Technical Working Group on Fast Reactors: German Report

International Nuclear Information System (INIS)

Maschek, W.

2012-01-01

General German situation: • After Fukushima decision to phase out definitely until 2022; • Currently only 9 reactors left (2 BWRs and 7 PWRs); • Start of new search for final repository all over Germany (salt, clay); • Discussion on retrievable repository; • Search for methods to reduce risk from waste repositories. Future prospects: • Continuation of work within EU programs; • Continuation of work within international organizations (IAEA, NEA-OECD); • New CRP‘s are of interest; • Education and training; • Focus on safety: prevention and mitigation; • Embedded in work on transmutation and waste treatment; • Cycle studies to demonstrate both sustainability potential and waste burning capability (reduced loads for final repositories) of FR

The Nuclear Safety Convention - does it confirm existing German law, and update international law?

International Nuclear Information System (INIS)

Lindemann, C.

1995-01-01

Some selected examples are discussed that are intended to answer the question of whether the NSC in its essence represents a development in confirmation of existing German nuclear law, and whether, assuming its coming into effect, this Convention will mean a step forward in the development of international law. The author examines the value of this codification of international law as such, and some of the obligations and standards such as retrofitting measures or shutdown of reactors below safety standard, and continues with briefly discussing the relationship between the NSC and nuclear liability law, the planned provisions for radiological protection in Art. 15, and the obligations for transboundary notification of safety-relevant events. These stipulations are analysed in comparison to existing international law, and with a view to their implementation under German law. Some provisions of the NSC that are based on standards of international technical guidance are compared with German regulatory guides. (orig./HP) [de

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

The materials concept in German light water reactors. A contribution to plant safety, economic performance and damage prevention

International Nuclear Information System (INIS)

Ilg, Ulf

2008-01-01

Major decisions taken as early as in the planning and construction phases of nuclear power plants may influence overall plant life. Component quality at the beginning of plant life is determined very much also by a balanced inclusion of the 'design, choice of materials, manufacturing and inspection' elements. One example of the holistic treatment of design, choice of material, and manufacture of important safety-related components in pressurized water reactors is the reactor pressure vessel (RPV) in which the ferritic compound tubes, with inside claddings, for the control rod drive nozzles are screwed into the vessel top. Also the choice of Incoloy 800 for the steam generator tubes, and the design of the main coolant pipes with inside claddings as seamless pipe bends / straight pipes with integrated nozzles connected to mixed welds with austenitic pipes are other special design features of the Siemens/KWU plants. A demonstrably high quality standard by international comparison to this day has been exhibited by the austenitic RPV internals of boiling water reactors, which were made of a low-carbon Nb-stabilized austenitic steel grade by optimum manufacturing technologies. The same material is used for backfitting austenitic pipes. Reliable and safe operation of German nuclear power plants has been demonstrated for more than 4 decades. One major element in this performance is the materials concept adopted in Germany also in the interest of damage prevention. (orig.)

Plant specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events

International Nuclear Information System (INIS)

2011-01-01

The German Parliament requested (17 March 2011) a comprehensive inspection of German nuclear power plants. For this purpose independent expert commissions should perform a new risk analysis of all German NPPS and nuclear installations with respect to the lessons learned from the Fukushima (Japan) events and other extraordinary damage scenarios. The Reactor safety commission (RSK) was assigned by the German Bundesamt fuer Strahlenschutz to develop a catalogue of requirements for this safety inspection. The contribution summarizes the required inspection volume (status 30.03.2011) including the following events: natural events like earth quakes, floods, weather-based consequences and possible superposition. Additionally the following assumptions have to be considered: event independent postulated common failures or systematic faults, station blackout larger than 2 hours, long-term failure of the auxiliary cooling water supply; aggravating boundary conditions for the performance of emergency measures (non-availability of power supply), hydrogen generation and detonation hazard, restricted personnel availability, non-accessibility due to high radiation levels, impeded technical support from outside. (orig.)

The European pressurized water reactor. The French-German advanced PWR project

International Nuclear Information System (INIS)

Watteau, M.P.; Seidelberger, H.; Broecker, B.; Serviere, G.

1995-01-01

In order to derive full benefit from the Franco-German experience and to maintain a continuous development process, the EPR is of an evolutionary design. It is also an innovative product, intended to combine competitiveness, improved operability and enhanced safety. With a large electrical output, in the range of 1400-1500 MW, the EPR has an excellent cost-size ratio and is fully adapted to the scarcity of sites. It is also suitable for the development of scale-down products based on the same technology. The progress in safety that can be achieved with the EPR can be used to further enhance the already very high level of safety of the current French and German nuclear plants: (1) by an improvement of the preventive level of the defence-in-depth concept, and (2) by the implementation of additional features, mainly for the containment, to mitigate the consequences of severe accidents. Economically, the generation cost objective of the EPR will be at least as good as that of the French N4 reactor series. Improvements in safety may imply some additional investment costs, compared with those of the current plants, but the EPR is designed to achieve reductions of the other components of the generation costs, i.e. the fuel cycle costs and the operation and maintenance costs. The paper also describes some major design features of of the EPR: the safety systems, the containment and confinement functions, the arrangement of buildings, protection against external attacks, the man-machine interface, and instrumentation and control. 5 figs, 2 tabs

High-Speed Maglev Trains; German Safety Requirements

Science.gov (United States)

1991-12-31

This document is a translation of technology-specific safety requirements developed : for the German Transrapid Maglev technology. These requirements were developed by a : working group composed of representatives of German Federal Railways (DB), Tes...

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

The risks of nuclear energy technology. Safety concepts of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Inst. fuer Kern- und Energietechnk (IKET); Kessler, Guenter; Veser, Anke; Schlueter, Franz-Hermann

2014-11-01

Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

Impacts of the Chernobyl reactor accident on the territories of the former German Democratic Republic in 1989

International Nuclear Information System (INIS)

1992-08-01

Several reports by SAAS (the Nuclear Safety and Radiation Protection Board of the German Democratic Republic) have been discussing the effects of the Chernobyl reactor accident through 1989. Only a summary had been published for 1989 in the environmental radioactivity annual report. Institut fuer Umweltschutz had been in charge of the publication of a more detailed account as part of the 'environmental report' but the project was abandoned since the institute was wound up as of October 1990. The report under review concludes the separate German Demoncratic Republic reporting by publishing the part of the manuscript on environmental contamination caused by artificial radionuclides which gives the 1989 situation on the basis of the previous results on the effects of the Chernobyl reactor accident. The appendix lists the SAAS reports published in the past. (orig./BBR) [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

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.)

Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

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

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

Review of the German Reactor Safety Study

International Nuclear Information System (INIS)

Bayer, A.

1980-01-01

The accident risks in German nuclear power plants are reviewed. Parameters influencing the extent and consequences of an accident are differentiated. Risks of normal operation, war, and sabotage have not been considered. Measures taken by the operating shaft in the course of an accident are taken into account. The knowledge thus gained can be applied in the planning of emergency measures. (DG) [de

Safety concept of high-temperature reactors based on the experience with AVR and THTR

International Nuclear Information System (INIS)

Wachholz, Winfried; Kroeger, Wolfgang

1990-01-01

In the Federal Republic of Germany a reactor is considered safe if verification has been furnished that the requirements contained in paragraph 7 of the Federal German Atomic Energy Act are met for this reactor: demonstration of sufficient precautions against damage required according to the actual state of the art, and especially compliance with the dose rate limits for normal operation and accidental conditions. These requirements result in a deterministic multi-stage safety concept with specified requirements for the engineered safety systems. In recent years, proposals for enhanced safety of nuclear power reactors or a radical change in safety philosophy have been made. This is characterised by 'inherently safe', 'super safe' and similar slogans. A quantitative definition of these requirements has not yet been established, but it is clear as a common objective that the event of beyond design basis accidents evacuation, relocation, and large scale contamination of ground should not occur. As a consequence of the Chernobyl accident the safety of all the NPPs in Germany has been reviewed. This analysis was completed for the THTR reactor in 1988. The same has been done for AVR reactor. The final evaluation of the HTR specific safety features have been fully confirmed. The HTR concepts under development are based on this experience. The HTR-Modul unit is currently being designed

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

The development of the breeder reactor program of the German Federal Republic

International Nuclear Information System (INIS)

Haunschild, H.H.

1984-01-01

This article recapitulates the stages of the program which is in hand in the German Federal Republic concerning breeder reactors with fast neutrons. In particular, it describes in detail the causes for the delays in the construction of the demonstration reactor SRN 300 (Kalkar) for which the commissioning has now been fixed for 1987. It emphasizes the importance of the collaboration which exists in Europe for the establishment of the projects and the construction of fast neutron power stations, including the SNR 2 reactor; the construction of the latter could begin in the German Federal Republic at the end of the Eighties [fr

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

SWR 1000: an advanced boiling water reactor with passive safety features

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

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

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

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

Nordic studies in reactor safety

International Nuclear Information System (INIS)

Pershagen, N.

1993-01-01

The Nordic Nuclear Safety Research Programme SIK programme in reactor safety is part of a major joint Nordic research effort in nuclear safety. The report summarizes the achievements of the SIK programme, which was carried out during 1990-1993 in collaboration between Nordic nuclear utilities, safety authorities, and research institutes. Three main projects were successfully completed dealing with: 1) development and application of a living PSA concept for monitoring the risk of core damage, and of safety indicators for early warning of possible safety problems; 2) review and intercomparison of severe accident codes, case studies of potential core melt accidents in nordic reactors, development of chemical models for the MAAP code, and outline of a system for computerized accident management support; 3) compilation of information about design and safety features of neighbouring reactors in Germany, Lithuania and Russia, and for naval reactors and nuclear submarines. The report reviews the state-of-the-art in each subject matter as an introduction to the individual project summaries. The main findings of each project are highlighted. The report also contains an overview of reactor safety research in the Nordic countries and a summary of fundamental reactor safety principles. (au) (69 refs.)

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.

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

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

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

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

Lessons from American-German nuclear power plant construction. Quality, safety and costs of an attempt to integrate American and German nuclear power plant technology

International Nuclear Information System (INIS)

Buchwald, K.

1979-05-01

The 1300 MW nuclear power plant at Muelheim-Kaerlich has been under construction since the beginning of 1975. It is being equipped with a pressurised water reactor which has been adapted to the German client's requirements and German licensing practice, based on a license held by Babcock and Wilcox USA (B and W). The problems which have arisen in making this adaptation are the result of different requirements in the USA and the Federal Republic of Germany which make it very difficult to integrate the two technologies. Full integration will almost certainly be impossible, but integration to the widest possible extent is important because it might mean both greater safety and reduced costs. In this article it is intended to show where the problems of integration lie and how they might perhaps be overcome. (author)

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

Selecting of key safety parameters in reactor nuclear safety supervision

International Nuclear Information System (INIS)

He Fan; Yu Hong

2014-01-01

The safety parameters indicate the operational states and safety of research reactor are the basis of nuclear safety supervision institution to carry out effective supervision to nuclear facilities. In this paper, the selecting of key safety parameters presented by the research reactor operating unit to National Nuclear Safety Administration that can express the research reactor operational states and safety when operational occurrence or nuclear accident happens, and the interrelationship between them are discussed. Analysis shows that, the key parameters to nuclear safety supervision of research reactor including design limits, operational limits and conditions, safety system settings, safety limits, acceptable limits and emergency action level etc. (authors)

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

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

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

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

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

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)

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

German research reactor back-end provisions

International Nuclear Information System (INIS)

Koester, Siegfried; Gruber, Gerhard

2002-01-01

Germany has several types of Research Reactors in operation. These reactors use fuel containing uranium of U.S. origin. Basically all the fuel which will be spent until May 2006 will be returned to the U.S. under existing contracts with the U.S. Department of Energy. The contracts are based on the U.S. FRR SNF (Foreign Research Reactor Spent Nuclear Fuel) Program which started in May 1996 and which will last for 10 years. In 1990, the German Federal Government started a program to long-term store (approx. 40 years) and finally dispose of spent fuel in Germany after the so-called U.S. fuel return window will be closed. In order to long-term store the fuel, a special container was designed which covers all different types of spent fuel from the Research Reactors. The container called 'CASTOR MTR 2' is basically licensed and is already in use for the spent fuel of Russian origin from the 'Research Reactor Rossendorf' in the eastern part of Germany. All that fuel is expected to be stored in the existing intermediate storage facility, the so-called BZA (Brennelemente Zwischenlager Ahaus). BZA already accomodates spent fuel from the former THTR-300 high temperature reactor. A final repository does not yet exist in Germany. Alternative provisions to close the back-end of the Research Reactor fuel cycle are reprocessing at COGEMA (France) or in Russian facilities, perspectively. Waste return in a form to be agreed will be mandatory, at least in France. (author)

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

Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

International Nuclear Information System (INIS)

1997-01-01

In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

Basic aspects and results of the German risk study

International Nuclear Information System (INIS)

Bayer, A.; Heuser, F.W.

1981-01-01

This article presents an overview of the investigations and results of the German Risk Study (Phase A). Similar to its American counterpart [Reactor Safety Study (WASH-1400)], the German Risk Study assesses the societal risks associated with potential accidents in nuclear power plants sited in the Federal Republic of Germany. The technical part of the analysis was performed for a representative pressurized-water-reactor nuclear power plant of the 1300-MW(e) class. For the risk assessment, 19 sites were considered, with a total of 25 reactor units presently in operation, under construction, or undergoing the licensing procedure. In the spring of 1981 a translation of the main report [German Risk Study-Main Report (EPRI-NP-1804-SR)], including the investigations and results of Phase A, was published by the Electric Power Research Institute

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

Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

International Nuclear Information System (INIS)

1978-09-01

The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas

Assessment of accident risks from german nuclear plants

International Nuclear Information System (INIS)

Heuser, F.W.

1979-01-01

The German risk study are presented. The main objectives can be summed up as follows: (a) An assessment of the societal risk due to accidents in nuclear power plants with reference to German conditions; (b) To get experience in the field of risk analysis and to provide a basis for estimation of uncertainties; (c) To provide guidance for future activities in the German Reactor Safety Research Program. Finally several conclusions reached by this study are discussed. (author)

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

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.)

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)

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

The implications of the German Risk Study

International Nuclear Information System (INIS)

Birkhofer, A.; Koeberlein, K.

1980-01-01

The methods and results of the German Risk Study published in 1979 are summarized and its implications for reactor safety are discussed. It has led to suggestions that risk analysis should be more widely used for nuclear and other technological systems. It has also identified the need for specific system modifications and confirmed trends in safety research. (author)

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 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

Comparison of radioactive doses after the last protection layer insight the reactor structure for Russian VVER-1000 and German PWR-1300 reactors

International Nuclear Information System (INIS)

Rahimi, A.; Mansourshaiflu, N.; Alizadeh, M. R.

2004-01-01

In pressurized reactors (VVER and PWR), various protections layers are used for reducing the output core doses. At any protection layer, some amount of neutron and gamma doses is reduced. In this project the axial flux of neutron and gamma beams have been evaluated at various protection layers in the operation state the German PWR-1300 and Russian VVER-1000 reactors by the MCNP computer code. For the purpose of effective use of the MCNP code and assuring its correct performance about of fluxed beams common and series of scientific answers and bench marks should be considered and the results obtained by the MCNP code, be compared with this answers. Then by using appropriate method, for reducing the flux variants of neutron and gamma beams at various protection layers of German PWR-1300 and Russian VVER-1000 reactors of the operation state of both reactors have been accelerated. In this projects, bench marks are computations and numbers existing in PSAR's present at Bushehr nuclear power plant. At the end, by using the results obtained and the standard doses, the time which a person can have work activity at the reactor wall (after the last protection layer), was compared for the operation status of the German PWR-1300 and Russian VVER-1000 reactors

Breeders: operational experience with fast power reactors in five states - more intensive German-French breeder cooperation

International Nuclear Information System (INIS)

Hueper, R.

1978-01-01

In the past year contracts have been signed and implemented for German-French cooperation in LMFBR development and commercialization. - The first German nuclear power station with a sodium cooled fast reactor, KNK II in Karlsruhe, is going into operation. - Construction of the prototype SNR 300 at Kalkar (Lower Rhine) is slowing down awaiting a decision of the German Federal Constitutional Court. - On the international level, remarkable experience in the operation of fast power reactors has accumulated. - Possible fuel cycle alternatives are being evaluated by an international committee. (orig.) [de

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)

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

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

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

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

Space reactor safety, 1985--1995 lessons learned

International Nuclear Information System (INIS)

Marshall, A.C.

1995-01-01

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration

Space reactor safety, 1985--1995 lessons learned

Energy Technology Data Exchange (ETDEWEB)

Marshall, A.C.

1995-12-31

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

Catalogue of requirements for a plant-specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events; Anforderungskatalog fuer anlagenbezogene Ueberpruefungen deutscher Kernkraftwerke unter Beruecksichtigung der Ereignisse in Fukushima-I (Japan)

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-03-30

The catalogue of requirements for a plant-specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events worked out by the German RSK (reactor safety commission) includes the following inspection topics: natural events like earth quakes, floods, weather-based consequences and possible superposition; civilization-based events like airplane crash, gas release, reactor accident consequences for neighboring units, terroristic impacts, external attacks on computer-based control systems. Further event-independent assumptions have to be considered: station blackout, long-term emergency power supply requirement, failure of auxiliary cooling water supply, efficacy of preventive measures, aggravating boundary conditions for the performance of emergency measures.

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)

Study of the accidental risk of the German fast breeder prototype reactor SNR-300

International Nuclear Information System (INIS)

Koeberlein, K.

1983-01-01

A fact-finding committee of the German Federal Parliament in July 1980 recommended to perform a 'risk-oriented study' of the SNR-300, the German 300 MWe fast breeder prototype reactor being under construction at Kalkar. The main aim of this study was to allow a comparative safety evaluation between the SNR-300 and a modern PWR, thus to prepare a basis for a political decision on the SNR-300. Methods and main results of the study are presented in this paper. In the first step of the study six groups of accidents have been identified which may initiate core destruction. By reliability analyses, expected frequency of each group has been estimated. Conditional probabilities for conceivable reactor tank failure modes have been analysed. Tank failure after core destruction leads to release of energy and radioactive material into the containment system. Such accident sequences have been pursued further. Based on a number of core destruction initiators and tank failure modes and various combinations of success and failure states of the containment systems, detailed calculations of different containment scenarious were carried out. From the results of the plant systems analysis, five release categories have been defined. Possible effects of external events and releases of radioactivity from the spent fuel storage pool have also been analysed. In order to quantify the degree of uncertainty of the calculated frequencies, subjective probability distributions of fixed, but inaccurately known quantities have been propagated through the calculations. Using the release categories as input, accident consequences were calculated for the site Kalkar. Though the uncertainty bandwidths for the accident frequencies estimated in the SNR-300 analysis are much wider than for the PWR, the analysis indicates that frequencies of severe accidents, and consequences, are smaller for the SNR-300 than for the PWR as analysed in the 'German Risk Study'. (orig.)

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

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)

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.

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.)

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)

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

The AFR. An approved network of research reactors

Energy Technology Data Exchange (ETDEWEB)

Hampel, Gabriele [Mainz Univ. (Germany). Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren (AFR)

2012-10-15

AFR (Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren) is the German acronym for 'Association for Research Reactor Operation and Safety Issues' which was founded in 1959. Reactor managers of European research reactors mainly from the German linguistic area meet regularly for their mutual benefit to exchange experience and knowledge in all areas of operating, managing and utilization of research reactors. In the last 2 years joint meetings were held together with the French association of research reactors CER (Club d'Exploitants des Reacteurs). In this contribution the AFR, its members, work and aims as well as the French partner CER are presented. (orig.)

German Phase B [risk study] highlights the role of [reactor] accident management

International Nuclear Information System (INIS)

Anon.

1989-01-01

Phase B of the German probabilistic risk assessment study, now scheduled for publication this month, suggests that reactor accident management measures can prevent or mitigate about 90 per cent of event sequences. (author)

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.)

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

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 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 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

EPR (European Pressurized water Reactor) The advanced nuclear reactor

International Nuclear Information System (INIS)

2004-01-01

Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

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

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)

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)

Reactor safety in industrial nuclear power plants. Developments in a political and technical environment as it prevails in the German Federal Republic; Reaktorsicherheit fuer Leistungskernkraftwerke. Die Entwicklung im politischen und technischen Umfeld der Bundesrepublik Deutschland

Energy Technology Data Exchange (ETDEWEB)

Laufs, Paul [Stuttgart Univ. (Germany)

2013-07-01

The present book describes the development of reactor safety in German light-water nuclear power plants from the beginnings with its multifarious links to model solutions seen in other countries, national and international research projects and accidents in conventional and nuclear power plants. It contains detailed, richly illustrated articles on specific safety techniques such as rupture safety of the pressurised envelope, safeguarding of emergency cooling in the event of accidents involving coolant loss, control technology and protection of the surroundings through containment. The results of national and international risk studies are discussed. It is shown how efforts on the part of industry, government and science to enhance safety resources and improve safety culture have been driven by a civic environment in which the use of nuclear energy has become a central issue of political debate. The book is written in readily comprehensible language and offers a wealth of material for further study.

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.)

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

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.)

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)

IRSN research programs concerning reactor safety

International Nuclear Information System (INIS)

Bardelay, J.

2005-01-01

This paper is made up of 3 parts. The first part briefly presents the missions of IRSN (French research institute on nuclear safety), the second part reviews the research works currently led by IRSN in the following fields : -) the assessment of safety computer codes, -) thermohydraulics, -) reactor ageing, -) reactivity accidents, -) loss of coolant, -) reactor pool dewatering, -) core meltdown, -) vapor explosion, and -) fission product release. In the third part, IRSN is shown to give a major importance to experimental programs led on research or test reactors for collecting valid data because of the complexity of the physical processes that are involved. IRSN plans to develop a research program concerning the safety of high or very high temperature reactors. (A.C.)

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

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)

[Infection control and safety culture in German hospitals].

Science.gov (United States)

Hansen, Sonja; Schwab, Frank; Gropmann, Alexander; Behnke, Michael; Gastmeier, Petra

2016-07-01

Healthcare-associated infections (HAI) are the most frequent adverse events in the healthcare setting and their prevention is an important contribution to patient safety in hospitals. To analyse to what extent safety cultural aspects with relevance to infection control are implemented in German hospitals. Safety cultural aspects of infection control were surveyed with an online questionnaire; data were analysed descriptively. Data from 543 hospitals with a median of [IQR] 275 [157; 453] beds were analysed. Almost all hospitals (96.6 %) had internal guidelines for infection control (IC) in place; 82 % defined IC objectives, most often regarding hand hygiene (HH) (93 %) and multidrug resistant organisms (72 %) and less frequently for antibiotic stewardship (48 %) or prevention of specific HAI. In 94 % of hospitals, a reporting system for adverse events was in place, which was also used to report low compliance with HH, outbreaks and Clostridium difficile-associated infections. Members of the IC team were most often seen to hold daily responsibility for IC in the hospital, but rarely other hospital staff (94 versus 19 %). Safety cultural aspects are not fully implemented in German hospitals. IC should be more strongly implemented in healthcare workers' daily routine and more visibly supported by hospital management.

Safety evaluation of the nuclear power plant at Cattenom

International Nuclear Information System (INIS)

Anon.

1987-01-01

This is a systematic compilation of the material which was dealt with at the level of the German-French Commission (on questions of the safety of nuclear installations) in this discussions about the nuclear power plant at Cattenom. As a supplement to the report published already in 1982, the Commission has officially released its deliberation results on the subjects constructive safety measures, radiological effects, and precautions in case of an emergency. The allegations according to which the installation is wanting in safety are countered by the joint statement of the chairmen of GPR (Permanent Group on Reactors) and RSK (German Commission on Reactor Safety) of August 29, 1986. (HSCH) [de

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

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)

Experiences with interpretation and application of the German Atomic Energy Act for the German research reactors in Geesthacht FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1980-01-01

The German research reactors FRG-1 and FRG-2 have passed different types of licensing procedures in the past years. It is reported about the experiences we have got in the interpretation and application of section 7 of the German Atomic Energy Act. Following these experiences an estimation is done for the licensing procedure for the reduction of the uranium enrichment. (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

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

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)

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

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)

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)

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

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)

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

Light water reactor safety research project

International Nuclear Information System (INIS)

Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

1980-07-01

The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

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

Meeting on reactor safety research

International Nuclear Information System (INIS)

1982-09-01

The meeting 'Reactor Safety Research' organized for the second time by the GRS by order of the BMFT gave a review of research activities on the safety of light water reactors in the Federal Repulbic of Germany, international co-operation in this field and latest results of this research institution. The central fields of interest were subjects of man/machine-interaction, operational reliability accident sequences, and risk. (orig.) [de

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

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)

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)

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.)

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

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

Integral fast reactor safety features

International Nuclear Information System (INIS)

Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

1988-01-01

The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: a pool-type primary system, and advanced ternary alloy metallic fuel, and an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

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

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)

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

Perspective channel-type reactor with enhanced safety

International Nuclear Information System (INIS)

Adamov, E.O.; Grozdov, I.I.; Kuznetsov, S.P.; Petrov, A.A.; Rozhdestvensky, M.I.; Cherkashov, Yu.M.

1994-01-01

Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: The design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness. (orig.)

Safety aspects of pressurised water reactors

International Nuclear Information System (INIS)

1985-01-01

This submission to the Health and Safety Executive has been prepared by the Institution of Professional Civil Servants (IPCS) as a contribution to the debate on safety aspects associated with Pressurized Water Reactors (PWRs). Although supporting an energy policy which includes the development of nuclear power, assurances are sought on a number of safety issues if it is decided that this should be generated by a PWR-type reactor. These issues are listed. In particular the following are mentioned: the wider publication of design information, the use of elastic-plastic fracture mechanics as the basis for determining pressure vessel integrity, the failure rate of steam generating units, water coolant quality control, greater investigation of two-phase flow accident conditions, the components of the reactor cooling system and training of reactor personnel in the understanding of LOCA effects. (U.K.)

Reactor safety

International Nuclear Information System (INIS)

Meneley, D.A.

The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

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.)

The European pressurized water reactor

International Nuclear Information System (INIS)

Leny, J.C.

1993-01-01

The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

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)

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

Integral fast reactor safety features

International Nuclear Information System (INIS)

Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

1988-01-01

The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

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)

Technical mechanics in constructional reactor safety

International Nuclear Information System (INIS)

Matthees, W.

1979-01-01

Reactor safety is based on close cooperation between a number of technical and scientific disciplines; most problems of reactor technology can be solved with the aid of technical mechanics. At the 5th International Conference on Structural Mechanics in Reactor Technology (5th SMIRT), one of the biggest conferences in the field of applied technical mechanics, about 800 papers were read giving the latest state of knowledge in the field of constructional reactor safety. The main subject of the conference was the analysis of material behaviour under high loads; the information and methods of these analysis go far beyond what is required in the conventional field. (orig./UA) [de

New perspectives on reactor safety

International Nuclear Information System (INIS)

Avery, R.

1986-01-01

Over the past few years a number of changes and new perspectives have come about in our approach to reactor safety. These changes have occurred over a period of time extending from as long ago as 1975, when WASH-1400 came out representing the first major application of probabilistic risk analysis (PRA) to US reactor plants. The period of change has extended from that time to the present, and includes new areas of focus such as safety goals, source term studies, and severe accident policy statement and approaches, including the IDCOR Program. It has also included a greatly increased interest in inherent safety. These areas are discussed in this paper

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)

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

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

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 Management at PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Ligam, A.S.; Zarina Masood; Ahmad Nabil Abdul Rahim

2011-01-01

Adequate safety measures and precautions, which follow relevant safety standards and procedures, should be in place so that personnel safety is assured. Nevertheless, the public, visitor, contractor or anyone who wishes to enter or be in the reactor building should be well informed with the safety measures applied. Furthermore, these same elements of safety are also applied to other irradiation facilities within the premises of Nuclear Malaysia. This paper will describes and explains current safety management system being enforced especially in the TRIGA PUSPATI Reactor (RTP) namely radiation monitoring system, safety equipment, safe work instruction, and interconnected internal and external health, safety and security related departments. (author)

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

The next generation of power reactors - safety characteristics

International Nuclear Information System (INIS)

Modro, S.M.

1995-01-01

The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

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)

Nuclear reactor safety in the USA

International Nuclear Information System (INIS)

Vigil, J.C.

1983-01-01

Nuclear reactor safety in the USA has emphasized a defense-in-depth approach to protecting the public from reactor accidents. This approach was severely tested by the Three Mile Island accident and was found to be effective in safeguarding the public health and safety. However, the economic impact of the TMI accident was very large. Consequently, more attention is now being given to plant protection as well as public-health protection in reactor-safety studies. Sophisticated computer simulations at Los Alamos are making major contributions in this area. In terms of public risk, nuclear power plants compare favorably with other large-scale alternatives to electricity generation. Unfortunately, there is a large gulf between the real risks of nuclear power and the present public perception of these risks

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)

Reactor safety systems

International Nuclear Information System (INIS)

Kafka, P.

1975-01-01

The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.) [de

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.

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

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

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

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

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.

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

Nuclear power reactor safety

International Nuclear Information System (INIS)

Pon, G.A.

1976-10-01

This report is based on the Atomic Energy of Canada Limited submission to the Royal Commission on Electric Power Planning on the safety of CANDU reactors. It discusses normal operating conditions, postulated accident conditions, and safety systems. The release of radioactivity under normal and accident conditions is compared to the limits set by the Atomic Energy Control Regulations. (author)

Risk reduction category (RRC-A) accident studies in the safety analysis report of the EPR trademark reactor

International Nuclear Information System (INIS)

Poehlmann, M.; Bleher, G.; Ismaier, A.; Knoll, A.; Levi, P.; Garcia, E. Vera; Schels, A.; Seitz, H.; Lima Campos, L.

2013-01-01

The Risk Reduction Category (RRC-A) is considered in the safety demonstration of nuclear reactors in addition to design basis operating conditions (Plant Condition Category, PCC), in order to analyze with a risk reduction approach any operating conditions with multiple failures. As extending the operating conditions of the plant 'beyond design basis', the Risk Reduction Category (RRC-A) is also denoted as Design Extension Condition (DEC-A). In the German licensing framework, the RRCA (or DEC-A) transients correspond to safety assessment level '4b' of the 'Sicherheitsanforderungen an Kernkraftwerke' (Safety Requirements for Nuclear Power Plants), Az. RS I 5 - 13303/01 of the German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety. These RRC-A (or DEC-A) operating conditions require specific design provisions (implemented by manual or automatic action), known as RRC-A measures, intended to render consequences of accumulated failures admissible. In contrast, RRC-B constitute severe accidents that lead to core melt. Identification of RRC-A operating conditions and corresponding RRC-A measures is based on the use of results of probabilistic safety assessments. After the Fukushima accident the RRC-A accidents like Station Black Out (SBO) or Loss of Ultimate Heat Sink (LUHS) are of particular interest in the safety assessment of nuclear new builds. In several chapters of the Safety Analysis Report it is demonstrated that the AREVA EPRTM design is resistant at RRC-A accident conditions. (orig.)

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

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

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

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

Safety of research reactors. Topical issues paper no. 4

International Nuclear Information System (INIS)

Alcala-Ruiz, F.; Ferraz-Bastos, J.L.; Kim, S.C.; Voth, M.; Boeck, H.; Dimeglio, F.; Litai, D.

2001-01-01

Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety review of the research reactor facility and to verify compliance with the IAEA Safety Standards. The methods used during an INSARR mission have been collected and analysed. Some of the important issues identified are the following: general ageing of the facility; uncertain status of many research reactors (in extended shutdown); indefinite deferral of return to operation or decommissioning; inadequate regulatory supervision; insufficient systematic (periodic) reassessment of safety; lack of quality assurance (QA) programmes; lack of an international safety convention or arrangement; lack of financial support for safety measures (e.g. safety reassessment, safety upgrading, decommissioning) and utilization; lack of clear utilization programmes; inadequate emergency preparedness; inadequate safety documentation (e.g. safety analysis report, operating rules and procedures, emergency plan); inadequate funding of shutdown reactors; weak safety culture; loss of expertise and corporate memory; loss of information concerning radioactive materials contained in retired experimental devices stored in the facility indefinitely; obsolescence of equipment and lack of spare parts; inadequate training and qualifications of regulators and operators; safety implications of new fuel types. These issues have been addressed by the IAEA Secretariat and the chairman of the International Nuclear Safety Advisory Group (INSAG). INSAG has identified three major safety issues that are: the increasing age of research reactors, the number of research reactors that are not operating anymore but have not been decommissioned, and the number of research reactors in countries that do not have appropriate regulatory authorities. This issue paper discusses the concerns generated by an analysis of the results of INSARR missions and those expressed by INSAG. The

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)

German data for risk based fire safety assessment

International Nuclear Information System (INIS)

Roewekamp, M.; Berg, H.P.

1998-01-01

Different types of data are necessary to perform risk based fire safety assessments and, in particular, to quantify the fire event tree considering the plant specific conditions. Data on fire barriers, fire detection and extinguishing, including also data on secondary effects of a fire, have to be used for quantifying the potential hazard and damage states. The existing German database on fires in nuclear power plants (NPPs) is very small. Therefore, in general generic data, mainly from US databases, are used for risk based safety assessments. Due to several differences in the plant design and conditions generic data can only be used as conservative assumptions. World-wide existing generic data on personnel failures in case of fire fighting have only to be adapted to the plant specific conditions inside the NPP to be investigated. In contrary, unavailabilities of fire barrier elements may differ strongly depending on different standards, testing requirements, etc. In addition, the operational behaviour of active fire protection equipment may vary depending on type and manufacturer. The necessity for more detailed and for additional plant specific data was the main reason for generating updated German data on the operational behaviour of active fire protection equipment/features in NPPs to support risk based fire safety analyses being recommended to be carried out as an additional tool to deterministic fire hazard analyses in the frame of safety reviews. The results of these investigations revealed a broader and more realistic database for technical reliability of active fire protection means, but improvements as well as collection of further data are still necessary. (author)

The reactor accident in Fukushima Daiichi. The consequence of design deficiencies and inadequate safety engineering; Der Reaktorunfall in Fukushima Daiichi. Folge fehlerhafter Auslegung und unzureichender Sicherheitstechnik

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-03-15

The reactor accident in Fukushima Daiichi is discussed in the frame of design deficiencies and inadequate safety engineering. The progress of the accident as consequence of the earthquake and the tsunami is described. The radiological situation for the public is supposed to be blow the dose limit of 20 mSv/year. The WHO and UNSCEAR (United Nations Scientific Committee on the Effects of Atomic radiation) did not observe acute radiation injuries. The Japanese authorities have classified the accident to 7 of the INES scale. The German Atomforum e.V. considers the safety engineering of German NPPs to be superior to the Japanese situation due to higher emergency energy supply, extensive measures to reduce the hydrogen accumulation and mitigating measures for the accident management. German NPPS are considered highly robust as the EU stress tests have shown.

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

Advanced nuclear reactor safety issues and research needs

International Nuclear Information System (INIS)

2002-01-01

On 18-20 February 2002, the OECD Nuclear Energy Agency (NEA) organised, with the co-sponsorship of the International Atomic Energy Agency (IAEA) and in collaboration with the European Commission (EC), a Workshop on Advanced Nuclear Reactor Safety Issues and Research Needs. Currently, advanced nuclear reactor projects range from the development of evolutionary and advanced light water reactor (LWR) designs to initial work to develop even further advanced designs which go beyond LWR technology (e.g. high-temperature gas-cooled reactors and liquid metal-cooled reactors). These advanced designs include a greater use of advanced technology and safety features than those employed in currently operating plants or approved designs. The objectives of the workshop were to: - facilitate early identification and resolution of safety issues by developing a consensus among participating countries on the identification of safety issues, the scope of research needed to address these issues and a potential approach to their resolution; - promote the preservation of knowledge and expertise on advanced reactor technology; - provide input to the Generation IV International Forum Technology Road-map. In addition, the workshop tried to link advancement of knowledge and understanding of advanced designs to the regulatory process, with emphasis on building public confidence. It also helped to document current views on advanced reactor safety and technology, thereby contributing to preserving knowledge and expertise before it is lost. (author)

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)

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

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)

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

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

Announcement of recommendations of the Reaktor-Sicherheitskommission. As of 24 July 1997. Joint recommendations of RSK and GPR for safety requirements of future nuclear PWR-type power plants. English versions published in the years 1995 through 1997

International Nuclear Information System (INIS)

1997-01-01

The recommendations, most parts given in English, refer to the European Pressurized Water Reactor (EPR) and have been established by the German RSK (reactor safety commission), the corresponding French organization GPR and the German SSK (radiation protection commission). This publication continues earlier joint recommendations by the national bodies, last published by the German BMU (responsible German ministry) on 5 May 1995, in BAnz. page 7452. The safety recommendations establish the basis for further activities in the Franco-German project for development of the EPR, a PWR type reactor of the next generation. (CB) [de

Safety upgrades to the NRU research reactor

International Nuclear Information System (INIS)

DeAbreu, B.; Mark, J.M.; Mutterback, E.J.

1998-01-01

The NRU (National Research Universal) Reactor is a 135 MW thermal research facility located at Chalk River Laboratories, and is owned and operated by Atomic Energy of Canada Limited. One of the largest and most versatile research reactors in the world, it serves as the R and D workhorse for Canada's CANDU business while at the same time filling the role as one of the world's major producers of medical radioisotopes. AECL plans to extend operation of the NRU reactor to approximately the year 2005 when a new replacement, the Irradiation Research Facility (IRF) will be available. To achieve this, AECL has undertaken a program of safety reassessment and upgrades to enhance the level of safety consistent with modem requirements. An engineering assessment/inspection of critical systems, equipment and components was completed and seven major safety upgrades are being designed and installed. These upgrades will significantly reduce the reactor's vulnerability to common mode failures and external hazards, with particular emphasis on seismic protection. The scheduled completion date for the project is 1999 December at a cost approximately twice the annual operating cost. All work on the NRU upgrade project is planned and integrated into the regular operating cycles of the reactor; no major outages are anticipated. This paper describes the safety upgrades and discusses the technical and managerial challenges involved in extending the operating life of the NRU reactor. (author)

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)

Modular high-temperature reactor launched (and wallchart)

International Nuclear Information System (INIS)

Steinwarz, W.

1987-01-01

In view of the need for a technically unsophisticated, safe and economic reactor system, the KWU group has integrated the experience gained from German light-water reactor engineering and from successful operation of the German AVR experimental high-temperature reactor into the development of the High-Temperature Reactor (HTR)-module. The main components are illustrated and explained and technical data for the HTR-module is given. Safety is also considered. This includes graphs of core heat-up temperature for pebble-bed HTR and a graph of the temperature load of the fuel elements. The operation, control and applications are considered. The latter includes use in combined heat and power generation and community heating. Feasibility studies have shown that the HTR-module is cheaper, comparatively, than coal-fired power stations. (U.K.)

Problems of nuclear reactor safety. Vol. 1

International Nuclear Information System (INIS)

Shal'nov, A.V.

1995-01-01

Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

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

The organization of research reactor safety in the UKAEA

International Nuclear Information System (INIS)

Redpath, W.

1983-01-01

The present state of organization and development of research reactor safety in the UKAEA are outlined by addressing the fundamental safety principles which have been adopted in keeping with national health and safety requirement. The organisation, assessment and monitoring of research reactor safety on complex multi-discipline and multi-activity nuclear research and development site are discussed. Methods of safety assessment, such as probabilistic risk assessment and risk acceptance criteria, which have been developed and applied in practice are explained, and some indication of the directions in which some of the current developments in the safety of UKAEA research reactors is also included. (A.J.)

Fast neutron reactors: the safety point of view

International Nuclear Information System (INIS)

Laverie, M.; Avenas, M.

1984-01-01

All versions of nuclear reactors present favourable and unfavourable characteristics from the point of view of safety. The safety of the installations is obtained by making efforts to utilize in the best possible way those which are favourable and by taking proper steps in the face of those which are unfavourable. The present article shows how this general principle has been applied as regards the fast neutron reactors of integrated design which have been developped in France, taking into account the specific features of this version. A qualitative method to compare the safety of this version with that of pressurized water reactors which has been widely put to the test commercially all over the world is presented. These analyses make, generally speaking, several positive characteristics stand out for these fast neutron reactors from the safety aspects [fr

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

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

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.

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.

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)

Safety report on WWR-S reactor

International Nuclear Information System (INIS)

Horyna, J.; Kaisler, L.; Listik, E.

1981-04-01

The present Safety Report of the WWR-S reactor summarizes findings obtained during the trial and partially also permanent operation of the reactor after two stages of its reconstruction implemented between 1974 and 1976. Most data are presented necessary for assessing probable risks of possible accident conditions whose consequences pose health hazards to individuals of the population, radiation personnel and the facilities themselves. Attention is devoted to the description of the locality, to components and systems, heat removal from the core, design aspects, the quality of new and old parts of the technological circuits, the systems of protection and control, the emergency core cooling system, the problems of radiation safety, and to the safety analyses of the abnormal states envisaged. The Report was compiled with regard to IAEA and CMEA recommendations concerning safe operation of research reactors and to the recommendations and binding decisions of the Czechoslovak Atomic Energy Commission. (author)

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

International Nuclear Information System (INIS)

Raisic, N.

1962-09-01

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

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

A new safety approach in the design of fast reactors

International Nuclear Information System (INIS)

Neuhold, R.J.; Marchaterre, J.F.; Waltar, A.E.

1987-01-01

A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

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

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

Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

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

Development of Safety Review Guide for the Periodic Safety Review of Reactor Vessel Internals

International Nuclear Information System (INIS)

Park, Jeongsoon; Ko, Hanok; Kim, Seonjae; Jhung, Myungjo

2013-01-01

Aging management of the reactor vessel internals (RVIs) is one of the important issues for long-term operation of nuclear power plants (NPPs). Safety review on the assessment and management of the RVI aging is conducted through the process of a periodic safety review (PSR). The regulatory body should check that reactor facilities sustain safety functions in light of degradation due to aging and that the operator of a nuclear power reactor establishes and implements management program to deal with degradation due to aging in order to guarantee the safety functions and the safety margin as a result of PSR. KINS(Korea Institute of Nuclear Safety) has utilized safety review guides (SRG) which provide guidance to KINS staffs in performing safety reviews in order to assure the quality and uniformity of staff safety reviews. The KINS SRGs for the continued operation of pressurized water reactors (PWRs) published in 2006 contain areas of review regarding aging management of RVIs in chapter 2 (III.2.15, Appendix 2.0.1). However unlike the SRGs for the continued operation, KINS has not officially published the SRGs for the PSR of PWRs, but published them as a form of the research report. In addition to that, the report provides almost same review procedures for aging assessment and management of RVIs with the ones provided in the SRGs for the continued operation, it cannot provide review guidance specific to PSRs. Therefore, a PSR safety review guide should be developed for RVIs in PWRs. In this study, a draft PSR safety review guide for reactor vessel internals in PWRs is developed and provided. In this paper, a draft PSR safety review guide for reactor vessel internals (PSR SRG-RVIs) in PWRs is introduced and main contents of the draft are provided. However, since the PSR safety review guides for areas other than RVIs in the pressurized water reactors (PWRs) are expected to be developed in the near future, the draft PSR SRG-RVIs should be revisited to be compatible with

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)

Thermohydraulic accident behavior of reactors

International Nuclear Information System (INIS)

Horche, W.; Kirmse, R.; Reichenbach, D.; Weber, J.P.

1992-01-01

GRS, on behalf of the German Federal Ministry for the Environment, conducted an assessment of the technical safety of the Greifswald nuclear generating units of the Soviet WWER-440/W-230 and W-213 reactor lines, respectively. The evaluation of existing accident analyses and the execution of some first calculations by GRS added to the know-how of GRS. This is reflected in the increased participation by GRS in international expert bodies investigating safety problems of WWER-440 plants. The contributions made towards international WWER projects within the framework of IAEA missions or as a result of bilateral consultations strengthen international partnership in the field of reactor safety in Central and Eastern Europe. (orig.) [de

The emphasis is on reactor safety research

International Nuclear Information System (INIS)

Anon.

1982-01-01

For the second time the Association for Reactor Safety mbH (GRS), Koeln, organised on behalf of the BMFT the conference 'Reactor safety research'. About 400 visitors took part. The public who were interested were given a review of the activities which are being undertaken by the BMFT in the programme 'Research and safety of light-water reactors'. The series of conference papers initiated by the BMFT is to be developed into a permanent information source which will be of interest to those working on nuclear questions such as official quarters, industry and high schools, and experts who have to give judgements. The most important statements by various research groups in industry, high schools and also associations of experts, are summarised. (orig.) [de

Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

Energy Technology Data Exchange (ETDEWEB)

Kliem, S.

2017-09-15

The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

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)

Proceedings of the international topical meeting on advanced reactors safety: Volume 2

International Nuclear Information System (INIS)

1997-01-01

In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

Safety of RBMK reactors: Major results and prospects

International Nuclear Information System (INIS)

Sidorenko, V.A.

1996-01-01

The paper considers the following issues: basic reasons for the advent of NPPs with RBMK reactors; the logic of identifying top-priority measures immediately after the accident; top-priority measures for improving the safety and reliability of NPPs with RBMK reactors; upgrading NPPs with RBMK reactors in compliance with the Norms; programmes for retrofitting and upgrading of NPPs of the ''Rosnergoatom'' Concern and progress with their implementation as of April 1996; the safety of RBMK plants and the programmes of its enhancement with regard to modern requirements in the light of national and international assessment; objective indicators of safety, reliability, and economic efficiency of NPPs with RBMK reactors; economics: rationale for continuing plants operation till the end of their design lifetime. 8 refs, 3 figs

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

Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

International Nuclear Information System (INIS)

Wittenbrock, N.G.

1982-01-01

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

Current safety issues related to research reactor operation

International Nuclear Information System (INIS)

Alcala-Ruiz, F.

2000-01-01

The Agency has included activities on research reactor safety in its Programme and Budget (P and B) since its inception in 1957. Since then, these activities have traditionally been oriented to fulfil the Agency's functions and obligations. At the end of the decade of the eighties, the Agency's Research Reactor Safety Programme (RRSP) consisted of a limited number of tasks related to the preparation of safety related publications and the conduct of safety missions to research reactor facilities. It was at the beginning of the nineties when the RRSP was upgraded and expanded as a subprogramme of the Agency's P and B. This subprogramme continued including activities related to the above subjects and started addressing an increasing number of issues related to the current situation of research reactors (in operation and shut down) around the world such as reactor ageing, modifications and decommissioning. The present paper discusses some of the above issues as recognised by various external review or advisory groups (e.g., Peer Review Groups under the Agency's Performance Programme Appraisal System (PPAS) or the standing International Nuclear Safety Advisory Group (INSAG)) and the impact of their recommendations on the preparation and implementation of the part of the Agency's P and B relating to the above subject. (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.

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.)

The unique safety challenges of space reactor systems

International Nuclear Information System (INIS)

Lanes, S.J.; Marshall, A.C.

1991-01-01

Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power

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)

German precursor study: methods and results

International Nuclear Information System (INIS)

Hoertner, H.; Frey, W.; von Linden, J.; Reichart, G.

1985-01-01

This study has been prepared by the GRS by contract of the Federal Minister of Interior. The purpose of the study is to show how the application of system-analytic tools and especially of probabilistic methods on the Licensee Event Reports (LERs) and on other operating experience can support a deeper understanding of the safety-related importance of the events reported in reactor operation, the identification of possible weak points, and further conclusions to be drawn from the events. Additionally, the study aimed at a comparison of its results for the severe core damage frequency with those of the German Risk Study as far as this is possible and useful. The German Precursor Study is a plant-specific study. The reference plant is Biblis NPP with its very similar Units A and B, whereby the latter was also the reference plant for the German Risk Study

Pumps for German pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Dernedde, R.

1984-01-01

The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

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)

Historical perspective of thermal reactor safety in light water reactors

International Nuclear Information System (INIS)

Levy, S.

1986-01-01

A brief history of thermal reactor safety in U.S. light water reactors is provided in this paper. Important shortcomings in safety philosophy evolution versus time are identified and potential corrective actions are suggested. It should be recognized, that this analysis represents only one person's opinion and that most historical accountings reflect the author's biases and specific areas of knowledge. In that sense, many of the examples used in this paper are related to heat transfer and fluid flow safety issues, which explains why it has been included in a Thermal Hydraulics session. One additional note of caution: the value of hindsight and the selective nature of human memory when looking at the past cannot be overemphasized in any historical perspective

Safety status of Russian research reactors

International Nuclear Information System (INIS)

Morozov, S.I.

2001-01-01

Gosatomnadzor of Russia is conducting the safety regulation and inspection activity related to nuclear and radiation safety at nuclear research facilities, including research reactors, critical assemblies and sub-critical assemblies. It implies implementing three major activities: 1) establishing the laws and safety standards in the field of research reactors nuclear and radiation safety; 2) research reactors licensing; and 3) inspections (or license conditions tracking and inspection). The database on nuclear research facilities has recently been updated based on the actual status of all facilities. It turned out that many facilities have been shutdown, whether temporary or permanently, waiting for the final decision on their decommissioning. Compared to previous years the situation has been inevitably changing. Now we have 99 nuclear research facilities in total under Gosatomnadzor of Russia supervision (compared to 113 in previous years). Their distribution by types and operating organizations is presented. The licensing and conduct of inspection processes are briefly outlined with emphasis being made on specific issues related to major incidents that happened in 2000, spent fuel management, occupational exposure, effluents and emissions, emergency preparedness and physical protection. Finally, a summary of problems at current Russian research facilities is outlined. (author)

German offsite accident consequence model for nuclear facilities: further development and application

International Nuclear Information System (INIS)

Bayer, A.

1985-01-01

The German Offsite Accident Consequence Model - first applied in the German Risk Study for nuclear power plants with light water reactors - has been further developed with the improvement of several important submodels in the areas of atmospheric dispersion, shielding effects of houses, and the foodchains. To aid interpretation, the presentation of results has been extended with special emphasis on the presentation of the loss of life expectancy. The accident consequence model has been further developed for application to risk assessments for other nuclear facilities, e.g., the liquid metal fast breeder reactor (SNR-300) and the high temperature gas cooled reactor. Moreover the model have been further developed in the area of optimal countermeasure strategies (sheltering, evacuation, etc.) in the case of the Central European conditions. Preliminary considerations has been performed in connection with safety goals on the basis of doses

Code of Conduct on the Safety of Research Reactors

International Nuclear Information System (INIS)

2006-09-01

The Board of Governors of the International Atomic Energy Agency (IAEA) adopted the Code of Conduct on the Safety of Research Reactors on 8 March 2004. The Board's action was the culmination of several years of work to develop the Code and obtain a consensus on its provisions. The process leading to the Code began in 1998, when the International Nuclear Safety Advisory Group (INSAG) informed the Director General of concerns about the safety of research reactors. In 2000, INSAG recommended that the Secretariat begin developing an international protocol or a similar legal instrument to address those concerns. In September 2000, in resolution GC(44)/RES/14, the General Conference requested the Secretariat ''within its available resources, to continue work on exploring options to strengthen the international nuclear safety arrangements for civil research reactors, taking due account of input from INSAG and the views of other relevant bodies''. A working group convened by the Secretariat pursuant to that request recommended that ''the Agency consider establishing an international action plan for research reactors'' and that the action plan include preparation of a Code of Conduct ''that would clearly establish the desirable attributes for management of research reactor safety''. In September 2001, the Board requested that the Secretariat develop and implement, in conjunction with Member States, an international research reactor safety enhancement plan which included preparation of a Code of Conduct on the Safety of Research Reactors. Subsequently, in resolution GC(45)/RES/10.A, the General Conference endorsed the Board's request. Pursuant to that request, a Code of Conduct on the Safety of Research Reactors was drafted at two meetings of an Open-ended Working Group of Legal and Technical Experts. This draft Code of Conduct was circulated to all Member States for comment. On the basis of the responses received, a revised draft of the Code was prepared by the Secretariat

Safety significance of ATR [Advanced Test Reactor] passive safety response attributes

International Nuclear Information System (INIS)

Atkinson, S.A.

1989-01-01

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab

Refurbishment and safety upgradation of research reactor Cirus

International Nuclear Information System (INIS)

Marik, S.K.; Rao, D.V.H.; Bhatnagar, A.; Pant, R.C.; Tikku, A.C.; Sankar, S.

2006-01-01

Cirus, a 40 MW t, vertical tank type research reactor, having wide range of research facilities, was commissioned in the year 1960. This research reactor, situated at Mumbai, India has been operated and utilized extensively for isotope production, material testing and neutron beam research for nearly four decades. With a view to assess the residual life of the reactor, detailed ageing studies were carried out during the early 1990s. Based on these studies, refurbishment of Cirus for its life extension was taken up. During refurbishment, additional safety features were incorporated in various systems to qualify them for the current safety standards. This paper gives the details of the operating experiences, utilization of the reactor along with methodologies followed for carrying out detailed ageing studies, refurbishment and safety upgradation for its life extension

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

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.

Tritium management in fusion reactors

International Nuclear Information System (INIS)

Galloway, T.R.

1978-05-01

This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-09-01

The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

Safety of intrinsically safe and economical reactor (ISER)

International Nuclear Information System (INIS)

Asahi, Y.; Sugawara, I.; Yamanaka, K.

1988-01-01

Inherent safety of a reactor may be quantified by the grace period at various safety levels such as maintenance of fuel integrity, maintenance of fuel coolability and avoidance of core-melt. It is important to find out the grace period especially at the safety level of maintenance of fuel integrity. It has been conducted to design the ISER, which is characterized by the steel-made reactor pressure vessel. In addition to the passive nature of the safety design of the reactor itself, the ISER is equipped in the secondary system with a subsystem called the passive safety and shutdown system (PSSS), which will help to increase the grace period. It was found by the null transient analysis that check valves are needed at the top hot/cold interface. The analysis of the station blackout, which is one of the severest accident conceivable for the ISER, was made to examine inherent safety of the ISER with and without the PSSS. This paper reports that found out that the PSSS enhances inherent safety of the ISER

Perspectives on reactor safety

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Perspectives on reactor safety

International Nuclear Information System (INIS)

Haskin, F.E.

1994-03-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

Perspectives on reactor safety. Revision 1

International Nuclear Information System (INIS)

Haskin, F.E.; Hodge, S.A.

1997-11-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

Perspectives on reactor safety. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

1997-11-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

VVER Reactor Safety in Eastern Europe and Former Soviet Union

Science.gov (United States)

Papadopoulou, Demetra

2012-02-01

VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

Old and new ways in reactor technology. Reactor concepts and reactor safety

Energy Technology Data Exchange (ETDEWEB)

Schulten, R

1989-01-01

Compared to developments of other technical-scale systems, the period between the recognition of the underlying physics of nuclear fission and the development of a functioning nuclear reactor and its further development to the present level of maturity has been relatively short. The whole development is based on the chain reaction and is rendered safe by the possible auto-stabilization of this reaction. Consequently, the safety of nuclear reactors properly designed is based on automatic mechanisms, which prevent spreads of radioactivity even in major accidents. Controversial opinions about nuclear power uses are mostly based on wrong perceptions both of reactor safety and of radioactive waste, unless they are characterized by sheer ideology. The use of nuclear power worldwide has assumed an important, growing role in the combined uses of a variety energy sources in a surprisingly short period of time and will continue to make a safe, economic, and thus responsible contribution in the long run.

Concept of safety related I and C and power supply systems in the passive safety concept of the HTR-module

International Nuclear Information System (INIS)

Juengst, U.

1990-01-01

The main motivation for the passive safety concepts is to gain a better quality of safety or at least to achieve higher public acceptance for nuclear power plants. This strategy has been introduced into the European Fast Reactor (EER), a common project of France, UK and Germany is applied stringently to the German high-temperature gas-cooled reactor ''HTR - Module''. The following fields are briefly described in the paper: Safety design features of the HTR - Module, overview of I and C concept, reactor protection system, emergency control room, power supply concept, system arrangement and protection against external hazards, accidents sequence of station black-out. (author). 3 figs

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

Cost/benefit analyses of reactor safety systems

International Nuclear Information System (INIS)

1988-01-01

The study presents a methodology for quantitative assessment of the benefit yielded by the various engineered safety systems of a nuclear reactor containment from the standpoint of their capacity to protect the environment compared to their construction costs. The benefit is derived from an estimate of the possible damage from which the environment is protected, taking account of the probabilities of occurrence of malfunctions and accidents. For demonstration purposes, the methodology was applied to a 1 300-MWe PWR nuclear power station. The accident sequence considered was that of a major loss-of-coolant accident as investigated in detail in the German risk study. After determination of the benefits and cost/benefit ratio for the power plant and the containment systems as designed, the performance characteristics of three subsystems, the leakoff system, annulus exhaust air handling system and spray system, were varied. For this purpose, the parameters which describe these systems in the activity release programme were altered. The costs were simultaneously altered in order to take account of the performance divergences. By varying the performance of the individual sub-systems an optimization in design of these systems can be arrived at

The program for Central and Eastern Europe organized by the German Federal Ministry for the Environment

International Nuclear Information System (INIS)

Gelfort, E.; Wolf, J.; Weil, L.

1996-01-01

One of the main objectives of the German Federal Ministry for the Environment, within the overall concept of the Federal Government to help establish democracy and a social market economy in Eastern Europe, is to contribute to improving the safety of nuclear power plants. Safety evaluations of Soviet-type reactors in the former German Democratic Republic have given rise to findings since 1990 which resulted in qualified cooperation especially in Russia and Ukraine. These activities mainly serve to strengthen the role of supervisory authorities in the interest of nuclear safety, and to make available technical inspection systems for nuclear power plants. German experts also conducted safety analyses and radioactivity measurements in situ. The financial expenditure between 1990 and 1995 totaled approx. DM 150 million. The money was spent on some 75 different projects for which contracts were awarded to qualified German firms. All measures, from analyses to the provision of equipment to training seminars, serve the purpose of providing help towards self-help, together with an adequate transfer of know-how. (orig.) [de

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

I. Reactor safety (including comments on criticisms of WASH-1400)

International Nuclear Information System (INIS)

1976-01-01

A major concern in any nuclear power programme is a reactor accident resulting in a large release of radioactivity to the environment. Serious reactor accidents are possible and the risk of such accidents cannot be reduced to zero i.e. absolute safety cannot be assured. All that can be expected is that the measures used to ensure safety in the design and operation of a reactor are such that the risk of accident is reduced to acceptably low levels. No member of the general public is known to have died or been injured as a result of an accident in over 1000 commercial nuclear power reactor-years. Some accidents in power reactors in operation today have come close enough to an environmental release of radioactivity to cause serious public concern about future safety. Apparent inadequacies in safety practices disclosed by former members of the nuclear power industry have added to this concern. To obtain an objective appraisal of the reactor safety issue this report examines the measures taken in the design and operation of nuclear reactors to reduce the probability of accident to acceptably low levels

Monitoring and reviewing research reactor safety in Australia

International Nuclear Information System (INIS)

Cairns, R.C.; Greenslade, G.K.

1990-01-01

Th research reactors operated by the Australian Nuclear Science and Technology Organization (ANSTO) comprise the 10 MW reactor HIFAR and the 100 kW reactor Moata. Although there are no power reactors in Australia the problems and issues of public concern which arise in the operation of research reactors are similar to those of power reactors although on a smaller scale. The need for independent safety surveillance has been recognized by the Australian Government and the ANSTO Act, 1987, required the Board of ANSTO to establish a Nuclear Safety Bureau (NSB) with responsibility to the Minister for monitoring and reviewing the safety of nuclear plant operated by ANSTO. The Executive Director of ANSTO operates HIFAR subject to compliance with requirements and arrangements contained in a formal Authorization from the Board of ANSTO. A Ministerial Direction to the Board of ANSTO requires the NSB to report to him, on a quarterly basis, matters relating to its functions of monitoring and reviewing the safety of ANSTO's nuclear plant. Experience has shown that the Authorization provides a suitable framework for the operational requirements and arrangements to be organised in a disciplined and effective manner, and also provides a basis for audits by the NSB by which compliance with the Board's safety requirements are monitored. Examples of the way in which the NSB undertakes its monitoring and reviewing role are given. Moata, which has a much lower operating power level and fission product inventory than HIFAR, has not been subject to a formal Authorization to date but one is under preparation

The safety analysis report for nuclear power plants in the Federal Republic of Germany

International Nuclear Information System (INIS)

Kohler, H.A.G.

1976-01-01

On the basis of the standard format for drawing up safety analysis reports for stationary fission reactors - this format was published in 1959 by the competent Federal Ministry for Atomic Energy - a report is made on the format and scope of German safety analysis reports. The reasons for drawing up a standard safety analysis report for nuclear power plants with pressurized water reactor or boiling water reactor and a KWU draft are discussed. (orig./RW) [de

The experiences of research reactor accident to safety improvement

International Nuclear Information System (INIS)

Wiranto, S.

1999-01-01

The safety of reactor operation is the main factor in order that the nuclear technology development program can be held according the expected target. Several experience with research reactor incidents must be learned and understood by the nuclear program personnel, especially for operators and supervisors of RSG-GA. Siwabessy. From the incident experience of research reactor in the world, which mentioned in the book 'Experience with research reactor incidents' by IAEA, 1995, was concluded that the main cause of research reactor accidents is understandless about the safety culture by the nuclear installation personnel. With learn, understand and compare between this experiences and the condition of RSG GA Siwabessy is expended the operators and supervisors more attention about the safety culture, so that RSG GA Siwabessy can be operated successfull, safely according the expected target

Overview of fusion reactor safety

International Nuclear Information System (INIS)

Cohen, S.; Crocker, J.G.

1981-01-01

Present trends in magnetic fusion research and development indicate the promise of commercialization of one of a limited number of inexhaustible energy options early in the next century. Operation of the large-scale fusion experiments, such as the Joint European Torus (JET) and Takamak Fusion Test Reactor (TFTR) now under construction, are expected to achieve the scientific break even point. Early design concepts of power producing reactors have provided problem definition, whereas the latest concepts, such as STARFIRE, provide a desirable set of answers for commercialization. Safety and environmental concerns have been considered early in the development of magnetic fusion reactor concepts and recognition of proplem areas, coupled with a program to solve these problems, is expected to provide the basis for safe and environmentally acceptable commercial reactors. First generation reactors addressed in this paper are expected to burn deuterium and tritium fuel because of the relatively high reaction rates at lower temperatures compared to advanced fuels such as deuterium-deuterium. This paper presents an overwiew of the safety and environmental problems presently perceived, together with some of the programs and techniques planned and/or underway to solve these problems. A preliminary risk assessment of fusion technology relative to other energy technologies is made. Improvements based on material selection are discussed. Tritium and neutron activation products representing potential radiological hazards in fusion reactor are discussed, and energy sources that can lead to the release of radioactivity from fusion reactors under accident conditions are examined. The handling and disposal of radioactive waste are discussed; the status of biological effects of magnetic fields are referenced; and release mechanisms for tritium and activation products, including analytical methods, are presented. (orig./GG)

The PEC reactor. Safety analysis: Detailed reports

Energy Technology Data Exchange (ETDEWEB)

1988-01-01

In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.

MAPLE research reactor safety uncertainty assessment methodology

International Nuclear Information System (INIS)

Sills, H.E.; Duffey, R.B.; Andres, T.H.

1999-01-01

The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

Physics and safety of advanced research reactors

International Nuclear Information System (INIS)

Boening, K.; Hardt, P. von der

1987-01-01

Advanced research reactor concepts are presently being developed in order to meet the neutron-based research needs of the nineties. Among these research reactors, which are characterized by an average power density of 1-10 MW per liter, highest priority is now generally given to the 'beam tube reactors'. These provide very high values of the thermal neutron flux (10 14 -10 16 cm -2 s -1 ) in a large volume outside of the reactor core, which can be used for sample irradiations and, in particular, for neutron scattering experiments. The paper first discusses the 'inverse flux trap concept' and the main physical aspects of the design and optimization of beam tube reactors. After that two examples of advanced research reactor projects are described which may be considered as two opposite extremes with respect to the physical optimization principle just mentioned. The present situation concerning cross section libraries and neutronic computer codes is more or less satisfactory. The safety analyses of advanced research reactors can largely be updated from those of current new designs, partially taking advantage of the immense volume of work done for power reactors. The paper indicates a few areas where generic problems for advanced research reactor safety are to be solved. (orig.)

Safety features of the MAPLE-X10 reactor design

International Nuclear Information System (INIS)

Lee, A.G.; Bishop, W.E.; Heeds, W.

1990-01-01

This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

Safety Design Criteria of Indian Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

2013-01-01

• Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

Code on the safety of nuclear research reactors: Design

International Nuclear Information System (INIS)

1992-01-01

The main objective of this publication is to provide a safety basis for the design of a research reactor and for the assessment of the design. Another objective is to cover certain aspects related to regulatory supervision, siting and quality assurance, as far as these are related to activities for the design of a research reactor. These objectives are expressed in terms of requirements and recommendations for the design of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop specific regulations and safety criteria for its research reactor programme.

Integral fast reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

Integral Fast Reactor concept inherent safety features

International Nuclear Information System (INIS)

Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

1986-01-01

The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

International Nuclear Information System (INIS)

Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

2003-01-01

The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)

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'

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

Overview of fast reactor safety research and development in the USA

International Nuclear Information System (INIS)

Neuhold, R.J.; Avery, R.; Marchaterre, J.F.

1986-01-01

The liquid metal reactor (LMR) safety R and D program in the U.S. is presently focused on support of two modular innovative reactor concepts: PRISM - the General Electric Power Reactor Inherently Safe Module and SAFR - the Rockwell International Sodium Advanced Fast Reactor. These reactor plant concepts accommodate the use of either oxide fuel or the metal fuel which is under development in the Argonne National Laboratory (ANL) Integral Fast Reactor (IFR) program. Both concepts emphasize prevention of accidents through enhancement of inherent and passive safety characteristics. Enhancement of these characteristics is expected to be a major factor in establishing new and improved safety criteria and licensing arrangements with regulatory authorities for advanced reactors. Limited work is also continuing on the Large Scale Prototype Breeder (LSPB), a large pool plant design. Major elements of the current and restructured safety program are discussed. (author)

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.

Nuclear Reactor RA Safety Report, Format and Contents

International Nuclear Information System (INIS)

1986-11-01

This is a new complete version of the safety report of nuclear reactor RA is made according to the recommendations of the IAEA. Report includes all the relevant data needed for evaluation of safe operation of this nuclear facility. Each of seven volumes of this report cover separate topics as follows: (1) introduction; (2) Site characteristics; (3) description of the reactor building and installations; (4) description of the reactor; (5) description of the coolant system; (6) description of the regulation and safety instrumentation; (7) description of the power supply system; (8) description of the auxiliary systems; (9) radiation protection issues; (10) radioactive waste management (11) reactor operation; (12) accident analysis during previous operation; (13) analysis of possible accident causes; (14) safety analysis and preventive actions: (15) analysis of significant accidents; (16) analysis of maximum possible accident; (17) environmental impact analysis in case of accident [sr

The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

International Nuclear Information System (INIS)

Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

2006-06-01

Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

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.)

Proceedings of the international symposium on research reactor safety operations and modifications

International Nuclear Information System (INIS)

1990-03-01

The International Symposium on Research Reactor Safety, Operations and Modifications was organized by the International Atomic Energy Agency in cooperation with Atomic Energy of Canada Limited-Research Company. The main objectives of this Symposium were: (1) to exchange information and to discuss current perspectives and concerns relating to all aspects to research reactor safety, operations, and modifications; and, (2) to present views and to discuss future initiatives and directions for research reactor design, operations, utilization, and safety. The symposium topics included: research reactor programmes and experience; research reactor design safety and analysis; research reactor modifications and decommissioning; research reactor licensing; and new research reactors. These topics were covered during eight oral sessions and three poster sessions. These Proceedings include the full text of the 93 papers presented. The subject of Symposium was quite wide-ranging in that it covered essentially all aspects of research reactor safety, operations, and modifications. This was considered to be appropriate and timely given the 326 research reactors currently in operation in some 56 countries; given the degree of their utilization which ranges from pure and applied research to radioisotopes production to basic training and manpower development; and given that many of these reactors are undergoing extensive modifications, core conversions, power upratings, and are becoming the subject of safety reassessment and regulatory reviews. Although the Symposium covered many topics, the majority of papers and discussions tended to focus mainly on research reactor safety. This was seen as a clear sign of the continuing recognition of the fundamental importance of identifying and addressing, particularly through international cooperation, issues and concerns associated with research reactor safety

The safety characteristics of the HTR 500 reactor plant

International Nuclear Information System (INIS)

Wachholz, W.

1987-01-01

The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

Reactor safety research in Sweden

International Nuclear Information System (INIS)

Pershagen, B.

1980-02-01

Objectives, means and results of Swedish light water reactor safety research during the 1970s are reviewed. The expenditure is about 40 Million Swkr per year excluding industry. Large efforts have been devoted to experimental studies of loss of coolant accidents. Large scale containment response tests for simulated pipe breaks were carried out at the Marviken facility. At Studsvik a method for testing fuel during fast power changes has been developed. Stress corrosion, crack growth and the effect of irradiation on the strength ductility of Zircaloy tube was studied. A method for determining the fracture toughness of pressure vessel steel was developed and it was shown that the fracture toughness was lower than earlier assumed. The release of fission products to reactor water was studied and so was the release, transport and removal of airborne radioactive matter for Swedish BWRs and PWRs. Test methods for iodine filter systems were developed. A system for continuous monitoring of radioactive noble gas stack release was developed for the Ringhals plant. Attention was drawn to the importance of the human factor for reactor safety. Probabilistic methods for risk analysis were applied to the Barsebaeck 2 and Forsmark 3 boiling water reactors. Procedures and working conditions for operator personnel were investigated. (GBn)

Chernobyl and the safety of nuclear reactors in OECD countries

International Nuclear Information System (INIS)

1987-01-01

This report assesses the possible bearing of the Chernobyl accident on the safety of nuclear reactors in OECD countries. It discusses analyses of the accident performed in several countries as well as improvements to the safety of RBMK reactors announced by the USSR. Several remaining questions are identified. The report compares RBMK safety features with those of commercial reactors in OECD countries and evaluates a number of issues raised by the Chernobyl accident

Application of Code Of Conduct on the Safety of Research Reactor (RTP)

International Nuclear Information System (INIS)

Ligam, A.S.; Ahmad Nabil Abd Rahim; Zarina Masood

2014-01-01

The implementation and the practices of the effective safety system at research reactors are important to ensure that the worker, public and environment do not receive any abnormal causes. Many international safety related support agencies for research reactor such as International Atomic Energy Agency (IAEA) providing guidelines that can be applied to enhance and strengthen the enforcement of safety namely Code of Conduct on the Safety of Research Reactor (IAEA/CODEOC/RR/2006). The excellent safety management, reliability, and maintainability of RTP reactor structures, coupled with personnel numerous lessons and experiences learned, Reactor TRIGA PUSPATI research reactor providing Nuclear Malaysia personnel and visitor the very safe working and visiting environment. This paper will discuss the status, practices and improvement strategies over the past few years. (author)

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

The impact of WASH-1400 on reactor safety evaluation

International Nuclear Information System (INIS)

Tanguy, P.Y.

1976-01-01

Trends in reactor safety evaluation in France following the publication of WASH-1400 (the Rasmussen Report) are presented. What is called 'the meteorite case' is first schematically presented as follows: WASH-1400 shows nuclear risk equivalent to meteorite risk and reasonable corrections cannot make many orders of magnitude, consequently present safety rules are adequate. The very impact of WASH-1400 on safety approach is then discussed as for: assistance to deterministic safety analysis, introduction of probabilistic safety criteria, acceptable level of risk, and the use of results in research and reactor operating experience

Considerations on nuclear reactor passive safety systems

International Nuclear Information System (INIS)

2016-01-01

After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

Safety of NPP with WWER-440 and WWER-1000 reactors

Energy Technology Data Exchange (ETDEWEB)

Balabanov, E [Energoproekt, Sofia (Bulgaria); Gledachev, J; Angelov, D [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1996-12-31

The WWER-440 and WWER-1000 reactors used at the Kozloduy NPP have been analyzed in terms of safety. There are currently 4 reactors WWER-440/230 and 2 reactors WWER-1000/320. The former do not comply completely with the modern safety requirements due to the regulations acted in the sixties when they have been designed. The main features of these reactors are: low power density in the core; three levels of reactor control and protection; six primary loops; horizontal steam generators; two turbines; large number of cross-unit connections. The low thermal density in the core, the low specific thermal loading in the rods and the large coolant inventory enhance the safety, while the major deficiencies are identified as follows: insufficient capabilities for emergency core cooling; low diversification and physical separation of the safety systems; old fashioned control systems; inadequate fire protection; lack of full containment. It is pointed out that several design and operation actions have been completed in the Kozloduy NPP in order to enhance their safety. The WWER-1000 units are 320 model and feature a high safety level, complying completely with OPB-82 regulations and with all current international safety standards. 3 refs., 7 figs., 1 tab.

Safety of NPP with WWER-440 and WWER-1000 reactors

International Nuclear Information System (INIS)

Balabanov, E.; Gledachev, J.; Angelov, D.

1995-01-01

The WWER-440 and WWER-1000 reactors used at the Kozloduy NPP have been analyzed in terms of safety. There are currently 4 reactors WWER-440/230 and 2 reactors WWER-1000/320. The former do not comply completely with the modern safety requirements due to the regulations acted in the sixties when they have been designed. The main features of these reactors are: low power density in the core; three levels of reactor control and protection; six primary loops; horizontal steam generators; two turbines; large number of cross-unit connections. The low thermal density in the core, the low specific thermal loading in the rods and the large coolant inventory enhance the safety, while the major deficiencies are identified as follows: insufficient capabilities for emergency core cooling; low diversification and physical separation of the safety systems; old fashioned control systems; inadequate fire protection; lack of full containment. It is pointed out that several design and operation actions have been completed in the Kozloduy NPP in order to enhance their safety. The WWER-1000 units are 320 model and feature a high safety level, complying completely with OPB-82 regulations and with all current international safety standards. 3 refs., 7 figs., 1 tab

Trends in fusion reactor safety research

International Nuclear Information System (INIS)

Herring, J.S.; Holland, D.F.; Piet, S.J.

1991-01-01

Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g. for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions. 18 refs

Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

International Nuclear Information System (INIS)

1986-11-01

This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

Comments on nuclear reactor safety in Ontario

International Nuclear Information System (INIS)

1987-08-01

The Chalk River Technicians and Technologists Union representing 500 technical employees at the Chalk River Nuclear Laboratories of AECL submit comments on nuclear reactor safety to the Ontario Nuclear Safety Review. Issues identified by the Review Commissioner are addressed from the perspective of both a labour organization and experience in the nuclear R and D field. In general, Local 1568 believes Ontario's CANDU nuclear reactors are not only safe but also essential to the continued economic prosperity of the province

Operational safety and reactor life improvements of Kyoto University Reactor

International Nuclear Information System (INIS)

Utsuro, M.; Fujita, Y.; Nishihara, H.

1990-01-01

Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

Energy Technology Data Exchange (ETDEWEB)

Stosic, Zoran V. [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)], E-mail: Zoran.Stosic@areva.com; Brettschuh, Werner; Stoll, Uwe [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)

2008-08-15

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of

Aspects of nuclear reactor safety

International Nuclear Information System (INIS)

Hardt, P. von der; Rottger, H.

1980-01-01

The Colloquium on 'Irradiation Tests for Reactor Safety Programmes' has been organised by JRC Petten in order to determine the present state of technology in the field. The role of research and test reactors for studies of structural material and fuel elements under transient and off-normal conditions was to be explained. The Colloquium has been attended by 110 participants from outside and inside Europe. 27 papers were presented covering the major ongoing projects in Japan, the United States, and in Europe, and elaborating in particular: - design rationale and layout of safety irradiation experiments; - design, manufacture, and performance of irradiation equipment with particular attention to generation and control of transient conditions, fast response in-pile instrumentation and its out-of-pile data retrieval; - post-irradiation evaluation; - results and analytical support

High-temperature gas-cooled reactor safety-reliability program plan

Energy Technology Data Exchange (ETDEWEB)

1981-03-01

The purpose of this document is to present a safety plan as part of an overall program plan for the design and development of the High Temperature Gas-Cooled Reactor (HTGR). This plan is intended to establish a logical framework for identifying the technology necessary to demonstrate that the requisite degree of public risk safety can be achieved economically. This plan provides a coherent system safety approach together with goals and success criterion as part of a unifying strategy for licensing a lead reactor plant in the near term. It is intended to provide guidance to program participants involved in producing a technology base for the HTGR that is fully responsive to safety consideration in the design, evaluation, licensing, public acceptance, and economic optimization of reactor systems.

On application of CFD codes to problems of nuclear reactor safety

International Nuclear Information System (INIS)

Muehlbauer, Petr

2005-01-01

The 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in May 2002 at Aix-en-Province, France, recommended formation of writing groups to report the need of guidelines for use and assessment of CFD in single-phase nuclear reactor safety problems, and on recommended extensions to CFD codes to meet the needs of two-phase problems in nuclear reactor safety. This recommendations was supported also by Working Group on the Analysis and Management of Accidents and led to formation oaf three Writing Groups. The first writing Group prepared a summary of existing best practice guidelines for single phase CFD analysis and made a recommendation on the need for nuclear reactor safety specific guidelines. The second Writing Group selected those nuclear reactor safety applications for which understanding requires or is significantly enhanced by single-phase CFD analysis, and proposed a methodology for establishing assesment matrices relevant to nuclear reactor safety applications. The third writing group performed a classification of nuclear reactor safety problems where extension of CFD to two-phase flow may bring real benefit, a classification of different modeling approaches, and specification and analysis of needs in terms of physical and numerical assessments. This presentation provides a review of these activities with the most important conclusions and recommendations (Authors)

Considerations concerning the reliability of reactor safety equipment

International Nuclear Information System (INIS)

Furet, J.; Guyot, Ch.

1967-01-01

A review is made of the circumstances which favor a good collection of maintenance data at the C.E.A. The large amount of data to be treated has made necessary the use of a computer for analyzing automatically the results collected. Here, only particular aspects of the reliability from the point of view of the electronics used for nuclear reactor control will be dealt with: sale and unsafe failures; probability of survival (in the case of reactor safety); availability. The general diagrams of the safety assemblies which have been drawn up for two types of reactor (power reactor and low power experimental reactor) are given. Results are presented of reliability analysis which could be applied to the use of functional modular elements, developed industrially in France. Improvement of this reliability appears to be fairly limited by an increase in the redundancy; on the other hand it is shown how it may be very markedly improved by the use of automatic tests with different frequencies for detecting unsafe failures rates of measurements for the sub-assemblies and for the logic sub-assemblies. Finally examples are given to show the incidence of the complexity and of the use of different technologies in reactor safety equipment on the reliability. (authors) [fr

The European Pressurized Water Reactor (EPR). State of the art after the preliminary design phase

International Nuclear Information System (INIS)

Bouteille, F.; Schneider, D.

2002-01-01

The European Pressurized Water Reactor (EPR) is an evolutionary development of the pressurized water reactor product lines built by Framatome and Siemens in France and Germany. Under the technical leadership of both nuclear power plant suppliers (now merged in Framatome ANP, a joint venture of AREVA and Siemens) the future-oriented plant concepts was developed in close cooperation with German and French utilities and in compliance with the European Utility Requirements. The EPR has safety features with which even extremely improbable, beyond design-basis events can be controlled and their effects can be limited to such an extent that no emergency response actions need be taken outside of the immediate plant site. This also means that safety systems prevent containment failure even in the improbable case of a core melt. This was confirmed by the French and German reactor safety authorities. The selected high thermal output also insures the economic viability of the innovative reactor concept, so that the power generation costs which can be achieved with the EPR will be absolutely competitive with those of fossil energy carriers. Framatome ANP has thus developed a pressurized water reactor ready for offer at the right time, which can completely fulfill the most rigorous requirements in terms of nuclear safety and economy. (Author)

Experts' discussion on reactor safety research

International Nuclear Information System (INIS)

1980-01-01

The experts' discussion on reactor safety research deals with risk analysis, political realization, man and technics, as well as with the international state of affairs. Inspite of a controversy on individual issues and on the proportion of governmental and industrial involvment in reactor safety research, the continuation and intensification of corresponding research work is said to be necessary. Several participants demanded to consider possible 'conventional accidents' as well as a stronger financial commitment by the industry in this sector. The ratio 'man and technics' being an interface decisive for the proper functioning or failure of complex technical systems requires even more research work to be done. (GL) [de

Reactor Safety Research: Semiannual report, July-December 1986

Energy Technology Data Exchange (ETDEWEB)

1987-11-01

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 the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and 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 tehcnology base supporting licensing decisions.

Reactor Safety Research: Semiannual report, July-December 1986

International Nuclear Information System (INIS)

1987-11-01

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 the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and 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 tehcnology base supporting licensing decisions

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

Introduction to German and Russian terminology in the field of reactor safety, taking the high-temperature reactor as a specific example. Einfuehrung in die deutsche und die russische Terminologie der Reaktorsicherheit am Beispiel des Hochtemperaturreaktors

Energy Technology Data Exchange (ETDEWEB)

Patt, B

1987-01-01

The need for international cooperation in the field of reactor safety engineering has been accepted by the two countries. The diploma thesis in hand is intended as a glossary and guide for translators who start to familiarize themselves with the subject field and the terminology. The glossary sets out the terminology applied for describing the safety features and the safety systems of the reactor itself and does not cover the power plant as a whole. The thesis presents the terminology in its textual environment, and in an alphabetical word list.

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

Recent achievement within the French-German safety approach for future PWRs

International Nuclear Information System (INIS)

Gros, G.; Rollinger, F.; Frisch, W.; Simon, M.

1999-12-01

The development of the common French-German safety approach was accomplished on three working levels: the technical safety organisations GRS and IPSN provided the technical basis, the advisory groups GPR and RSK developed common recommendations, and the authorities BMU and DSIN adopted and issued the recommendations. The general safety approach issued in May 1993 contains safety objectives, general principles and some technical principles for future PWRs. Based on this general approach, more detailed recommendations have been developed in 1994 on key issues. The following period from 1995 on was characterised by a further refinement of the recommendations and the treatment of some new subjects such as digital I and C, man-machine-interface and core design. (authors)

Food for thought: change and continuity in German food safety policy

NARCIS (Netherlands)

Paul, K.T.

2007-01-01

This paper examines German food safety policy from a discourse‐theoretically informed perspective. It draws on qualitative research, including textual analysis and in‐depth interviews. Its aim is to understand why the occurrence of ‘mad‐cow‐disease’ (BSE) in Germany in the year 2000 led to what some

Design and safety aspects of nuclear district heating reactors

International Nuclear Information System (INIS)

Brogli, R.; Mathews, D.; Pelloni, S.

1989-01-01

Extensive studies on the rationale, the potential and the technology of nuclear district heating have been performed in Switzerland. Beside economics the safety aspects were of primary importance. Due to the high costs to transport heat the heating reactor tend to be small and therefore, minimally staffed and located close to population centers. Stringed safety rules are therefore applying. Gas cooled reactors are well suited as district heating reactors since they have due to their characteristics several inherent features, significant safety margins and a remarkable radioactivity retention potential. Some ways to mitigate the effects of water ingress and graphite corrosion are under investigation. (author). 5 refs, 3 figs

LWR safety studies. Analyses and further assessments relating to the German Risk Assessment Study on Nuclear Power Plants. Vol. 3

International Nuclear Information System (INIS)

1983-01-01

Critical review of the analyses of the German Risk Assessment Study on Nuclear Power Plants (DRS) concerning the reliability of the containment under accident conditions and the conditions of fission product release (transport and distribution in the environment). Main point of interest in this context is an explosion in the steam section and its impact on the containment. Critical comments are given on the models used in the DRS for determining the accident consequences. The analyses made deal with the mathematical models and database for propagation calculations, the methods of dose computation and assessment of health hazards, and the modelling of protective and safety measures. Social impacts of reactor accidents are also considered. (RF) [de

Safety studies concerning nuclear power reactors

International Nuclear Information System (INIS)

Bailly, Jean; Pelce, Jacques

1980-01-01

The safety of nuclear installations poses different technical problems, whether concerning pressurized water reactors or fast reactors. But investigating methods are closely related and concern, on the one hand, the behavior of shields placed between fuel and outside and, on the other, analysis of accidents. The article is therefore in two parts based on the same plan. Concerning light water reactors, the programme of studies undertaken in France accounts for the research carried out in countries where collaboration agreements exist. Concerning fast reactors, France has the initiative of their studies owing to her technical advance, which explains the great importance of the programmes under way [fr

Passive safety features in current and future water cooled reactors

International Nuclear Information System (INIS)

1990-11-01

Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

Some safety considerations in laser-controlled thermonuclear reactors. Final report

International Nuclear Information System (INIS)

Botts, T.E.; Breton, D.; Chan, C.K.; Levy, S.I.; Sehnert, M.; Ullman, A.Z.

1978-07-01

A major objective of this study was to identify potential safety questions for laser controlled thermonuclear reactors. From the safety viewpoint, it does not appear that the actual laser controlled thermonuclear reactor conceptual designs present hazards very different than those of magnetically confined fusion reactors. Some aspects seem beneficial, such as small lithium inventories, and the absence of cryogenic devices, while other aspects are new, for example the explosion of pressure vessels and laser hazards themselves. Major aspects considered in this report include: (a) general safety considerations, (b) tritium inventories, (c) system behavior during loss of flow accidents, and (d) safety considerations of laser related penetrations

Advanced power reactors with improved safety characteristics

International Nuclear Information System (INIS)

Birkhofer, A.

1994-01-01

The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

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).

Reactor safety after TMI (Three-Mile Island)

International Nuclear Information System (INIS)

Smidt, D.

1980-01-01

The present status of safety design in Germany and the governing principles behind it, are presented. The problem of human failure and what could be done about it, are discussed. This is followed by the description of some research projects for further increasing the accident prevention capability of German plants. (E.G.) [pt

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

New training reactor at Dresden Technical University

International Nuclear Information System (INIS)

Hansen, W.; Knorr, J.; Wolf, T.

2006-01-01

A total of 14 low-power (up to 10 W) training reactors have been operated at German universities, 9 of them officially classified as being operational in 2004, though for very different uses. This number is expected to drop sharply. The only comprehensive upgrading of a training reactor took place at Dresden Technical University: AKR-2, the most modern facility in Germany, started routine operation in April 2005, under a newly granted license pursuant to Sec. 7, Subsec. 1 of the German Atomic Energy Act, for training students in nuclear technology, for suitable research projects, and a a center of information about reactor technology and nuclear technology for the interested public. One special aspect of this refurbishment was the installation of digital safety I and C systems of the TELEPERM XS line, which are used also in other modern plants. This fact, plus the easy possibility to use the plant for many basic experiments in reactor physics and radiation protection, make the AKR-2 attractive also to other users (e.g. for training reactor personnel or other persons working in nuclear technology). (orig.)

The European pressurized water reactor. Result of the French-German cooperation of experienced NPP suppliers and operators

International Nuclear Information System (INIS)

Fischer, U.

1999-01-01

In 1989 Framatome and Siemens, the two most experienced European nuclear power plant suppliers, decided to join the efforts for the development of a new reactor type for the next generation in their equally owned subsidiary nuclear power international (NPI). In 1992 Electricite de France and the major German utilities operating nuclear power plants merged their own development programs with that of nuclear power international and initiated the European pressurized water reactor (EPR) project. In order to reach the two major targets of the project, the licensability in both countries, France and Germany, and the competitiveness of nuclear energy with other alternative energy sources, the design basis which had differently developed in the two countries needed to be harmonized. In parallel, the licensing authorities of both countries extended their existing cooperation in the field of a safety survey of existing nuclear power plants to the definition of safety criteria for the next generation of nuclear power plants. Through this cooperation the licensability of EPR in France and Germany will be assured. Continuously performed cost analysis show in addition that also the second target of the project, the competitiveness with alternative primary energy sources, can be achieved. Thanks to the fruitful cooperation between all parties involved, satisfactory results have been achieved not by a simple superposition of existing design features but through a careful evaluation and combination of the best available alternatives. At the end of 1997 the basic design results were compiled in a final report. Subsequently an optimization phase was launched that further improves the competitiveness of the power generation costs. (orig.)

Some views on nuclear reactor safety

Energy Technology Data Exchange (ETDEWEB)

Tanguy, P.Y. [Electricite de France, Paris (France)

1995-04-01

This document is the text of a speech given by Pierre Y. Tanguy (Electricite de France) at the 22nd Water Reactor Safety Meeting held in Bethesda, MD in 1994. He describes the EDF nuclear program in broad terms and proceeds to discuss operational safety results with EDF plants. The speaker also outlines actions to enhance safety planned for the future, and he briefly mentions French cooperation with the Chinese on the Daya Bay project.

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Problems in the assessment of inherent safety characteristics of nuclear reactors

International Nuclear Information System (INIS)

Garribba, S.F.; Vivante, C.

1988-01-01

A number of proposals are being made for an increased RD and D effort on advanced nuclear power reactors that would display outstanding safety performance. A common characteristic of the different reactor concepts would be their limited reliance upon active engineered systems under major accident conditions. However, when submitted to a more close scrutiny reactor concept options may reveal diverging safety behaviors and also development opportunities. In this respect, three issues are explored in this paper. A first question is the meaning of non-active, i.e. inherent and passive safety features. Next, is the ranking of advanced and new reactor concepts from the viewpoint of inherent and passive safety. Multiple correspondence analysis may provide a simple tool, whose use is shown for the case of HTR-500, AP600 and PRISM. Conversely, probabilistic risk assessment would allow quantitative comparisons, although lack of information and data is an obstacle. Finally, is demonstration of safety performances as a step toward market deployment of the new reactor systems

Safety of evolutionary and innovative nuclear reactors: IAEA activities and world efforts

International Nuclear Information System (INIS)

Saito, T.; Gasparini, M.

2004-01-01

'Defence in Depth' approach constitutes the basis of the IAEA safety standards for nuclear power plants. Lessons learned from the current generation of reactors suggest that, for the next generation of reactor designs, the Defence in Depth philosophy should be retained, and that its implementation should be guided by the probabilistic insights. Recent developments in the area of general safety requirements based on Defence in Depth approach are examined and summarized. Global efforts to harmonize safety requirements for evolutionary nuclear power plants have involved many countries and organizations such as IAEA, US EPRI and European Utility EUR Organization. In recent years, developments of innovative nuclear power plants are also being discussed. The IAEA is currently developing a safety approach specifically for innovative nuclear reactors. This approach will eventually lead to a proposal of safety requirements for innovative reactors. Such activities related to safety requirements of evolutionary and innovative reactors are introduced. Various evolutionary and innovative reactor designs are reported in the world. The safety design features of evolutionary large LWRs, innovative LWRs, Modular High Temperature Gas Reactors and Small Liquid Metal Cooled LMRs are also introduced. Enhanced safety features proposed in such reactors are discussed and summarized according to the levels of Defence in Depth. For future nuclear plants, international cooperation and harmonization, especially in the area of safety, appear to be inevitable. Based on the past experience with many member states, the IAEA believes itself to be the uniquely positioned international organization to play this key role. (authors)

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)

Nuclear safety research project (PSF). 1999 annual report

International Nuclear Information System (INIS)

Muehl, B.

2000-08-01

The reactor safety R and D work of the Karlsruhe Research Centre (FZK) has been part of the Nuclear Safety Research Project (PSF) since 1990. The present annual report summarizes the R and D results of PSF during 1999. The research tasks cover three main topics: Light Water Reactor safety, innovative systems, and studies related to the transmutation of actinides. The importance of the Light Water Reactor safety, however, has decreased during the last year in favour of the transmutation of actinides. Numerous institutes of the research centre contribute to the PSF programme, as well as several external partners. The tasks are coordinated in agreement with internal and external working groups. The contributions to this report, which are either written in German or in English, correspond to the status of early/mid 2000. (orig.) [de

Nuclear reactor safety program in US department of energy and future perspectives

International Nuclear Information System (INIS)

Song, Y.T.

1988-01-01

The US Department of Energy (DOE) establishes policy, issues orders, and assures compliance with requirements. The contractors who design, construct, modify, operate, maintain and decommission DOE reactors, set forth the assessment of the safety of cognizant reactors and implement DOE orders. Teams of experts in the Department, through scheduled and unscheduled review programs, reassess the safety of reactors in every phases of their lives. As new technology develops, the safety programs are reevaluated and policies are modified to accommodate these new technologies. The diagnostic capabilities of the computer using multiple alarms to enhance detection of defects and control of a reactor have been greatly utilized in reactor operating systems. The Application of artificial intelligence technologies for diagnostic and even for the decision making process in the event of reactor accidents would be one of the future trends in reactor safety programs

Safety concept of nuclear power plants in the Federal Republic of Germany

International Nuclear Information System (INIS)

Schnurer, H.L.; Seipel, H.G.

1983-01-01

The safety philosophy of the Federal Republic of Germany (FRG) pertaining to nuclear power plants is presented. Its general approach makes much use of research and experience in other countries, especially the United States, but the FRG has also evolved approaches and aspects unique to itself. The article discusses the institutional interrelationships of the entities affecting German nuclear power plant safety and presents the resulting German safety philosophy. This philosophy is based, as in the United States, on multilevel defenses but with greater reliance on a number of accident-prevention and/or -mitigation concepts, including: (1) automated responses to off-normal conditions, (2) an operating-limit system that acts before initiation of protective actions by the reactor protection system, (3) multistrand independent safety trains, and (4) long-term self-contained safety under accident conditions without operator intervention. The practical realization of this philosophy in the design and operation of the chief reactor systems is described

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

Analysis of dynamic stability and safety of the reactor system by reactor simulator

International Nuclear Information System (INIS)

Raisic, N.

1963-11-01

This document defines the approximations done for establishing a mathematical model of a reactor. Since the model should be used for safety analysis, it was important to choose a mathematical model less stable than the reactor itself. The analysis was performed on the analog computer RAS. Results obtained and conclusions concerned with three possible reactor accidents are presented [sr

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

Development of the reactor safety film

International Nuclear Information System (INIS)

Sheheen, N.N.; Hodson, P.J.

1981-01-01

The first computer-generated film of LASL's Reactor Safety efforts was developed using the ANIMATE framework, a program that adds visual capabilities to MAPPER. Numerous software limitations had to be overcome within a very limited production schedule. A significant achievement was the 15,000-vector-per-frame sequence depicting a pressurized water reactor core with parts flashing while pumps circulate fluid through the system

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.)

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)

A bibliography of AECL publications on reactor safety

International Nuclear Information System (INIS)

Hawley, N.J.

1979-12-01

AECL Publications on Reactor Safety in CANDU Reactors are listed in this bibliography. The listing is chronological and the accompanying index is by subject. The bibliography will be brought up to date annually. (auth)

Overview of the reactor safety study consequence model

International Nuclear Information System (INIS)

Wall, I.B.; Yaniv, S.S.; Blond, R.M.; McGrath, P.E.; Church, H.W.; Wayland, J.R.

1977-01-01

The Reactor Safety Study (WASH-1400) is a comprehensive assessment of the potential risk to the public from accidents in light water power reactors. The engineering analysis of the plants is described in detail in the Reactor Safety Study: it provides an estimate of the probability versus magnitude of the release of radioactive material. The consequence model, which is the subject of this paper, describes the progression of the postulated accident after the release of the radioactive material from the containment. A brief discussion of the manner in which the consequence calculations are performed is presented. The emphasis in the description is on the models and data that differ significantly from those previously used for these types of assessments. The results of the risk calculations for 100 light water power reactors are summarized

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

Development of 3D CFD simulation method in nuclear reactor safety analysis

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

Safety design study of fast breeder reactors in Japan

International Nuclear Information System (INIS)

Miura, M.; Inagaki, T.

1992-01-01

This paper reports on two fast breeder reactor (FBR) concepts, the tank type and the loop type, that have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle fo the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant

Nuclear technology and reactor safety engineering. The situation ten years after the Chernobyl reactor accident

International Nuclear Information System (INIS)

Birkhofer, A.

1996-01-01

Ten years ago, on April 26, 1986 the most serious accident ever in the history of nuclear tgechnology worldwide happened in unit 4 of the nuclear power plant in Chernobyl in the Ukraine, this accident unveiling to the world at large that the Soviet reactor design lines are bearing unthought of safety engineering deficits. The dimensions of this reactor accident on site, and the radioactive fallout spreading far and wide to many countries in Europe, vividly nourished the concern of great parts of the population in the Western world about the safety of nuclear technology, and re-instigated debates about the risks involved and their justification. Now that ten years have elapsed since the accident, it is appropriate to strike a balance and analyse the situation today. The number of nuclear power plants operating worldwide has been growing in the last few years and this trend will continue, primarily due to developments in Asia. The Chernobyl reactor accident has pushed the international dimension of reactor safety to the foreground. Thus the Western world had reason enough to commit itself to enhancing the engineered safety of reactors in East Europe. The article analyses some of the major developments and activities to date and shows future perspectives. (orig.) [de

Extension and update of the reliability data for fire protection equipment in German light water reactors; Ergaenzung und Aktualisierung von Zuverlaessigkeitskenngroessen fuer Brandschutzeinrichtungen in deutschen Leichtwasserreaktoren

Energy Technology Data Exchange (ETDEWEB)

Forell, Burkhard; Einarsson, Svante

2016-12-15

Within the BMUB project 3610R01370 ''Extension and update of reliability data for fire protection equipment in German light water reactors'' plant-specific as well as generic failure rates for the technical reliability of active fire protection features in German nuclear power plants have been calculated. Based on results of previous projects, in this project observation times of components were updated and extended and additional components and functions were assessed. Now, the data evaluated results from a total of six German reference plants with seven reactor units.

International nuclear safety experts complete IAEA peer review of German regulatory system

International Nuclear Information System (INIS)

2008-01-01

policies, the regulatory framework and activities identified by the IRRS team were: Both BMU and UM BW have high quality and experienced staff, which are supported by experienced and competent technical support organizations. Both organizations have a high commitment to learning and self improvement; The team has also made recommendations and suggestions related to areas where the regulatory system as a whole could be improved. Examples include: There is room for improving information exchange and communication between the Federal and State regulatory bodies. Such enhanced communication would also lead to increased mutual trust and public confidence. In order to fulfill their regulatory responsibilities, BMU and UM BW should be staffed appropriately and develop a plan for a succession planning strategy. It could be very beneficial for the regulatory system, and helpful for worldwide learning, if all German 'Laender' (states) with nuclear power plants also took the opportunity to benefit from the experience of an IRRS mission, including making self- assessments against international best practice. The peer review consisted of an analysis of technical regulatory information, interviews, and discussions with key personnel at BMU and UM BW, as well as with regulatory inspectors at Neckarwestheim-1 Nuclear Power Plant to witness a regulatory inspection. Regulatory aspects were also discussed with Federal Minister Sigmar Gabriel and Minister Tanja Goenner of Baden-Wuerttemberg. Other organisations such as the Federal Office for Radiation Protection (BFS), the Reactor Safety Commission (RSK), the technical support organisation Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS), and the Nuclear Safety Standards Commission (ILK) were also involved. (author)

Design of Safety Parameter Monitoring Function in a Research Reactor Facility

Energy Technology Data Exchange (ETDEWEB)

Park, Jaekwan; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

Design of Safety Parameter Monitoring Function in a Research Reactor Facility

International Nuclear Information System (INIS)

Park, Jaekwan; Suh, Yongsuk

2014-01-01

The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

Proceedings of the Third Scientific Presentation on Reactor Safety Technology

International Nuclear Information System (INIS)

1998-01-01

These proceedings contains the results of research and development on reactor safety technology which carried out by Reactor Safety Technology Centre, National Atomic Energy Agency, Serpong, Indonesia during 1997/1998 fiscal year. The presentation was held on 13-14 May 1998 at Serpong,Indonesia

The safety of Ontario's nuclear reactors

International Nuclear Information System (INIS)

1980-06-01

A Select Committee of the Legislature of Ontario was established to examine the affairs of Ontario Hydro, the provincial electrical utility. Extensive public hearings were held on several topics including the safety of nuclear power reactors operating in Ontario. The Committee found that these reactors are acceptably safe. Many of the 24 recommendations in this report deal with the licensing process and public access to information. (O.T.)

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)

Advanced reactor concepts and safety

International Nuclear Information System (INIS)

Lipsett, J.J.

1988-06-01

The need for some consistency in the terms used to describe the evolution of methods for ensuring the safety of nuclear reactors has been identified by the IAEA. This is timely since there appears to be a danger that the precision of many valuable words is being diluted and that a new jargon may appear that will confuse rather than aid the communication of important but possibly diverse philosophies and concepts. Among the difficulties faced by the nuclear industry is promoting and gaining a widespread understanding of the risks actually posed by nuclear reactors. In view of the importance of communication to both the public and to the technical community generally, the starting point for the definition of terms must be with dictionary meanings and common technical usage. The nuclear engineering community should use such words in conformance with the whole technical world. This paper addresses many of the issues suggested in the invitation to meet and also poses some additional issues for consideration. Some examples are the role of the operator in either enhancing or degrading safety and how the meaning or interpretation of the word 'safety' can be expected to change during the next few decades. It is advantageous to use criteria against which technologies and ongoing operating performance can be judged provided that the criteria are generic and not specific to particular reactor concepts. Some thoughts are offered on the need to frame the criteria carefully so that innovative solutions and concepts are fostered, not stifled

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)

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

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

Software reliability and safety in nuclear reactor protection systems

Energy Technology Data Exchange (ETDEWEB)

Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

1993-11-01

Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

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)

Software reliability and safety in nuclear reactor protection systems

International Nuclear Information System (INIS)

Lawrence, J.D.

1993-11-01

Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor

Cost benefit analysis of reactor safety systems

International Nuclear Information System (INIS)

Maurer, H.A.

1984-01-01

Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)

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

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

Operating experience feedback from safety significant events at research reactors

Energy Technology Data Exchange (ETDEWEB)

Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

2015-05-15

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Operating experience feedback from safety significant events at research reactors

International Nuclear Information System (INIS)

Shokr, A.M.

2015-01-01

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

The safety of future reactors

International Nuclear Information System (INIS)

Tanguy, P.

1992-01-01

To sum up, I would like to underline once again the importance of experience feedback. This issue can only be properly handled by reversing the thought process which lay behind the construction of the current NPP's. The design was the springboard for building the reactors and then operating them. Throughout construction and at times during operation, many difficulties arose, which were overcome by modifications. The need today is to go back down the line in the opposite direction : to use operational and constructional experience to restructure the design. Furthermore, the design of future reactors appears to me as a process which must be founded upon two guiding principles : defense in depth and a PSA-type probabilistic approach. They seem to me ideally fitted to underpin such a process, especially in the case of an evolutionary-type reactor project. Such a strategy requires the cooperation of many participants supported by a high level of safety culture, as defined in the report published by the IAEA in 1991 : a permanent questioning attitude, a prudent approach and efficient communication between all of the individuals and organizations involved. Failure to make such an effort might well compromise the safety goals mentioned earlier in this paper. (author) any other organization. (author)

Evaluation of the Community's nuclear reactor safety research programme

International Nuclear Information System (INIS)

Brandstetter, A.; Goedkoop, J.A.; Jaumotte, A.; Malhouitre, G.; Tomkins, B.; Zorzoli, G.B.

1986-01-01

This report describes an evaluation of the 1980-85 CEC reactor safety programme prepared, at the invitation of the Commission, by a panel of six independent experts by means of examining the relevant document and by holding hearings with the responsible CEC staff. It contains the recommendations made by the panel on the following topics: the need for the JRC to continue to make its competence in the reactor safety field available to the Community; the importance of continuity in the JRC and shared-cost action programmes; the difficulty of developing reactor safety research programmes which satisfy the needs of users with diverse needs; the monitoring of the utilization of the research results; the maintenance of the JRC computer codes used by the Member States; the spin-off from research results being made available to other industrial sectors; the continued contact between the JRC researchers and the national experts; the coordination of LWR safety research with that of the Member States; and, the JRC work on fast breeders to be planned with regard to the R and D programmes of the Fast Reactor European Consortium

Current state of research on pressurized water reactor safety

International Nuclear Information System (INIS)

Couturier, Jean; Schwarz, Michel; Roubaud, Sebastien; Lavarenne, Caroline; Mattei, Jean-Marie; Rigollet, Laurence; Scotti, Oona; Clement, Christophe; Lancieri, Maria; Gelis, Celine; Jacquemain, Didier; Bentaib, Ahmed; Nahas, Georges; Tarallo, Francois; Guilhem, Gilbert; Cattiaux, Gerard; Durville, Benoit; Mun, Christian; Delaval, Christine; Sollier, Thierry; Stelmaszyk, Jean-Marc; Jeffroy, Francois; Dechy, Nicolas; Chanton, Olivier; Tasset, Daniel; Pichancourt, Isabelle; Barre, Francois; Bruna, Gianni; Evrard, Jean-Michel; Gonzalez, Richard; Loiseau, Olivier; Queniart, Daniel; Vola, Didier; Goue, Georges; Lefevre, Odile

2018-03-01

For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study - loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. -, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today's reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

Science.gov (United States)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion