01

objective of the CRP was to determine a common method for reliability assessment of passive safety system performance. Such a method would facilitate the application of risk informed approaches in design optimization and safety qualification of future advanced reactors, thereby contributing to their enhanced safety levels and improved economics. Five Member States participated, representing seven research institutes and organizations in Argentina, France, India, Italy and the Russian Federation. This publication is the outcome of the different tasks performed and extensive discussions held in the technical meetings, and summarizes the information provided by the technical experts within the CRP over the four year period

Continuous Assessment of Safety Margin for the 14-MW TRIGA Reactor

International Nuclear Information System (INIS)

Ciocanescu, M.; Georgescu, D.; Doru, O.

2008-01-01

The assessment of reactor safety implies analyses of the reactor and its systems response to a range of postulated initiating events (such as malfunction or failures of equipment, operator errors, external events and so on which could lead to either anticipated operational occurrences or to accident conditions. Decreasing in heat removal by the reactor cooling system may be considered as a process disturbance which may lead to a postulated initiating event. The cold source for the reactor cooling system, in case of TRIGA-14 MW reactor is the atmosphere by the secondary cooling towers. The ability to evacuate the heat produced by the reactor core ranges between the outlet temperature of the core flow and the outdoors temperature in air, which is subject to season and day variation. Selected values for safety limits, safety system settings and limiting condition(s) are derived from safety analysis and are consistent with the operational state of the reactor. When a limiting condition for safe operation is not satisfied, the operating personal is supposed to take the appropriate action(s) to ensure safety. Operating requirements and the safety system are presented. The reactor operating safety parameters from the main Data Acquisition System are transferred to an AT personal computer. These selected parameters are the following: - average inlet temperature which is calculated as an average temperature measured by 20 type K thermocouples distributed within a 4 x 5 matrix located on the top of the reactor core; - average outlet temperature which is calculated as an average record from 10 type K thermocouples placed in the outlet pipe; - average flow rate which is calculated as an average value from four transducers (two for the inlet flow rate and two for the outlet flow rate). Due to its high instability, this value is also filtered using a two-pole low-pass filter (software); - reactor thermal power value derivable from the previous parameters or obtained from the

IAEA safety standards and approach to safety of advanced reactors

International Nuclear Information System (INIS)

Gasparini, M.

2004-01-01

The paper presents an overview of the IAEA safety standards including their overall structure and purpose. A detailed presentation is devoted to the general approach to safety that is embodied in the current safety requirements for the design of nuclear power plants. A safety approach is proposed for the future. This approach can be used as reference for a safe design, for safety assessment and for the preparation of the safety requirements. The method proposes an integration of deterministic and risk informed concepts in the general frame of a generalized concept of safety goals and defence in depth. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor including small and medium sized reactors with innovative safety features.(author)

Reactor safety research. The CEC contribution

International Nuclear Information System (INIS)

Krischer, W.

1990-01-01

The involvement of the EC Commission in the reactor safety research dates back almost to the implementation of the EURATOM Treaty and has thus lasted for thirty years. The need for close collaboration and for general consensus on some crucial problems of concern to the public, has made the role of international organizations and, as far as Europe is concerned, the role of the European Community particularly important. The areas in which the CEC has been active during the last five years are widespread. This is partly due to the fact that, after TMI and Chernobyl, the effort and the interest of the different countries in reactor safety was considerable. Reactor Safety Research represents the proceedings of a seminar held by the Commission at the end of its research programme 1984-88 on reactor safety. As such it gives a comprehensive overview of the recent activities and main results achieved in the CEC Joint Research Centre and in national laboratories throughout Europe on the basis of shared cost actions. In a concluding chapter the book reports on the opinions, expressed during a panel by a group of major exponents, on the needs for future research. The main topics addressed are, with particular reference to Light Water Reactors (LWRS): reliability and risk evaluation, inspection of steel components, primary circuit components end-of-life prediction, and abnormal behaviour of reactor cooling systems. As far as LMFBRs are concerned, the topics covered are: severe accident modelling, material properties and structural behaviour studies. There are 67 pages, all of which are indexed separately. Reactor Safety Research will be of particular interest to reliability and safety engineers, nuclear engineers and technicians, and mechanical and structural engineers. (author)

Reports covering research projects in the field of reactor safety supported by the German Federal Ministry for Research and Technology

International Nuclear Information System (INIS)

1976-03-01

Investigations on the safety of Light Water Reactors (LWR) being performed in the framework of the safety program 'Reactor Safety' are sponsored by the Bundesminister fuer Forschung und Technologie (BMFT - Secretary of State for Research and Technology). Objective of this program is to continue improving the safety of LWR, in order to minimize the risk for the environment. With grant assistance from the Bundesminister des Innern. (BMI - Secretary of State for Home Affairs) research cont racts in the field of reactor safety are being performed. Results of these projects should contribute to resolve questions arising nuclear licensing procedures. The Forschungsbetreuung (FB - research supervision department) at the Institute for Reactor Safety (IRS), as consultants to BMFT and BMI, provides information about the progress of investigations. Individual reports will be prepared and put into standard forms by the research contractors. Each report gives information on: 1) the work accomplished, 2) the results obtained, 3) the work planned to be continued. Initial reports of research projects describe in addition the purpose of the work. A BMFT-research program on the safety of Fast Breeders (Schneller Brutreaktor - SBR) is presently under discussion. In order to define several problems, investigations included in the present compilation (RS 139, 140, 143, 162) will be previously performed. (orig.) [de

Safety considerations for research reactors in extended shutdown

International Nuclear Information System (INIS)

2004-01-01

According to the IAEA Research Reactor Database, in the last 20 years, 367 research reactors have been shut down. Of these, 109 have undergone decommissioning and the rest are in extended shutdown with no clear definition about their future. Still other research reactors are infrequently operated with no meaningful utilization programme. These two situations present concerns related to safety such as loss of corporate memory, personnel qualification, maintenance of components and systems and preparation and maintenance of documentation. There are many reasons to shut down a reactor; these may include: - the need to carry out modifications in the reactor systems; - the need for refurbishment to extend the lifetime of the reactor; - the need to repair reactor structures, systems, or components; - the need to remedy technical problems; - regulatory or public concerns; - local conflicts or wars; - political convenience; - the lack of resources. While any one of these reasons may lead to shutdown of a reactor, each will present unique problems to the reactor management. The large variations from one research reactor to the next also will contribute to the uniqueness of the problems. Any option that the reactor management adopts will affect the future of the facility. Options may include dealing with the cause of the shutdown and returning to normal operation, extending the shutdown period waiting a future decision, or decommissioning. Such options are carefully and properly analysed to ensure that the solution selected is the best in terms of reactor type and size, period of shutdown and legal, economic and social considerations. This publication provides information in support of the IAEA safety standards for research reactors

Engineering safety features for high power experimental reactors

International Nuclear Information System (INIS)

Doval, A.; Villarino, E.; Vertullo, A.

2000-01-01

In the present analysis we will focus our attention in the way engineering safety features are designed in order to prevent fuel damage in case of abnormal or accidental situations. To prevent fuel damage two main facts must be considered, the shutdown of the reactor and the adequate core cooling capacity, it means that both, neutronic and thermohydraulic aspects must be analysed. Some neutronic safety features are common to all power ranges like negative feedback reactivity coefficients and the required number of control rods containing the proper absorber material to shutdown the reactor. From the thermohydraulic point of view common features are siphon-breaker devices and flap valves for those powers requiring cooling in the forced convection regime. For the high power reactor group, the engineering safety features specially designed for a generic reactor of 20 MW, will be presented here. From the neutronic point of view besides the common features, and to comply with our National Regulatory Authority, a Second Shutdown System was designed as a redundant shutdown system in case the control plates fail. Concerning thermohydraulic aspects besides the pump flywheels and the flap valves providing the natural convection loop, a metallic Chimney and a Chimney Water Injection System were supplied. (author)

Application of best estimate thermalhydraulic codes for the safety analysis of research reactors

International Nuclear Information System (INIS)

Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.

2006-01-01

An established international expertise in relation to computational tools, procedures for their application including Best Estimate (BE) methods supported by uncertainty evaluation, and comprehensive experimental database exists within the safety technology of Nuclear Power Plant (NPP). The importance of transferring NPP safety technology tools and methods to RR safety technology has been noted in recent IAEA activities. However, the ranges of parameters of interest to RR are different from those for NPP: this is namely true for fuel composition, system pressure, adopted materials and overall system geometric configuration. The large variety of research reactors prevented so far the achievement of systematic and detailed lists of initiating events based upon qualified Probabilistic Safety Assessment (PSA) studies with results endorsed by the international community. However, bounding and generalized lists of events are available from IAEA documents and can be considered for deeper studies in the area. In the area of acceptance criteria, established standards accepted by the international community are available. Therefore no major effort is needed, but an effort appears worthwhile to check that those standards are adopted and that the related thresholds are fulfilled. The importance of suitable experimental assessment is recognized. A large amount of data exists as the kinetic dynamic core behaviour form SPERT reactors tests. However, not all data are accessible to all institutions and the relationship between the range of parameters of experiments and the range of parameters relevant to RR technology is not always established. However, code-assessment through relevant set of experimental data are recorded and properly stored. An established technology exists for development, qualification and application of system thermal-hydraulics codes suitable to be adopted for accident analysis in research reactors. This derives from NPP technology. The applicability of

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Final report-passive safety optimization in liquid sodium-cooled reactors

International Nuclear Information System (INIS)

Cahalana, J. E.; Hahn, D.

2007-01-01

This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

Passive safety systems for integral reactors

International Nuclear Information System (INIS)

Kuul, V.S.; Samoilov, O.B.

1996-01-01

In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs

Passive safety systems for integral reactors

Energy Technology Data Exchange (ETDEWEB)

Kuul, V S; Samoilov, O B [OKB Mechanical Engineering (Russian Federation)

1996-12-01

In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs.

Overview of fusion reactor safety

International Nuclear Information System (INIS)

Cohen, S.; Crocker, J.G.

1981-01-01

Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR

International Nuclear Information System (INIS)

CHENG, L.; HANSON, A.; DIAMOND, D.; XU, J.; CAREW, J.; RORER, D.

2004-01-01

Detailed reactor physics and safety analyses have been performed for the 20 MW D 2 O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core

Nuclear reactor safety systems

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1980-01-01

A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

Energy Technology Data Exchange (ETDEWEB)

Guentay, S [ed.; Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs.

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

International Nuclear Information System (INIS)

Guentay, S.

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs

Safety considerations concerning light water reactors in Sweden

International Nuclear Information System (INIS)

Nilsson, T.

1977-01-01

In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

Evaluation of reactor safety

Energy Technology Data Exchange (ETDEWEB)

NONE

1960-04-15

Although the operation of nuclear reactors has a remarkably good record of safety, the prevention of possible reactor accidents is one of the major factors that atomic planners have to contend with. At the same time, excessive caution may breed an attitude that hampers progress, either by resisting new development or by demanding unnecessarily elaborate and expensive precautions out of proportion to the actual hazards involved. The best course obviously is to determine the possible dangers and adopt adequate measures for their prevention, providing of course, for a reasonable margin of error in judging the hazards and the effectiveness of the measures. The greater the expert understanding and thoroughness with which this is done, the narrower need the margin be. This is the basic idea behind the evaluation of reactor safety

Safety research for evolutionary light water reactors

International Nuclear Information System (INIS)

Cacuci, D.G.

1996-01-01

The development of nuclear energy has been characterized by a continuous evolution of the technological and philosophical underpinnings of reactor safety to enable operation of the plant without causing harm to either the plant operators or the public. Currently, the safety of a nuclear plant is assured through the combined use of procedures and engineered safety features together with a system of multiple protective barriers against the release of radioactivity. This approach is embodied in the concept of Design-Basis Accidents (DBA), which requires the designers to demonstrate that all credible accidents have been identified and that all safety equipment and procedures perform their functions extremely reliably. Particularly important functions are the automatic protection to shut the reactor down and to remove the decay heat while ensuring the integrity of the containment structure. Within the DBA concept, the so-called severe accidents were conveniently defined to be those accidents that lie beyond the DBA envelope; hence, they did not form part of the safety case. (author)

Safety research for evolutionary light water reactors

Energy Technology Data Exchange (ETDEWEB)

Cacuci, D G [Karlsruhe Univ. (T.H.) (Germany). Universitaetsbibliothek

1996-12-01

The development of nuclear energy has been characterized by a continuous evolution of the technological and philosophical underpinnings of reactor safety to enable operation of the plant without causing harm to either the plant operators or the public. Currently, the safety of a nuclear plant is assured through the combined use of procedures and engineered safety features together with a system of multiple protective barriers against the release of radioactivity. This approach is embodied in the concept of Design-Basis Accidents (DBA), which requires the designers to demonstrate that all credible accidents have been identified and that all safety equipment and procedures perform their functions extremely reliably. Particularly important functions are the automatic protection to shut the reactor down and to remove the decay heat while ensuring the integrity of the containment structure. Within the DBA concept, the so-called severe accidents were conveniently defined to be those accidents that lie beyond the DBA envelope; hence, they did not form part of the safety case. (author).

Nuclear safety as applied to space power reactor systems

International Nuclear Information System (INIS)

Cummings, G.E.

1987-01-01

Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety

Distinctive safety aspects of the CANDU-PHW reactor design

International Nuclear Information System (INIS)

Kugler, G.

1980-01-01

Two lectures are presented in this report. They were prepared in response to a request from IAEA to provide information on the 'Special characteristics of the safety analysis of heavy water reactors' to delegates from member states attending the Interregional Training Course on Safety Analysis Review, held at Karlsruhe, November 19 to December 20, 1979. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (auth)

Safety research needs for Russian-designed reactors. Requirements situation

International Nuclear Information System (INIS)

Brown, R. Allan; Holmstrom, Heikki; Reocreux, Michel; Schulz, Helmut; Liesch, Klaus; Santarossa, Giampiero; Hayamizu, Yoshitaka; Asmolov, Vladimir; Bolshov, Leonid; Strizhov, Valerii; Bougaenko, Sergei; Nikitin, Yuri N.; Proklov, Vladimir; Potapov, Alexandre; Kinnersly, Stephen R.; Voronin, Leonid M.; Honekamp, John R.; Frescura, Gianni M.; Maki, Nobuo; Reig, Javier; ); Bekjord, Eric S.; Rosinger, Herbert E.

1998-01-01

In June 1995, an OECD Support Group was set up to perform a broad study of the safety research needs of Russian-designed reactors. The emphasis of the study is on the VVER-type reactors in part because of the larger base of knowledge within the NEA Member countries related to LWRs. For the RBMKs, the study does not make the judgement that such reactors can be brought to acceptable levels of safety but focuses on near term efforts that can contribute to reducing the risk to the public. The need for the safety research must be evaluated in context of the lifetime of the reactors. The principal outcome of the work of the Support Group is the identification of a number of research topics which the members believe should receive priority attention over the next several years if risk levels are to be reduced and public safety enhanced. These appear in the Conclusions and Recommendations section of the report, and are the following: - The most important near-term need for VVER and RBMK safety research is to establish a sound technical basis for the emergency operating procedures used by the plant staff to prevent or halt the progression of accidents (i.e., Accident Management) and for plant safety improvements. - Co-operation of Western and Eastern experts should help to avoid East-West know-how gaps in the future, as safety technology continues to improve. - Safety research in Eastern countries will make an important contribution to public safety as it has in OECD countries. - RBMK safety research, including verification of codes, starts from a smaller base of experience than VVER, and is at an earlier stage of development. Technical Conclusions: - Research to improve human performance and operational safety of VVER and RBMK plants is extremely important. - VVER thermal-hydraulic and reactor physics research should focus on full validation of codes to VVER-specific features, and on extension of experimental data base. - Methods of assessing VVER pressure boundary

Investigation of the possibility of a calculative reactor safety estimation in the licence procedure for nuclear reactors

International Nuclear Information System (INIS)

Adler, B.; Kampf, T.

1975-12-01

Up to now it is impossible to calculate completely the safety of nuclear reactors. Therefore the authors have collected and employed a number of at a high degree independent safety parameters for mathematical evaluation of the reactor safety. By means of computer programs such parameters from about 400 research reactors have been analysed and the fluctuation ranges of their greatest density were determined. The limits of these fluctuation ranges are quickly available and can be used as recommended values for the layout and for the safety estimation of research reactors. A comparison of the existing layout recommendations and the determined fluctuation ranges in most cases shows a good agreement. In some cases corrections and new layout recommendations have been proposed. The determined fluctuation ranges found their first practical application in the estimation of the Rossendorf Equipment for Critical Experiments (RAKE). (author)

Licensing procedures and safety criteria for research reactors in France

International Nuclear Information System (INIS)

Berry, J.L.; Lerouge, B.

1980-11-01

This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. Some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors. At the end, a few considerations are given to the consequences of the Osiris core conversion

The RA nuclear research reactor at VINCA Institute as an engineering and scientific challenge

International Nuclear Information System (INIS)

Mesarovic, M.

1997-01-01

The RA nuclear research at the Vinca Institute of Nuclear Sciences is the largest nuclear research facility in Yugoslavia and belongs to that generation of research reactors which have had an important contribution to nuclear technology development. As these older reactors were generally not built to specific nuclear standards, new safety systems had to be installed at the RA reactor for a renewal of its operating licence in 1984 and it was shut down, after 25 years of operation. Although all the required and several additional systems were built for the restart of the RA reactor, a disruption of foreign delivery of new control equipment caused its conversion to a 'dormant' facility, and it is still out of operation. Therefore, the future status of the RA reactor presents an engineering and scientific challenge to the engineers and scientists from Yugoslavia and other countries that may be interested to participate. To attract their attention on the subject, principal features of the RA reactor and its present status are described in detail, based on a recent engineering economic and safety evaluation. A comparative review of the world research reactors is also presented.(author)

Problems of nuclear reactor safety. Vol. 2

International Nuclear Information System (INIS)

Goncharov, L.A.

1995-01-01

Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

Nuclear Safety Bureau: safety objectives and principles for the proposed ANSTO reactor

International Nuclear Information System (INIS)

Westall, D.

1993-01-01

Siting criteria and safety assessment principles were previously promulgated by the Australian Atomic Energy Commission (AAEC), and have been applied by ANSTO and the Nuclear Safety Bureau (NSB). The NSB is revising these criteria and principles to take account of evolving nuclear safety standards and practices. The NSB Safety and Siting Assessment Principles (SSAP) are presented and it is estimated that it will provide a comprehensive basis for the safety assessment of research reactors in Australia, and be applicable to all stages of a reactor project: siting: design and construction; operation; modification; and decommissioning. The SSAP are similar to the principles promulgated by the AAEC, in that probabilistic safety criteria are set for assessment of design, however these criteria are complimentary to a deterministic design basis approach. This is a similar approach to that recently published by the UK Nuclear Installations Inspectorate 4 . Siting principles are now also included, where they were previously separate, and require a consideration of the consequences of severe accidents which are an extension of accidents catered for by the design of the plant. Criteria for radiation doses due to normal operations and design basis accidents are included in the principles for safety assessment. 9 refs

Nuclear safety requirements for operation licensing of Egyptian research reactors

International Nuclear Information System (INIS)

Ahmed, E.E.M.; Rahman, F.A.

2000-01-01

From the view of responsibility for health and nuclear safety, this work creates a framework for the application of nuclear regulatory rules to ensure safe operation for the sake of obtaining or maintaining operation licensing for nuclear research reactors. It has been performed according to the recommendations of the IAEA for research reactor safety regulations which clearly states that the scope of the application should include all research reactors being designed, constructed, commissioned, operated, modified or decommissioned. From that concept, the present work establishes a model structure and a computer logic program for a regulatory licensing system (RLS code). It applies both the regulatory inspection and enforcement regulatory rules on the different licensing process stages. The present established RLS code is then applied to the Egyptian Research Reactors, namely; the first ET-RR-1, which was constructed and still operating since 1961, and the second MPR research reactor (ET-RR-2) which is now in the preliminary operation stage. The results showed that for the ET-RR-1 reactor, all operational activities, including maintenance, in-service inspection, renewal, modification and experiments should meet the appropriate regulatory compliance action program. Also, the results showed that for the new MPR research reactor (ET-RR-2), all commissioning and operational stages should also meet the regulatory inspection and enforcement action program of the operational licensing safety requirements. (author)

Summary view on demonstration reactor safety

International Nuclear Information System (INIS)

Satoh, Kazuziro; Kotake, Shoji; Tsukui, Yutaka; Inagaki, Tatsutoshi; Miura, Masanori

1991-01-01

This work presents a summary view on safety design approaches for the demonstration fast breeder reactor (DFBR). The safety objective of DFBR is to be at lea as safe as a LWR. Major safety issues discussed in this paper are; reduction of sodium void reactivity worth, adoption of self-actuated mechanism in the backup shutdown system, use of the direct reactor auxiliary cooling system (DRACS), provision of the containment system. (author)

Development of safety analysis technology for integral reactor

International Nuclear Information System (INIS)

Kim, Hee Cheol; Kim, K. K.; Kim, S. H.

2002-04-01

The state-of-the-arts for the integral reactor was performed to investigate the safety features. The safety and performance of SMART were assessed using the technologies developed during the study. For this purpose, the computer code system and the analysis methodology were developed and the safety and performance analyses on SMART basic design were carried out for the design basis event and accident. The experimental facilities were designed for the core flow distribution test and the self-pressurizing pressurizer performance test. The tests on the 2-phase critical flow with non-condensable gas were completed and the results were used to assess the critical flow model. Probabilistic Safety Assessment(PSA) was carried out to evaluate the safety level and to optimize the design by identifying and remedying any weakness in the design. A joint study with KINS was carried out to promote licensing environment. The generic safety issues of integral reactors were identified and the solutions were formulated. The economic evaluation of the SMART desalination plant and the activities related to the process control were carried out in the scope of the study

The safety features of an integrated maritime reactor

International Nuclear Information System (INIS)

Miyakoshi, Junichi; Yamada, Nobuyuki; Kuwahara, Shin-ichi

1975-01-01

The EFDR-80, a typical integrated maritime reactor, which is being developed in West Germany is outlined. The safety features of the integrated maritime reactor are presented with the analysis of reactor accidents and hazards, and are compared with those of the separated maritime reactor. Furthermore, the safety criteria of maritime reactors in Japan and West Germany are compared, and some of the differences are presented from the viewpoint of reactor design and safety analysis. In this report the authors express an earnest desire that the definite and reasonable safety criteria of the integrated maritime reactor should be established and that the safety criteria of the nuclear ship should be standardized internationally. (auth.)

IAEA activities in the field of research reactors safety

International Nuclear Information System (INIS)

Ciuculescu, C.; Boado Magan, H.J.

2004-01-01

IAEA activities in the field of research reactor safety are included in the programme of the Division of Nuclear Installations Safety. Following the objectives of the Division, the results of the IAEA missions and the recommendations from International Advisory Groups, the IAEA has conducted in recent years a certain number of activities aiming to enhance the safety of research reactors. The following activities will be presented: (a) the new Requirements for the Safety of Research Reactors, main features and differences with previous standards (SS-35-S1 and SS-35-S2) and the grading approach for implementation; (b) new documents being developed (safety guides, safety reports and TECDOC's); (c) activities related to the Incident Reporting System for Research Reactor (IRSRR); (d) the new features implemented for the INSARR missions; (e) the Code of Conduct on the Safety of Research Reactors adopted by the Board of Governors on 8 March 2004, following the General Conference Resolution GC(45)/RES/10; and (f) the survey on the safety of research reactors published on the IAEA website on February 2003 and the results obtained. (author)

Safety evaluation of the Dalat research reactor operation

International Nuclear Information System (INIS)

Long, V.H.; Lam, P.V.; An, T.K.

1989-01-01

After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

Safety of research reactors (Design and Operation)

International Nuclear Information System (INIS)

Dirar, H. M.

2012-06-01

The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

2014-01-01

The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

Space nuclear reactor safety

International Nuclear Information System (INIS)

Damon, D.; Temme, M.; Brown, N.

1990-01-01

Definition of safety requirements and design features of the SP-100 space reactor power system has been guided by a mission risk analysis. The analysis quantifies risk from accidental radiological consequences for a reference mission. Results show that the radiological risk from a space reactor can be made very low. The total mission risk from radiological consequences for a shuttle-launched, earth orbit SP-100 mission is estimated to be 0.05 Person-REM (expected values) based on a 1 mREM/yr de Minimus dose. Results are given for each mission phase. The safety benefits of specific design features are evaluated through risk sensitivity analyses

Swedish Nuclear Power Inspectorate, Office of Reactor Safety. Research plans for the period 1997-1999

International Nuclear Information System (INIS)

1997-02-01

Office of Reactor Safety research is carried out within the following areas: Safety evaluation, Safety analysis, MTO, Materials and chemistry, Non-Destructive Testing, Strength of materials, Thermohydraulics, Nuclear fuel, Serious accidents and Process control. Research is carried out to fulfill SKIs overall goals in accordance with the directives from the Swedish government and parliament, in particular to be a driving force in safety related work when justified by operating experience, research results and technical progress, towards licensees as well as in international cooperation in safety; to promote the maintenance and development of competence in the safety related work at the SKI as well as the licensees and generally in the country, and as a specific role for the Office of Reactor Safety as designated in the internal routines to take initiative to encourage and carry out research into areas of importance for the Office as well as ensuring that research results are disseminated and used both within SKI and in the general work concerning nuclear safety. Research efforts within the Office of Reactor safety are carried out in the form of separate projects which form part of the priority work plans. Project managers, the necessary personnel resources and the budget for each year are included in the Annual Plan and the work is followed up in the same manner as other efforts. Research is performed in different ways, that can vary from laboratory studies to more consultative efforts, and be organised in many different ways such as examination projects, post-graduate studies, work sponsored at research institutes and companies in Sweden and abroad, collaboration in larger international projects, and participation in conferences which provide an important contribution to keeping SKI personnel informed within their specialist areas

The French-German common safety approach for future reactors

International Nuclear Information System (INIS)

Birkhofer, A.; Chevet, P.F.

1995-01-01

A common safety approach has been defined for future electronuclear plants in the framework of the French-German European Pressurised water Reactor (EPR) project. Improvements in the domain of containment are required in future reactors conception to prevent any risk of core fusion under high and low pressure. Another objective is to reduce significantly the radioactive releases due to other accidents in order to reduce spatial and temporal environmental and human protection procedures. Protection against external aggressions (plane fall, explosions, earthquakes,..), prevention of pipe rupture in the primary circuits, limitation of hydrogen production in the case of water-zirconium complete reaction, cooling of the reactor in the case of core fusion, and radiologic consequences of accidents are the main points discussed by the French-German safety authorities to define the common safety standards of the EPR project. (J.S.)

Common safety approach for future pressurized water reactors in France and Germany

International Nuclear Information System (INIS)

Queniart, D.; Gros, G.

1994-01-01

In France and Germany all major activities related to future pressurized water reactors are now proceeding in a coordinated way among the two countries. This holds for utilities and industry in the development of a joint PWR project, the ''European Pressurized Water Reactor (EPR)'' by Electricite de France (EDF), German utilities, Nuclear Power International (NPI), Framatome and Siemens as well as for the technical safety objectives for future evolutionary reactors on the basis of a common safety approach adopted by the safety authorities of both countries for plants to operate form the beginning of the next century. The proposed paper covers this common development of a safety approach and particular technical safety objectives. (authors). 5 refs. 1 fig

Molten salt reactors - safety options galore

International Nuclear Information System (INIS)

Gat, U.; Dodds, H.L.

1997-01-01

Safety features and attributes of molten salt reactors (MSR) are described. The unique features of fluid fuel reactors of on-line continuous processing and the ability for so-called external cooling result in simple and safe designs with low excess reactivity, low fission product inventory, and small source term. These, in turn, make a criticality accident unlikely and reduce the severity of a loss of coolant to where they are no longer severe accidents. A melt down is not an accident for a reactor that uses molten fuel. The molten salts are stable, non-reactive and efficient heat transfer media that operate at high temperatures at low pressures and are highly compatible with selected structural materials. All these features reduce the accident plethora. Freeze valves can be used for added safety. An ultimate safe reactor (U.S.R) is described with safety features that are passive, inherent and non-tamperable (PINT)

Proposal for a technology-neutral safety approach for new reactor designs

International Nuclear Information System (INIS)

2007-09-01

Many states are considering an expansion of their nuclear power generation programmes. Many of the technologies and concepts are new and innovative. The current design and licensing rules are applicable to mostly large water reactors and there are no accepted rules in place for design, safety assessment and licensing for new innovative nuclear power plants. This TECDOC proposes a (new) safety approach and a methodology to generate technology-neutral (i.e. independent of reactor technology) safety requirements and a 'safe design' for advanced and innovative reactors. The experience gained in decades of design and licensing, combined with the development of risk-based concepts, has provided insights that will form the basis for new safety rules and requirements. Many lessons learned acknowledge the importance of such concepts as safety goals and defence in depth and the benefits of integrating risk insights early in an iterative design process. A new safety approach will incorporate many of the new developments in these concepts. For example, the probabilistic elements of defence in depth will help define the cumulative provisions to compensate for uncertainty and incompleteness of our knowledge of accident initiation and progression. This TECDOC also identifies areas of work, which will require further definition, research and development and guidance on application. This publication is to be used as a guide to developing a new technology-neutral safety approach, and as a guide in the application of methodologies to define the safety requirements for an innovative reactor designs. The method proposes an integration of deterministic and probabilistic considerations with established principles and concepts such as safety goals and defence in depth. The TECDOC recommends that the structure of the new technology-neutral main pillars for the design and licensing of innovative nuclear reactors be developed following a top-down approach to reflect a newer risk-informed and

Fast breeder reactor safety : a perspective

International Nuclear Information System (INIS)

Kale, R.D.

1992-01-01

Taking into consideration India's limited reserves of natural and vast reserves of thorium, the fast reactor route holds a great promise for India's energy supply in future. The fast reactor fueled with 239 Pu/ 238 U (unused or depleted) produces (breeds) more fissionable fuel material 239 Pu than it consumes. Calculations show that a fast breeder reactor (FBR) increases energy potential of natural uranium by about 60 times. As the fast reactor can also convert 232 Th into 233 U which is a fissionable material, it can make India's thorium reserves a source of almost inexhaustible energy supply for a long time to come. Significant advantage of FBR plants cooled by sodium and their world-wide operating experience are reviewed. There are two main safety issues of FBR, one nuclear and the other non-nuclear. The nuclear issue concerns core disruptive accident and the non-nuclear one concerns the high chemical energy potential of sodium. These two issues are analysed and it is pointed that they are manageable by current design, construction and operational practices. Main findings of safety research during the last six to eight years in West European Countries and United States of America (US) are summarised. Three stage engineered safety provision incorporated into the design of the sodium cooled Fast Breeder Test Reactor (FBTR) commissioned at Kalpakkam are explained. The important design safety features of FBTR such as primary system containment, emergency core cooling, plant protection system, inherent safety features achieved through reactivity coefficients, and natural convection cooling are discussed. Theoretical analysis and experimental research in fast reactor safety carried out at the Indira Gandhi Centre for Atomic Research during the past some years are reviewed. (M.G.B.)

Safety in decommissioning of research reactors

International Nuclear Information System (INIS)

1986-01-01

This Guide covers the technical and administrative considerations relevant to the nuclear aspects of safety in the decommissioning of reactors, as they apply to the reactor and the reactor site. While the treatment, transport and disposal of radioactive wastes arising from decommissioning are important considerations, these aspects are not specifically covered in this Guide. Likewise, other possible issues in decommissioning (e.g. land use and other environmental issues, industrial safety, financial assurance) which are not directly related to radiological safety are also not considered. Generally, decommissioning will be undertaken after planned final shutdown of the reactor. In some cases a reactor may have to be decommissioned following an unplanned or unexpected event of a series or damaging nature occurring during operation. In these cases special procedures for decommissioning may need to be developed, peculiar to the particular circumstances. This Guide could be used as a basis for the development of these procedures although specific consideration of the circumstances which create the need for them is beyond its scope

National Institute for Occupational Safety and Health

Science.gov (United States)

... Submit Search The CDC The National Institute for Occupational Safety and Health (NIOSH) Note: Javascript is disabled or is not ... and Events NIOSH Contact Information Related Federal Agencies Occupational Safety and Health Administration Mine Safety and Health Administration Follow NIOSH ...

Proceedings of the Third CSNI Workshop on Iodine Chemistry in Reactor Safety

International Nuclear Information System (INIS)

Ishigure, K.; Saeki, M.; Soda, K.; Sugimoto, J.

1992-03-01

The Third CSNI Workshop on Iodine Chemistry in Reactor Safety was held at Tokai Research Establishment of Japan Atomic Energy Research Institute at Tokai-mura, Ibaraki-ken, Japan, on September 11 to 13, 1991. About 60 experts attended the Workshop from 10 countries and 2 international organizations. In the Workshop, 29 papers were presented in five sessions on various aspects of iodine chemistry in reactor safety, such as radiolytic and surface reactions of iodine species, fundamental and integral tests, modeling and code developments. The information exchanged and the discussions followed resulted in extended and promoted understanding of iodine behavior in accidents of light water reactors and also gave a large expectation for the further progresses coming in the future. It should be emphasized that a most important and unique forum has been established through the Workshop for exchanging information and collaboratively solving the important problems in the field of the iodine chemistry in nuclear reactor safety. At the Workshop, an effort was made to integrate all information for better use by safety analysts. There is no doubt that a lot of information on iodine behaviour has been accumulated, but in many cases this information needs to be co-ordinated and well organised for safety analyses of nuclear reactors. It is essential that the results of laboratory studies and integral experiments together with modelling activities are well co-ordinated. Therefore, the goal was: - to review the knowledge and understanding of the chemistry of iodine of relevance to the prediction of its behaviour in nuclear reactors during a range of operational and accident conditions; - to define those areas of chemistry which are important but poorly understood and require further study. As shown by the conclusions of the Workshop, there is no doubt that this objective was widely attained

Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

Bokhari, I.H.; Israr, M.; Pervez, S.

1999-01-01

Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

Recommendations of the Reactor Safety Commission, 1971--1974

International Nuclear Information System (INIS)

The Reactor Safety Commission (RSC) of the German Federal Republic was created in 1958 by the then Federal Minister for Atomic Energy and Hydroeconomy. Its task included the advising of the Federal Minister responsible for protection of the population from possible dangers occurring during peaceful utilization of nuclear energy and in all matters of the safety of nuclear oriented installations. The RSC consists of independent experts from differing specialty areas. The results of their consultations are passed on in recommendations to the appropriate Federal Minister. Since the reappointment and reorganization of the RSC in the year 1971 by the then Federal Minister for Education and Science, the recommendations have been published in the ''Federal Gazette.'' The report presented is divided into two parts with a three part appendix. All recommendations of the RSC made since the reorganization (68th Meeting on 3 November 1971) until expiration of the last appointment, (96th Meeting on 25 July 1974) are compiled in the first section. The second part contains all announcements of the RSC which have appeared in the ''Federal Gazette.'' These concern training, agenda, as well as current members. Appendix A includes the ''Guidelines for Pressurized Water Reactors'' approved in the 93rd Meeting on 24 April 1974. Two key word listings follow in Sections B and C of the Appendix. The report was prepared in behalf of the Federal Minister of the Interior through the Office of the RSC assigned to the Institute for Reactor Safety of the Technical Inspection Association (IRS). The report is continued at regular intervals following each meeting period

IAEA activities on research reactor safety

International Nuclear Information System (INIS)

Alcala-Ruiz, F.

1995-01-01

Since its inception in 1957, the International Atomic Energy Agency (IAEA) has included activities in its programme to address aspects of research reactors such as safety, utilization and fuel cycle considerations. These activities were based on statutory functions and responsibilities, and on the current situation of research reactors in operation around the world; they responded to IAEA Member States' general or specific demands. At present, the IAEA activities on research reactors cover the above aspects and respond to specific and current issues, amongst which safety-related are of major concern to Member States. The present IAEA Research Reactor Safety Programme (RRSP) is a response to the current situation of about 300 research reactors in operation in 59 countries around the world. (orig.)

Status and some safety philosophies of the China advanced research reactor CARR

International Nuclear Information System (INIS)

Luzheng Yuan

2001-01-01

The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21 st century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D 2 O reflector not less than 8 x 10 14 n/cm 2 . s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10 15 n/cm 2 . s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

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

International Nuclear Information System (INIS)

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

2010-03-01

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

EPRI program in water reactor safety

International Nuclear Information System (INIS)

Loewenstein, W.B.; Gelhaus, F.; Gopalakrishnan, A.

1975-01-01

The basis for EPRI's water reactor safety program is twofold. First is compilation and development of fundamental background data necessary for quantified light-water reactor (LWR) safety assurance appraisals. Second is development of realistic and experimentally bench-marked analytical procedures. The results are expected either to confirm the safety margins in current operating parameters, and to identify overly conservative controls, or, in some cases, to provide a basis for improvements to further minimize uncertainties in expected performance. Achievement of these objectives requires the synthesis of related current and projected experimental-analytical projects toward establishment of an experimentally-based analysis for the assurance of safety for LWRs

Safety regulations concerning instrumentation and control systems for research reactors

International Nuclear Information System (INIS)

El-Shanshoury, A.I.

2009-01-01

A brief study on the safety and reliability issues related to instrumentation and control systems in nuclear reactor plants is performed. In response, technical and strategic issues are used to accomplish instrumentation and control systems safety. For technical issues there are ; systems aspects of digital I and C technology, software quality assurance, common-mode software, failure potential, safety and reliability assessment methods, and human factors and human machine interfaces. The strategic issues are the case-by-case licensing process and the adequacy of the technical infrastructure. The purpose of this work was to review the reliability of the safety systems related to these technical issues for research reactors

Fusion reactor safety studies, FY 1977

International Nuclear Information System (INIS)

Darby, J.B. Jr.

1978-04-01

This report reviews the technical progress in the fusion reactor safety studies performed during FY 1977 in the Fusion Power Program at the Argonne National Laboratory. The subjects reported on include safety considerations of the vacuum vessel and first-wall design for the ANL/EPR, the thermal responses of a tokamak reactor first wall, the vacuum wall electrical resistive requirements in relationship to magnet safety, and a major effort is reported on considerations and experiments on air detritiation

Reactor safety in Eastern Europe

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

Issues affecting advanced passive light-water reactor safety analysis

International Nuclear Information System (INIS)

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-01-01

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

The nuclear safety case for the replacement research reactor

International Nuclear Information System (INIS)

Willers, A.; Garea, V.

2003-01-01

This paper presents a broad overview of the safety case being used in the licensing of Australia's Replacement Research Reactor. The reactor is a 20 MW pool-type research reactor and is being constructed at the Lucas Heights Science and Technology Centre in Sydney's south. It will be owned and operated by the Australian Nuclear Science and Technology Organisation (ANSTO) and will take over the duties currently performed by HIFAR, a DIDO-type reactor currently operating at the site. The safety case for the RRR considers all aspects of normal operation and anticipated occurrences and will be subject to periodic review and updated in line with evolving methodologies and modifications to plant and procedures. Its scope and degree of detail ensure that the risk posed to members of the public, operators and environment are all adequately low and well in the regulatory limits

PSA in design of passive/active safety reactors

International Nuclear Information System (INIS)

Sato, T.; Tanabe, A.; Kondo, S.

1995-01-01

PSAs in the design of advanced reactors are applied mainly in level 1 PSA areas. However, even in level 1 PSA, there are certain areas where special care must be taken depending on plant design concepts. This paper identifies these areas both for passive and active safety reactor concepts. For example, 'long-term PSA' and shutdown PSA are very important for a passive safety reactor concept from the standpoint of effectiveness of a grace period and passive safety systems. External events are also important for an active safety reactor concept. These kinds of special PSAs are difficult to conduct precisely in a conceptual design stage. This paper shows methods of conducting these kinds of special PSAs simply and conveniently and the use of acquired insights for the design of advanced reactors. This paper also clarifies the meaning or definition of a grace period from the standpoint of PSA

Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

International Nuclear Information System (INIS)

Kobayashi, Michiyuki; Aya, Izuo; Inasaka, Fujio; Murata, Hiroyuki; Odano, Naoteru; Shiozaki, Koki

1998-01-01

A research project from 1995-1999 had a plan to make experimental studies on (1) safety of nuclear ship loaded with an integral ship propulsion reactor (2) effects of pulsating flow on the thermo-hydraulic characteristics of ship propulsion reactor and (3) thermo-hydraulic behaviors of the reactor container at the time of accident in a passively safe ship propulsion reactor. Development of a data base for ship propulsion reactor was attempted using previous experimental data on the thermo-hydraulic characteristics of the reactor in the institute in addition to the present results aiming to make general analytical evaluation for the safety of the engineering-simulation system for nuclear ship. A general data base was obtained by integrating the data list and the analytical program for static characteristics. A test equipment which allows to visualize the pulsating flow was produced and visualization experiments have started. (M.N.)

The safety of pressurized water reactors

International Nuclear Information System (INIS)

Panossian, J.; Tanguy, P.

1991-01-01

In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

17. meeting of the Society for Reactor Safety. Proceedings

International Nuclear Information System (INIS)

1994-06-01

An autonomous and independent reactor safety research in Germany is indispensable. Three out of the four papers of the meeting deal with the protective aim concept of NPP. Deterministic safety assessment during periodic in-service inspections, a new generation of information engineering, and the incorporation of serious accidents in the containment design of new reactors are considered in detail. (DG) [de

Nuclear reactor safety system

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1983-01-01

The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

Energy Technology Data Exchange (ETDEWEB)

David Petti; Jim Kinsey; Dave Alberstein

2014-01-01

Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

Reactor safety research and safety technology. Pt. 2

International Nuclear Information System (INIS)

Theenhaus, R.; Wolters, J.

1987-01-01

The state of HTR safety research work reached permits a comprehensive and reliable answer to be given to questions which have been raised by the reactor accident at Chernobyl, regarding HTR safety. Together with the probability safety analyses, the way to a safety concept suitable for an HTR is cleared; instructions are given for design optimisation with regard to safety technique and economy. The consequences of a graphite fire, the neutron physics design and the consequenes of the lack of a safety containment are briefly described. (DG) [de

Reactor system safety assurance

International Nuclear Information System (INIS)

Mattson, R.J.

1984-01-01

The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

Institute for Safety Research. Annual report 1992

International Nuclear Information System (INIS)

Weiss, F.P.; Boehmert, J.

1993-11-01

The Institute is concerned with evaluating the design based safety and increasing the operational safety of technical systems which include serious sources of danger. It is further occupied with methods of mitigating the effects of incidents and accidents. For all these goals the institute does research work in the following fields: modelling and simulation of thermofluid dynamics and neutron kinetics in cases of accidents; two-phase measuring techniques; safety-related analyses and characterizing of mechanical behaviours of material; measurements and calculations of radiation fields; process and plant diagnostics; development and application of methods of decision analysis. This annual report gives a survey of projects and scientific contributions (e.g. Single rod burst tests with ZrNb1 cladding), lists publications, institute seminars and workshops, names the personal staff and describes the organizational structure. (orig./HP)

Paul Scherrer Institut annual report 1995. Annex IV: PSI nuclear energy and safety

International Nuclear Information System (INIS)

Birchley, J.; Roesel, R.; Doesburg, R. van

1996-01-01

Nuclear energy research in Switzerland is concentrated at PSI's Department F4. It is explicitly mentioned in the Institute's official charter and commands about one fifth of the Institute's federal resources. Presently, PSI invests approx. 200 py/a in nuclear energy research, one third of this being externally funded; the share of external funding in investment costs totals approx. 50%. This funding is provided by the Swiss utilities and the NAGRA, the Safety Authority (HSK) and the former National Fund for Energy Research (NEFF). PSI's activities in nuclear research concentrate on three main areas: safety of operating plants, safety features of future reactor concepts and waste management. 7% of personnel are invested in addressing global aspects of energy. (author) figs., tabs., refs

Safety re-assessment of AECL test and research reactors

International Nuclear Information System (INIS)

Winfield, D.J.

1990-01-01

Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

List of reports from reactor safety research of BMFT, USNRC, EPRI and JSTA

International Nuclear Information System (INIS)

1979-01-01

This list reviews reports from the Federal Republic of Germany, from the USA and from Japan concerning special problems in the field of Reactor Safety Research. According to the cooperation of the Bundesminister fuer Forschung und Technologie (BMFT) with the USNRC, the Electric Power Research Institute (EPRI), and the Japan Science and Technology Agency (JSTA) these reports are available in the Gesellschaft fuer Reaktorsicherheit. The list pursues the following order: Country of origin, problem area concerned, according to the Reactor Safety Research Program of BMFT, reporting organisation. The list of reports appears quarterly. Requests for reports should be adressed to GRS, Forschungsbetreuung. Contractual view points have to be considered for the distribution of the reports. (orig./HP) [de

List of reports from reactor safety research of BMFT, USNRC, EPRI and JSTA

International Nuclear Information System (INIS)

1978-12-01

This list reviews reports from the Federal Republic of Germany, from the USA and from Japan concerning special problems in the field of Reactor Safety Research. According to the cooperation of the Bundesminister fuer Forschung und Technologie (BMFT) with the USNRC, the Electric Power Research Institute (EPRI), and the Japan Science and Technology Agency (JSTA) these reports are available in the Gesellschaft fuer Reaktorsicherheit. The list pursues the following order: Country of origin, problem area concerned, according to the Reactor Safety Research Program of BMFT, reporting organisation. The list of reports appears quarterly. Requests for reports should be addressed to GRS, Forschungsbetreuung. Contractual view points have to be considered for the distribution of the reports. (orig./HP) [de

Solution of safety problems for nuclear power plants with WWER-440 reactors

International Nuclear Information System (INIS)

Krett, V.; Pernitsa, R.; Pfann, Ya.; Zbeglik, J.

1982-01-01

Institute of nuclear research (INR) of Czechoslovakian Atomic Energy Commission isto fulfil the supervision functions within the field of nuclear power research and development. The problems of safe operation ensurance for the nuclear power plants (NPP) with WWER-440 reactors are studied within the frame of sever major issues: code standardization and devolopment of guiding materials for the state supervision; neutronic and thermohydrolic data processing for the accident analysis; operation reliability studies of the safety systems and estimates of separate component failure importance; assessment of the accidents resulting from the equipment misfunctioning and component failures; development of a controlled reliability program; evaluation of the atomic installationsimpact on the environment; ensurence of the reactor vessel reliability and durability under irradiation. The NPP safety analysis incorporates the calculations of transient and accidental regimes for the core, the primary loop and the entire plant. A number of codes has been produced which allow to determine the state of fuel elements during operation just before the accidents assessed, thermohydrolic conditions in the coolant and the temperature distribution within the fuel both for the stationary reactor conditions and for transient regimes. A mathematical model has been deveoloped, including the description of all the primary loop major components. The Soviet code DYNAMIKA has been adopted and adjusted for EC-1040 computer, there by the accident analysis for the entire NPP has been made possible. On the basis of american SAFTE code a faster SAFEDO-2 code has been developed employing the Monte Carlo method for the accident analysis of a complex system described by means of a failure tree. The discussed codes are used at the data assessment for the accident analysis part of the safety reports as well as for the reliability evaluation of the emergency core cooling system [ru

High temperature reactor safety and environment

International Nuclear Information System (INIS)

Brisbois, J.; Charles, J.

1975-01-01

High-temperature reactors are endowed with favorable safety and environmental factors resulting from inherent design, main-component safety margins, and conventional safety systems. The combination of such characteristics, along with high yields, prove in addition, that such reactors are plagued with few problems, can be installed near users, and broaden the recourse to specific power, therefore fitting well within a natural environment [fr

Review on conformance of JMTR reactor facility to safety design examination guides for water-cooled reactors for test and research

International Nuclear Information System (INIS)

Ide, Hiroshi; Naka, Michihiro; Sakuta, Yoshiyuki; Hori, Naohiko; Matsui, Yoshinori; Miyazawa, Masataka

2009-03-01

The safety design examination guides for water-cooled reactors for test and research are formulated as fundamental judgements on the basic design validity for licensing from a viewpoint of the safety. Taking the refurbishment opportunity of the JMTR, the conformance of the JMTR reactor facility to current safety design examination guides was reviewed with licensing documents, annexes and related documents. As a result, it was found that licensing documents fully satisfied the requirements of the current guides. Moreover, it was found that the JMTR reactor facility itself also satisfied the guides requirements as well as the safety performance, since the facility with safety function such as structure, systems, devices had been installed based on the licensing documents under the permission by the regulation authority. Important devices for safety have been produced under authorization of regulating authority. Therefore, it was confirmed that the licensing was conformed to guides, and that the JMTR has enough performance. (author)

Safety requirements in the design of research reactors: A Canadian perspective

International Nuclear Information System (INIS)

Lee, A.G.; Langman, V.J.

2000-01-01

In Canada, the formal development of safety requirements for the design of research reactors in general began under an inter-organizational Small Reactor Criteria Committee. This committee developed safety and licensing criteria for use by several small reactor projects in their licensing discussions with the Atomic Energy Control Board. The small reactor projects or facilities represented included the MAPLE-X10 reactor, the proposed SES-10 heating reactor and its prototype, the SDR reactor at the Whiteshell Laboratories, the Korea Multipurpose Research Reactor (a.k.a., HANARO) in Korea, the SCORE project, and the McMaster University Nuclear Reactor. The top level set of criteria which form a safety philosophy and serve as a framework for more detailed developments was presented at an IAEA Conference in 1989. AECL continued this work to develop safety principles and design criteria for new small reactors. The first major application of this work has been to the design, safety analysis and licensing of the MAPLE 1 and 2 reactors for the MDS Nordion Medical Isotope Reactor Project. This paper provides an overview of the safety principles and design criteria. Examples of an implementation of these safety principles and design criteria are drawn from the work to design the MAPLE 1 and 2 reactors. (author)

Examination of the bases for proposed innovations in reactor safety technology

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

This paper employs the criteria for evaluations from the Nuclear Power Option Viability Study to examine the bases for proposed innovations in light water reactor safety technology. These bases for innovation fall into four broad categories as follows: (1) virtually exclusive reliance on passive safety features to preclude core damage in all situations, (2) design simplification using some passive safety features to reduce the frequency of core damage to less than about 10 -6 per reactor-year, (3) passive containment to preclude releases from any accident, and (4) designing to limit licensing attention to one or at least a few systems. Of these, only the first two, and perhaps only the second, hold significant promise for providing for the viability of advanced light water reactors

Development of small reactor safety criteria in Canada

International Nuclear Information System (INIS)

Ernst, P.C.; French, P.M.; Axford, D.J.; Snell, V.G.

1990-01-01

A number of new small reactor designs have been proposed in Canada over the last several years and some have reached the stage where licensing discussions have been initiated with the Atomic Energy Control Board (AECB). An inter-organizational Small Reactor Criteria (SRC) working group was formed in 1988 to propose safety and licensing criteria for these small reactors. Two levels of criteria are proposed. The first level forms a safety philosophy and the second is a set of criteria for specific reactor applications. The safety philosophy consists of three basic safety objectives together with evaluation criteria, and fourteen fundamental principles measured by specific criteria, which must be implemented to meet the safety objectives. Two of the fourteen principles are prime: defence in depth, and safety culture; the other twelve principles can be seen as deriving from them. A benefit of this approach is that the concepts of defence in depth and safety culture become well-defined. The objectives and principles are presented in the paper and their criteria are summarized. The second level of criteria, under development, will form a safety application set and will provide small reactor criteria in a number of general areas, such as regulatory process and safety assessment, as well as for specific reactor life-cycle activities, from siting through to decommissioning. The criteria are largely deterministic. However, the frequencies and consequences of postulated accidents are assessed against numerical criteria to assist in judging the acceptability of plant design, operation, and proposed siting. All criteria proposed are designed to be testable in some evidentiary fashion, readily enabling an assessment of compliance for a given proposal

An Overview of the Safety Case for Small Modular Reactors

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2011-01-01

Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.

Power reactor core safety research

International Nuclear Information System (INIS)

Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

1981-01-01

As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

Criticality safety studies involved in actions to improve conditions for storing 'RA' research reactor spent fuel

International Nuclear Information System (INIS)

Matausek, M.; Marinkovic, N.

1998-01-01

A project has recently been initiated by the VINCA Institute of Nuclear Sciences to improve conditions in the spent fuel storage pool at the 6.5 MW research reactor RA, as well as to consider transferring this spent fuel into a new dry storage facility built for the purpose. Since quantity and contents of fissile material in the spent fuel storage at the RA reactor are such that possibility of criticality accident can not be a priori excluded, according to standards and regulations for handling fissile material outside a reactor, before any action is undertaken subcriticality should be proven under normal, as well as under credible abnormal conditions. To perform this task, comprehensive nuclear criticality safety studies had to be performed. (author)

List of BMFT, USNRC, EPRI, and STA reports on reactor safety research

International Nuclear Information System (INIS)

1980-07-01

This list reviews reports from the Federal Republic of Germany, from the United States of America and from Japan concerning special problems in the field of Reactor Safety Research. According to the cooperation of the Bundesminister fuer Forschung und Technologie (BMFT) with the United States Nuclear Regulatory Commission (USNRC), the Electric Power Research Institute (EPRI), and the Japan Science and Technology Agency (JSTA) these reports are available in the Gesellschaft fuer Reaktorsicherheit. The list pursues the following order : Country of origin, problem area concerned, according to the Reactor Safety Research Program of BMFT, reporting organisation. The list of reports appears quarterly. (orig./HP) [de

Safety Analysis for Power Reactor Protection System

International Nuclear Information System (INIS)

Eisawy, E.A.; Sallam, H.

2012-01-01

The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

Gas-cooled breeder reactor safety

Energy Technology Data Exchange (ETDEWEB)

Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

1981-01-15

The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

Current state of the auto-evaluation process of the behaviour code in the safety of research reactors in Mexico

International Nuclear Information System (INIS)

Mamani A, Y. R.; Salgado G, J. R.

2011-11-01

In Mexico, the regulator organism in nuclear matter is the National Commission of Nuclear Safety and Safeguards, and a nuclear research reactor exists, the TRIGA Mark III, operated by the National Institute of Nuclear Research. In this work the main aspects of the current state and the future challenges are presented with relationship to the installation of the auto-evaluation process of the behaviour code in the safety of research reactors for the TRIGA reactor case. Additionally, the legal mark of the licensing process for the nuclear activities in a research reactor is described in a brief way, and the main characteristics of the reactor, the uses for the isotopes production, the administration and the verification of the safety, the administration program of the radiological protection, the emergency plan and the operation personnel qualification are emphasized. (Author)

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

Savannah River Site K-Reactor Probabilistic Safety Assessment

International Nuclear Information System (INIS)

Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O'Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

1992-12-01

This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety

Tests for validation of fast neutron reactors safety

International Nuclear Information System (INIS)

Nagata, T.; Yamashita, H.

2001-01-01

Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

Refurbishment and safety up-gradation of Cirus Reactor

International Nuclear Information System (INIS)

Rao, D.V.H.

2004-01-01

Cirus, a 40 MWth, vertical tank type research reactor, having a wide range of research facilities, was commissioned in 1960. This research facility has been operated and utilized extensively for nearly four decades. With a view to assess the residual life of the reactor, detailed ageing studies were carried out. Based on this, refurbishment work for life extension was undertaken. During this work, additional safety features were incorporated to improve the overall safety of the reactor. This lecture details the methodologies used for ageing studies and refurbishment activities for life extension with enhanced safety. (author)

Defence-in-depth and development of safety requirements for advanced nuclear reactors

International Nuclear Information System (INIS)

Carnino, A.; Gasparini, M.

2002-01-01

The paper addresses a general approach for the preparation of the design safety requirements using the IAEA Safety Objectives and the strategy of defence-in-depth. It proposes a general method (top-down approach) to prepare safety requirements for a given kind of reactor using the IAEA requirements for nuclear power plants as a starting point through a critical interpretation and application of the strategy of defence-in-depth. The IAEA has recently developed a general methodology for screening the defence-in-depth of nuclear power plants starting from the fundamental safety objectives as proposed in the IAEA Safety Fundamentals. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor. Currently the IAEA is preparing the technical basis for the development of safety requirements for Modular High Temperature Gas Reactors, with the aim of showing the viability of the method. A draft TECDOC has been prepared and circulated among several experts for comments. This paper is largely based on the content of the draft TECDOC. (authors)

Safety Analysis Of Actinide Recycled Fast Power Reactor

International Nuclear Information System (INIS)

Taufik, Mohammad

2001-01-01

Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State

List of reports in reactor safety research by BMFT, USNRC, EPRI and JSTA

International Nuclear Information System (INIS)

1981-01-01

This list reviews reports from the Federal Republic of Germany, from the United States of America and from Japan concerning special problems in the field of Reactor Safety research. According to the cooperation of the Bundesminister fuer Forschung und Technologie (BMFT) with the United States Nuclear Regulatory Commission (USNRC), the Electric Power Research Institute (EPRI), and the Japan Science and Technology Agency (JSTA) these reports are available in the Gesellschaft fuer Reaktorsicherheit. The list pursues the following order: Country of origin, problem area concerned, according to the Reactor Safety Research Program of BMFT, reporting organisation. The list of reports appears quarterly. (orig.) [de

Safety approach to the selection of design criteria for the CRBRP reactor refueling system

International Nuclear Information System (INIS)

Meisl, C.J.; Berg, G.E.; Sharkey, N.F.

1979-01-01

The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents. The process steps are illustrated by examples

Passive Safety Features for Small Modular Reactors

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2010-01-01

The rapid growth in the size and complexity of commercial nuclear power plants in the 1970s spawned an interest in smaller, simpler designs that are inherently or intrinsically safe through the use of passive design features. Several designs were developed, but none were ever built, although some of their passive safety features were incorporated into large commercial plant designs that are being planned or built today. In recent years, several reactor vendors are actively redeveloping small modular reactor (SMR) designs with even greater use of passive features. Several designs incorporate the ultimate in passive safety they completely eliminate specific accident initiators from the design. Other design features help to reduce the likelihood of an accident or help to mitigate the accidents consequences, should one occur. While some passive safety features are common to most SMR designs, irrespective of the coolant technology, other features are specific to water, gas, or liquid-metal cooled SMR designs. The extensive use of passive safety features in SMRs promise to make these plants highly robust, protecting both the general public and the owner/investor. Once demonstrated, these plants should allow nuclear power to be used confidently for a broader range of customers and applications than will be possible with large plants alone.

Safety aspects on dependability management for a TRIGA research reactor in Romania

International Nuclear Information System (INIS)

Vieru, Gheorghe

2002-01-01

Safety on the management for a nuclear research reactor involves a 'good dependability management' of the activities, such as: reliability, availability, maintainability and maintenance support. In order to evaluate the safety management aspects intended to be applied at a research reactor management, the performance dependability indicators and their impact over the availability and reactor safety have to be established. The document ISO 9000-4/IEC 300-1 'Dependability Management' (1995), describes five internationally agreed indicators of the reactor equipment dependability, each of them can be used for corrective maintenance or for preventive maintenance, such as: I 1 - equipment Maintenance Frequency; I 2 - equipment Maintenance Effort; I-3 - equipment Maintenance Downtime Factor; I 4 - equipment Maintenance Contribution to the System Function Downtime Factor; I 5 - equipment Maintenance Contribution to the reactor Capability Loss Factor. The paper presents an evaluation of those 5 mentioned indicators with referring only at the primary circuit of the INR's TRIGA research reactor and conclusion. The analyzed period was stated between 1994-1999. It is to be noted that this type of analyze is performed for the first time for a research reactor. (author)

Development of a numerical tool for safety assessment and emergency management of experimental reactors

International Nuclear Information System (INIS)

Maas, L.; Beuter, A.; Seropian, C.

2010-01-01

The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. Among its duties, one important item is to provide help for emergency situations management in case of an accident occurring in a French nuclear facility. In this framework, IRSN develops and applies numerical tools dealing with containment management issues. Up to now IRSN has not got any specific tool for experimental reactors. Accordingly, it has been then decided to extend the ASTEC code, devoted to severe accident scenarios for Pressurized Water Reactors, to this kind of reactors. This lumped-parameter code, co-developed by IRSN and GRS (Germany), covers the entire phenomenology from the initiating event up to fission products release outside the reactor containment, except for the steam explosion and the mechanical integrity of the containment. A first application to experimental reactors was carried out to assess the High Flux Reactor (HFR) operator's improvement proposal concerning the containment management during accidental situations. This reactor, located in Grenoble (France), is composed of a double wall containment with a pressurized containment annulus preventing any direct leakage into the environment. Until now, in case of severe accidents (mainly core melting in pool, explosive reactivity accident called BORAX), the HFR emergency management consisted in isolating the containment building in the early stage of the accident, to prevent any radioactive products release to the environment. The operator decided to improve this containment management during accidental situations by using an air filtering venting system able to maintain a slight sub-atmospheric pressure in the reactor building. The operator's demonstration of the efficiency of this new system is mainly based on containment pressure evaluations during accidental transients. IRSN assessed these calculations through ASTEC calculations. Finally, a global agreement was

Applicability of RELAP5 for safety analysis of AP600 and PIUS reactors

International Nuclear Information System (INIS)

Motloch, C.G.; Modro, S.M.

1990-01-01

An assessment of the applicability of using RELAP5 for performing safety analyses of the AP600 and PIUS advanced reactor concepts is being performed. This ongoing work is part of a larger safety assessment of advanced reactors sponsored by the United States Nuclear Regulatory Commission. RELAP5 models and correlations are being reviewed from the perspective of the new AP600 and PIUS phenomena and features that could be important to reactor safety. The purpose is to identify those areas in which new mathematical models of physical phenomena would be required to be added to RELAP5. In most cases, the AP600 and PIUS designs and systems and the planned and off-normal operations are similar enough to current Pressurized Water Reactors (PWR) that RELAP5 safety analysis applicability is unchanged. However, for AP600 the single most important systemic and phenomenological difference between it and current PWRs is in the close coupling between the reactor system and the containment during postulated Loss of Coolant Accident (LOCA) events. This close coupling may require the addition of some thermal-hydraulic models to RELAP5. And for PIUS, the most important new feature is the thermal density locks. These and other important safety-related features are discussed. This document presents general descriptions of RELAP5, AP600, and PIUS, describes the new features and phenomena of the reactors, and discusses the code/reactors safety-related issues. 32 refs., 4 figs., 2 tabs

Reactor safety in Eastern Europe. Proceedings

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP) [de

Safety management at nuclear installations with research reactors. A comparison of five European installations

International Nuclear Information System (INIS)

Troen, H.; Lauridsen, B.

1997-11-01

Five European institutions with nuclear research reactors were visited to compare safety management among institutions similar to Risoe. Risoe is a National Laboratory and the main activities are research and development. In 1996 it was decided to look into safety management at Risoe again; the last revision was in 1972. The purpose was to make it more efficient and to emphasise, that the responsibility lies in the operating organisation. Information such as nuclear facilities at the institutions, the safety management organisation, emergency preparedness, and lists of radiation doses to the employees from the years 1995 and 1996 is given in the report. Also international requirements and recommendations are given in short. Furthermore the report contains some reflections on the development in safety management organisations in resent years and the conclusions drawn from the information gathered

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 1997

Energy Technology Data Exchange (ETDEWEB)

Weiss, F P; Rindelhardt, U [eds.

1998-10-01

The research work of the institute aims at the assessment and increase of the safety of technical facilities. Subject of the investigation are equally nuclear plants and installations of process industries. To analyse thermo-fluiddynamic phenomena of accident scenarios physical models and computer codes are developed as well for multi-phase and multi-component flows as for the time and space dependent power release (neutron kinetics in light water reactors, reaction kinetics of exothermic chemical reactions). Emphasis is put on the description of spatial flows and the transient evolution of flow patterns. (orig.)

Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

International Nuclear Information System (INIS)

Mohamed, F.; Hassan, A.; Yahaya, R.; Rahman, I.; Maskin, M.; Praktom, P.; Charlie, F.

2015-01-01

Highlights: • Human Reliability Analysis (HRA) for Level 1 Probabilistic Safety Analysis (PSA) is performed on research nuclear reactor. • Implemented qualitative HRA framework is addressed. • Human Failure Events of significant impact to the reactor safety are derived. - Abstract: A Level 1 Probabilistic Safety Analysis (PSA) for the TRIGA Mark II research reactor of Malaysian Nuclear Agency has been developed to evaluate the potential risk in its operation. In conjunction to this PSA development, Human Reliability Analysis (HRA) is performed in order to determine human contribution to the risk. The aim of this study is to qualitatively analyze human actions (HAs) involved in the operation of this reactor according to the qualitative part of the HRA framework for PSA which is namely the identification, qualitative screening and modeling of HAs. By performing this framework, Human Failure Events (HFEs) of significant impact to the reactor safety are systematically analyzed and incorporated into the PSA structure. A part of the findings in this study will become the input for the subsequent quantitative part of the HRA framework, i.e. the Human Error Probability (HEP) quantification

Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; Okumura, Keisuke; Suzuki, Motoe; Mineo, Hideaki; Nakatsuka, Toru

2004-06-01

The present report contains the achievement of 'Research and Development on Reduced-moderation Light Water Reactor with Passive Safety Features', which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies. Our development target is 'Reduced-moderation Light Water Reactor with Passive Safety Features' with innovative technologies to achieve above mentioned requirement. Electric power is selected as 300 MWe considering anticipated size required for future deployment. The reactor core consists of MOX fuel assemblies with tight lattice arrangement to increase the conversion ratio. Design targets of the core specification are conversion ratio more than unity, negative void reactivity feedback coefficient to assure safety, discharged burnup more than 60 GWd/t and operation cycle more than 2 years. As for the reactor system, a small size natural circulation BWR with passive safety systems is adopted to increase safety and reduce construction cost. The results obtained are as follows: As regards core design study, core design was performed to meet the goal. Sequence of startup operation was constructed for the RMWR. As the plant design, plant system was designed to achieve enhanced economy using passive safety system effectively. In

Paul Scherrer Institut annual report 1994. Annex IV: PSI nuclear energy and safety research progress report 1994

Energy Technology Data Exchange (ETDEWEB)

Williams, T; Kallfelz, J M; Mathews, D [eds.; Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1995-10-01

Nuclear energy research in Switzerland is concentrated at PSI. It is explicitly mentioned in the Institute`s official charter and commands about one fifth of the Institute`s federal resources. Presently, PSI invests approx. 200 py/a in nuclear energy research, one third of this being externally funded; the share of external funding in investment costs totals approx. 50%. This funding is provided by the Swiss utilities and the NAGRA, the Safety Authority (HSK) and the former National Fund for Energy Research (NEFF). PSI`s activities in nuclear research concentrate on three main areas: safety of operating plants, safety features of future reactor concepts and waste management. 7% of personnel are invested in addressing global aspects of energy. (author) figs., tabs., refs.

Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor

International Nuclear Information System (INIS)

Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

1994-02-01

This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC's ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC's preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant's research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified

RB research reactor safety report; Izvestaj o sigurnsti istrazivackog reaktora RB

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Pesic, M; Vranic, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1979-04-15

This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document.

Periodic safety review of the experimental fast reactor JOYO. Review of the activity for safety

International Nuclear Information System (INIS)

Maeda, Yukimoto; Kashimura, Youichi; Suzuki, Toshiaki; Isozaki, Kazunori; Hoshiba, Hideaki; Kitamura, Ryoichi; Nakano, Tomoyuki; Takamatsu, Misao; Sekine, Takashi

2005-02-01

Periodic safety review (Review of the activity for safety) which consisted of 'Comprehensive evaluation of operation experience' and Incorporation of the latest technical knowledge' was carried out up to January 2005. 1. Comprehensive evaluation of operation experience. It was confirmed that the effectual activities for safety through the operation of JOYO were carried out in terms of (1) Operation management, (2) Maintenance management, (3) Fuel management, (4) Radiation management, (5) Radioactive waste management, (6) Emergency planning and (7) Feedback of incidents and failures. 2. Reflection of the latest technical knowledge. It was confirmed that the latest technical knowledge including regulation and guide line established by Nuclear Safety Commission of Japan until March 31st. 2003 were properly reflected in impressing the safety of the reactor. As a result, it was evaluated that the activity for safety was carried out effectually, and no additional measure was identified continual safe operation of the reactor. (author)

Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

International Nuclear Information System (INIS)

2002-01-01

A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

Safety Committees for Argentinean Research Reactor - Regulatory Issues

International Nuclear Information System (INIS)

Perrin, Carlos D.

2009-01-01

In the field of radiological and nuclear safety, the Nuclear Regulatory Authority (ARN) of Argentina controls three research reactors and three critical assemblies, by means of evaluations, audits and inspections, in order to ensure the fulfillment of the requirements established in the Licenses, in the Regulatory Standards and in the Mandatory Documentation in general. From the Nuclear Regulatory Authority's point of view, within the general process of research reactors safety management, the Operational Organization self verification of radiological and nuclear safety plays an outstanding role. In this aspect the ARN has established specific requirements in the Regulatory Standards, in the Operation Licenses and in the Operational Limits and Conditions. These requirements include the figure of different safety committees, which act as reviewers or advisers in diverse situations. This paper describes the main characteristics of the committees, their function, scope and the regulatory documents where the requirements are included. (author)

Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

2014-08-15

The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring