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Sample records for seismic safety guide

  1. Seismic Safety Guide

    International Nuclear Information System (INIS)

    Eagling, D.G.

    1985-01-01

    The Seismic Safety Guide provides facilities managers with practical guidelines for administering a comprehensive earthquake safety program. Most facilities managers, unfamiliar with earthquake engineering, tend to look for answers in techniques more sophisticated than required to solve the actual problems in earthquake safety. Often the approach to solutions to these problems is so academic, legalistic, and financially overwhelming that mitigation of actual seismic hazards simply does not get done in a timely, cost-effective way. The objective of the Guide is to provide practical advice about earthquake safety so that managers and engineers can get the job done without falling into common pitfalls, prolonged diagnosis, and unnecessary costs. It is comprehensive with respect to earthquakes in that it covers the most important aspects of natural hazards, site planning, rehabilitation of existing buildings, design of new facilities, operational safety, emergency planning, non-structural elements, life lines, and risk management. 5 references

  2. Seismic Safety Guide

    International Nuclear Information System (INIS)

    Eagling, D.G.

    1983-09-01

    This guide provides managers with practical guidelines for administering a comprehensive earthquake safety program. The Guide is comprehensive with respect to earthquakes in that it covers the most important aspects of natural hazards, site planning, evaluation and rehabilitation of existing buildings, design of new facilities, operational safety, emergency planning, special considerations related to shielding blocks, non-structural elements, lifelines, fire protection and emergency facilities. Management of risk and liabilities is also covered. Nuclear facilities per se are not dealt with specifically. The principles covered also apply generally to nuclear facilities but the design and construction of such structures are subject to special regulations and legal controls

  3. Seismic Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    Eagling, D.G. (ed.)

    1983-09-01

    This guide provides managers with practical guidelines for administering a comprehensive earthquake safety program. The Guide is comprehensive with respect to earthquakes in that it covers the most important aspects of natural hazards, site planning, evaluation and rehabilitation of existing buildings, design of new facilities, operational safety, emergency planning, special considerations related to shielding blocks, non-structural elements, lifelines, fire protection and emergency facilities. Management of risk and liabilities is also covered. Nuclear facilities per se are not dealt with specifically. The principles covered also apply generally to nuclear facilities but the design and construction of such structures are subject to special regulations and legal controls.

  4. Safety design guides for seismic requirements for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for seismic requirements for CANDU 9 describes the seismic design philosophy, defines the applicable earthquakes and identifies the structures and systems requiring seismic qualification to ensure that the essential safety function can be adequately satisfied following earthquake. The detailed requirements for structures, systems and components which must be seismically qualified are specified in the Appendix. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 1 fig., (Author) .new

  5. Evaluation of seismic hazards for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    The main objective of this Safety Guide is to provide recommendations on how to determine the ground motion hazards for a plant at a particular site and the potential for surface faulting, which could affect the feasibility of construction and safe operation of a plant at that site. The guidelines and procedures presented in this Safety Guide can appropriately be used in evaluations of site suitability and seismic hazards for nuclear power plants in any seismotectonic environment. The probabilistic seismic hazard analysis recommended in this Safety Guide also addresses the needs for seismic hazard analysis of external event PSAs conducted for nuclear power plants. Many of the methods and processes described may also be applicable to nuclear facilities other than power plants. Other phenomena of permanent ground displacement (liquefaction, slope instability, subsidence and collapse) as well as the topic of seismically induced flooding are treated in Safety Guides relating to foundation safety and coastal flooding. Recommendations of a general nature are given in Section 2. Section 3 discusses the acquisition of a database containing the information needed to evaluate and address all hazards associated with earthquakes. Section 4 covers the use of this database for construction of a seismotectonic model. Sections 5 and 6 review ground motion hazards and evaluations of the potential for surface faulting, respectively. Section 7 addresses quality assurance in the evaluation of seismic hazards for nuclear power plants

  6. Seismic Hazards in Site Evaluation for Nuclear Installations. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-08-15

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear installations. It supplements the Safety Requirements publication on Site Evaluation for Nuclear Installations. The present publication provides guidance and recommends procedures for the evaluation of seismic hazards for nuclear power plants and other nuclear installations. It supersedes Evaluation of Seismic Hazards for Nuclear Power Plants, IAEA Safety Standards Series No. NS-G-3.3 (2002). In this publication, the following was taken into account: the need for seismic hazard curves and ground motion spectra for the probabilistic safety assessment of external events for new and existing nuclear installations; feedback of information from IAEA reviews of seismic safety studies for nuclear installations performed over the previous decade; collective knowledge gained from recent significant earthquakes; and new approaches in methods of analysis, particularly in the areas of probabilistic seismic hazard analysis and strong motion simulation. In the evaluation of a site for a nuclear installation, engineering solutions will generally be available to mitigate, by means of certain design features, the potential vibratory effects of earthquakes. However, such solutions cannot always be demonstrated to be adequate for mitigating the effects of phenomena of significant permanent ground displacement such as surface faulting, subsidence, ground collapse or fault creep. The objective of this Safety Guide is to provide recommendations and guidance on evaluating seismic hazards at a nuclear installation site and, in particular, on how to determine: (a) the vibratory ground motion hazards, in order to establish the design basis ground motions and other relevant parameters for both new and existing nuclear installations; and (b) the potential for fault displacement and the rate of fault displacement that could affect the feasibility of the site or the safe operation of the installation at

  7. Pickering seismic safety margin

    International Nuclear Information System (INIS)

    Ghobarah, A.; Heidebrecht, A.C.; Tso, W.K.

    1992-06-01

    A study was conducted to recommend a methodology for the seismic safety margin review of existing Canadian CANDU nuclear generating stations such as Pickering A. The purpose of the seismic safety margin review is to determine whether the nuclear plant has sufficient seismic safety margin over its design basis to assure plant safety. In this review process, it is possible to identify the weak links which might limit the seismic performance of critical structures, systems and components. The proposed methodology is a modification the EPRI (Electric Power Research Institute) approach. The methodology includes: the characterization of the site margin earthquake, the definition of the performance criteria for the elements of a success path, and the determination of the seismic withstand capacity. It is proposed that the margin earthquake be established on the basis of using historical records and the regional seismo-tectonic and site specific evaluations. The ability of the components and systems to withstand the margin earthquake is determined by database comparisons, inspection, analysis or testing. An implementation plan for the application of the methodology to the Pickering A NGS is prepared

  8. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    E.N. Lindner

    2004-12-03

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

  9. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    International Nuclear Information System (INIS)

    E.N. Lindner

    2004-01-01

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

  10. Seismic safety of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Katona, T.

    1993-01-01

    An extensive program is underway at Paks NPP for evaluation of the seismic safety and for development of the necessary safety increasing measures. This program includes the following five measures: investigation of methods, regulations and techniques utilized for reassessment of seismic safety of operating NPPs and promoting safety; investigation of earthquake hazards; development of concepts for creating the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept, and preliminary evaluation of the seismic safety

  11. Nuclear criticality safety guide

    International Nuclear Information System (INIS)

    Ro, Seong Ki; Shin, Hee Seong; Park, Seong Won; Shin, Young Joon.

    1997-06-01

    Nuclear criticality safety guide was described for handling, transportation and storage of nuclear fissile materials in this report. The major part of the report was excerpted frp, TID-7016(revision 2) and nuclear criticality safety written by Knief. (author). 16 tabs., 44 figs., 5 refs

  12. Hydrogen Technologies Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    Rivkin, C. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Burgess, R. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Buttner, W. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2015-01-01

    The purpose of this guide is to provide basic background information on hydrogen technologies. It is intended to provide project developers, code officials, and other interested parties the background information to be able to put hydrogen safety in context. For example, code officials reviewing permit applications for hydrogen projects will get an understanding of the industrial history of hydrogen, basic safety concerns, and safety requirements.

  13. Technical evaluation of seismic qualification of safety-related equipment

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yang Hui; Park, Heong Gee; Park, Yeong Seok [Univ. of Incheon, Incheon (Korea, Republic of)

    1994-04-15

    This study is purposed to evaluate the technical acceptability of the procedures and techniques of seismic qualifications which were performed for the YGN 3 and 4 safety-related equipment.This study is also targeted to suggest a systematized technical procedure guide for the effective performance and review of the seismic qualification, which reflects the most up-to-date licensing requirements and state-of the-art.

  14. Probabilistic safety assessment for seismic events

    International Nuclear Information System (INIS)

    1993-10-01

    This Technical Document on Probabilistic Safety Assessment for Seismic Events is mainly associated with the Safety Practice on Treatment of External Hazards in PSA and discusses in detail one specific external hazard, i.e. earthquakes

  15. Seismic and tsunami safety margin assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

  16. Seismic and tsunami safety margin assessment

    International Nuclear Information System (INIS)

    2013-01-01

    Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

  17. Safety Information System Guide

    International Nuclear Information System (INIS)

    Bullock, M.G.

    1977-03-01

    This Guide provides guidelines for the design and evaluation of a working safety information system. For the relatively few safety professionals who have already adopted computer-based programs, this Guide may aid them in the evaluation of their present system. To those who intend to develop an information system, it will, hopefully, inspire new thinking and encourage steps towards systems safety management. For the line manager who is working where the action is, this Guide may provide insight on the importance of accident facts as a tool for moving ideas up the communication ladder where they will be heard and acted upon; where what he has to say will influence beneficial changes among those who plan and control his operations. In the design of a safety information system, it is suggested that the safety manager make friends with a computer expert or someone on the management team who has some feeling for, and understanding of, the art of information storage and retrieval as a new and better means for communication

  18. Seismic safety of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Katona, T.

    1995-01-01

    This paper contains an overview of the results concerning the following activities: investigation of methods, regulations and techniques for reassessment of seismic safety of operating NPPs and upgrading of safety; investigation of earthquake hazards; development of concept for creation of the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept and preliminary evaluation of the seismic safety. It is limited on investigation of dynamic features of building structures, the building dynamical experiments and experimental investigation of the equipment

  19. Safety study application guide

    International Nuclear Information System (INIS)

    1993-07-01

    Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly open-quotes lowclose quotes) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, open-quotes Technical Safety Requirements,close quotes and 5480.23, open-quotes Nuclear Safety Analysis Reports.close quotes A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis

  20. Seismic safety programme at NPP Paks. Propositions for coordinated international activity in seismic safety of the WWER-440 V-213

    International Nuclear Information System (INIS)

    Katona, T.

    1995-01-01

    This paper presents the Paks NPP seismic safety program, highlighting the specifics of the WWER-440/213 type in operation, and the results of work obtained so far. It covers the following scope: establishment of the seismic safety program (original seismic design, current requirements, principles and structure of the seismic safety program); implementation of the seismic safety program (assessing the seismic hazard of the site, development of the new concept of seismic safety for the NPP, assessing the seismic resistance of the building and the technology); realization of the seismic safety of higher level (technical solutions, drawings, realization); ideas and propositions for coordinated international activity

  1. Seismic safety research program plan

    International Nuclear Information System (INIS)

    1987-05-01

    This document presents a plan for seismic research to be performed by the Structural and Seismic Engineering Branch in the Office of Nuclear Regulatory Research. The plan describes the regulatory needs and related research necessary to address the following issues: uncertainties in seismic hazard, earthquakes larger than the design basis, seismic vulnerabilities, shifts in building frequency, piping design, and the adequacy of current criteria and methods. In addition to presenting current and proposed research within the NRC, the plan discusses research sponsored by other domestic and foreign sources

  2. Armenia nuclear power plant. Overview of the present situation and its seismic safety

    International Nuclear Information System (INIS)

    Godoy, A.

    1993-01-01

    The presentation covers the problems of seismic safety of the two Armenian WWER type NPPs in the context of the energy situation in Armenia. Since the seismicity of the region is hazardous the upgrading of the seismic level is necessary and is considered feasible. A more complete and systematic approach to this problem is required. In this regard recommendations for seismic and site related safety which should be implemented are cited in the paper and a two phase approach is proposed in view of IAEA Safety Codes and Safety Guides

  3. Seismic safety margins research program overview

    International Nuclear Information System (INIS)

    Tokarz, F.J.; Smith, P.D.

    1978-01-01

    A multiyear seismic research program has been initiated at the Lawrence Livermore Laboratory. This program, the Seismic Safety Margins Research Program (SSMRP) is funded by the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The program is designed to develop a probabilistic systems methodology for determining the seismic safety margins of nuclear power plants. Phase I, extending some 22 months, began in July 1978 at a funding level of approximately $4.3 million. Here we present an overview of the SSMRP. Included are discussions on the program objective, the approach to meet the program goal and objectives, end products, the probabilistic systems methodology, and planned activities for Phase I

  4. IAEA establishes International Seismic Safety Centre

    International Nuclear Information System (INIS)

    2008-01-01

    Full text: The IAEA today officially inaugurated an international centre to coordinate efforts for protecting nuclear installations against the effects of earthquakes. The International Seismic Safety Centre (ISSC), which has been established within the IAEA's Department of Nuclear Safety and Security, will serve as a focal point on seismic safety for nuclear installations worldwide. ISSC will assist countries on the assessment of seismic hazards of nuclear facilities to mitigate the consequences of strong earthquakes. 'With safety as our first priority, it is vital that we pool all expert knowledge available worldwide to assist nuclear operators and regulators to be well prepared for coping with major seismic events,' said Antonio Godoy, Acting Head of the IAEA's Engineering Safety Section and leader of the ISSC. 'The creation of the ISSC represents the culmination of three decades of the IAEA's active and recognized involvement in this matter through the development of an updated set of safety standards and the assistance to Member States for their application.' To further seismic safety at nuclear installations worldwide, the ISSC will: - Promote knowledge sharing among the international community in order to avoid or mitigate the consequences of extreme seismic events on nuclear installations; - Support countries through advisory services and training courses; and - Enhance seismic safety by utilizing experience gained from previous seismic events in member states. The centre is supported by a scientific committee of high-level experts from academic, industrial and nuclear safety authorities that will advise the ISSC on implementation of its programme. Experts have been nominated from seven specialized areas, including geology and tectonics, seismology, seismic hazard, geotechnical engineering, structural engineering, equipment, and seismic risk. Japan and the United States have both contributed initial funds for creation of the centre, which will be based at

  5. School Chemistry Laboratory Safety Guide

    Science.gov (United States)

    Brundage, Patricia; Palassis, John

    2006-01-01

    The guide presents information about ordering, using, storing, and maintaining chemicals in the high school laboratory. The guide also provides information about chemical waste, safety and emergency equipment, assessing chemical hazards, common safety symbols and signs, and fundamental resources relating to chemical safety, such as Material…

  6. Preclosure Safety Analysis Guide

    International Nuclear Information System (INIS)

    D.D. Orvis

    2003-01-01

    A preclosure safety analysis (PSA) is a required element of the License Application (LA) for the high- level radioactive waste repository at Yucca Mountain. This guide provides analysts and other Yucca Mountain Repository Project (the Project) personnel with standardized methods for developing and documenting the PSA. The definition of the PSA is provided in 10 CFR 63.2, while more specific requirements for the PSA are provided in 10 CFR 63.112, as described in Sections 1.2 and 2. The PSA requirements described in 10 CFR Part 63 were developed as risk-informed performance-based regulations. These requirements must be met for the LA. The PSA addresses the safety of the Geologic Repository Operations Area (GROA) for the preclosure period (the time up to permanent closure) in accordance with the radiological performance objectives of 10 CFR 63.111. Performance objectives for the repository after permanent closure (described in 10 CFR 63.113) are not mentioned in the requirements for the PSA and they are not considered in this guide. The LA will be comprised of two phases: the LA for construction authorization (CA) and the LA amendment to receive and possess (R and P) high-level radioactive waste (HLW). PSA methods must support the safety analyses that will be based on the differing degrees of design detail in the two phases. The methods described herein combine elements of probabilistic risk assessment (PRA) and deterministic analyses that comprise a risk-informed performance-based safety analysis. This revision to the PSA guide was prepared for the following objectives: (1) To correct factual and typographical errors. (2) To provide additional material suggested from reviews by the Project, the U.S. Department of Energy (DOE), and U.S. Nuclear Regulatory Commission (NRC) Staffs. (3) To update material in accordance with approaches and/or strategies adopted by the Project. In addition, a principal objective for the planned revision was to ensure that the methods and

  7. Seismic safety of nuclear power plants

    International Nuclear Information System (INIS)

    Guerpinar, A.; Godoy, A.

    2001-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  8. Seismic Safety Of Simple Masonry Buildings

    International Nuclear Information System (INIS)

    Guadagnuolo, Mariateresa; Faella, Giuseppe

    2008-01-01

    Several masonry buildings comply with the rules for simple buildings provided by seismic codes. For these buildings explicit safety verifications are not compulsory if specific code rules are fulfilled. In fact it is assumed that their fulfilment ensures a suitable seismic behaviour of buildings and thus adequate safety under earthquakes. Italian and European seismic codes differ in the requirements for simple masonry buildings, mostly concerning the building typology, the building geometry and the acceleration at site. Obviously, a wide percentage of buildings assumed simple by codes should satisfy the numerical safety verification, so that no confusion and uncertainty have to be given rise to designers who must use the codes. This paper aims at evaluating the seismic response of some simple unreinforced masonry buildings that comply with the provisions of the new Italian seismic code. Two-story buildings, having different geometry, are analysed and results from nonlinear static analyses performed by varying the acceleration at site are presented and discussed. Indications on the congruence between code rules and results of numerical analyses performed according to the code itself are supplied and, in this context, the obtained result can provide a contribution for improving the seismic code requirements

  9. Seismic failure modes and seismic safety of Hardfill dam

    Directory of Open Access Journals (Sweden)

    Kun Xiong

    2013-04-01

    Full Text Available Based on microscopic damage theory and the finite element method, and using the Weibull distribution to characterize the random distribution of the mechanical properties of materials, the seismic response of a typical Hardfill dam was analyzed through numerical simulation during the earthquakes with intensities of 8 degrees and even greater. The seismic failure modes and failure mechanism of the dam were explored as well. Numerical results show that the Hardfill dam remains at a low stress level and undamaged or slightly damaged during an earthquake with an intensity of 8 degrees. During overload earthquakes, tensile cracks occur at the dam surfaces and extend to inside the dam body, and the upstream dam body experiences more serious damage than the downstream dam body. Therefore, under the seismic conditions, the failure pattern of the Hardfill dam is the tensile fracture of the upstream regions and the dam toe. Compared with traditional gravity dams, Hardfill dams have better seismic performance and greater seismic safety.

  10. Nuclear criticality safety guide

    International Nuclear Information System (INIS)

    Pruvost, N.L.; Paxton, H.C.

    1996-09-01

    This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators

  11. Nuclear criticality safety guide

    Energy Technology Data Exchange (ETDEWEB)

    Pruvost, N.L.; Paxton, H.C. [eds.

    1996-09-01

    This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

  12. Nucelar reactor seismic safety analysis techniques

    International Nuclear Information System (INIS)

    Cummings, G.E.; Wells, J.E.; Lewis, L.C.

    1979-04-01

    In order to provide insights into the seismic safety requirements for nuclear power plants, a probabilistic based systems model and computational procedure have been developed. This model and computational procedure will be used to identify where data and modeling uncertainties need to be decreased by studying the effect of these uncertainties on the probability of radioactive release and the probability of failure of various structures, systems, and components. From the estimates of failure and release probabilities and their uncertainties the most sensitive steps in the seismic methodologies can be identified. In addition, the procedure will measure the uncertainty due to random occurrences, e.g. seismic event probabilities, material property variability, etc. The paper discusses the elements of this systems model and computational procedure, the event-tree/fault-tree development, and the statistical techniques to be employed

  13. Seismic safety research program plan

    International Nuclear Information System (INIS)

    1985-06-01

    This plan describes the safety issues, regulatory needs, and the research necessary to address these needs. The plan also discusses the relationship between current and proposed research within the NRC and research sponsored by other government agencies, universities, industry groups, professional societies, and foreign sources

  14. Safety review for seismic qualification on nuclear power plant equipment

    International Nuclear Information System (INIS)

    Fang Qingxian

    1995-01-01

    The standards and requirements for seismic qualification of nuclear power plant's component have been fully addressed, including the scope of seismic qualification, the approach and the method of common seismic qualification, the procedure of the seismic tests, and the criteria for the seismic qualification review. The problems discovered in the safety review and the solution for these problems and some other issues are also discussed

  15. Safety design guides for containment extension for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for containment extension describes the containment isolation philosophy and containment extension requirements. The metal extensions and components falling within the scope of ASME Section III are classified in accordance with the CAN/CSA-N285.0 and CAN/CSA-N285.3. The special consideration for the leak monitoring capability, seismic qualification and inspection requirements for containment extensions, etc., are defined in this design guide. In addition, the containment isolation systems are defined and summarized schematically in appendix A. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. (Author) .new

  16. Occupational radiation protection. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. In 1996, the Agency published Safety Fundamentals on Radiation Protection and the Safety of Radiation Sources (IAEA Safety Series No. 120) and International Basic Safety Standards for Protection against Ionizing, Radiation and for the Safety of Radiation Sources (IAEA Safety Series No. 115), both of which were jointly sponsored by the Food and Agriculture Organization of the United Nations, the IAEA, the International Labour Organisation, the OECD Nuclear Energy Agency, the Pan American Health Organization and the World Health Organization. These publications set out, respectively, the objectives and principles for radiation safety and the requirements to be met to apply the principles and to achieve the objectives. The establishment of safety requirements and guidance on occupational radiation protection is a major component of the support for radiation safety provided by the IAEA to its Member States. The objective of the IAEA's occupational protection programme is to promote an internationally harmonized approach to the optimization of occupational radiation protection, through the development and application of guidelines for restricting radiation exposures and applying current radiation protection techniques in the workplace. Guidance on meeting the requirements of the Basic Safety Standards for occupational protection is provided in three interrelated Safety Guides, one giving general guidance on the development of occupational radiation protection programmes and two giving more detailed guidance on the monitoring and assessment of workers' exposure due to external radiation sources and from intakes of radionuclides, respectively. These Safety

  17. Occupational radiation protection. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. In 1996, the Agency published Safety Fundamentals on Radiation Protection and the Safety of Radiation Sources (IAEA Safety Series No. 120) and International Basic Safety Standards for Protection against Ionizing, Radiation and for the Safety of Radiation Sources (IAEA Safety Series No. 115), both of which were jointly sponsored by the Food and Agriculture Organization of the United Nations, the IAEA, the International Labour Organisation, the OECD Nuclear Energy Agency, the Pan American Health Organization and the World Health Organization. These publications set out, respectively, the objectives and principles for radiation safety and the requirements to be met to apply the principles and to achieve the objectives. The establishment of safety requirements and guidance on occupational radiation protection is a major component of the support for radiation safety provided by the IAEA to its Member States. The objective of the IAEA's occupational protection programme is to promote an internationally harmonized approach to the optimization of occupational radiation protection, through the development and application of guidelines for restricting radiation exposures and applying current radiation protection techniques in the workplace. Guidance on meeting the requirements of the Basic Safety Standards for occupational protection is provided in three interrelated Safety Guides, one giving general guidance on the development of occupational radiation protection programmes and two giving more detailed guidance on the monitoring and assessment of workers' exposure due to external radiation sources and from intakes of radionuclides, respectively. These Safety

  18. Occupational radiation protection. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. In 1996, the Agency published Safety Fundamentals on Radiation Protection and the Safety of Radiation Sources (IAEA Safety Series No. 120) and International Basic Safety Standards for Protection against Ionizing, Radiation and for the Safety of Radiation Sources (IAEA Safety Series No. 115), both of which were jointly sponsored by the Food and Agriculture Organization of the United Nations, the IAEA, the International Labour Organisation, the OECD Nuclear Energy Agency, the Pan American Health Organization and the World Health Organization. These publications set out, respectively, the objectives and principles for radiation safety and the requirements to be met to apply the principles and to achieve the objectives. The establishment of safety requirements and guidance on occupational radiation protection is a major component of the support for radiation safety provided by the IAEA to its Member States. The objective of the IAEA's occupational protection programme is to promote an internationally harmonized approach to the optimization of occupational radiation protection, through the development and application of guidelines for restricting radiation exposures and applying current radiation protection techniques in the workplace. Guidance on meeting the requirements of the Basic Safety Standards for occupational protection is provided in three interrelated Safety Guides, one giving general guidance on the development of occupational radiation protection programmes and two giving more detailed guidance on the monitoring and assessment of workers' exposure due to external radiation sources and from intakes of radionuclides, respectively. These Safety

  19. Occupational radiation protection. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. In 1996, the Agency published Safety Fundamentals on Radiation Protection and the Safety of Radiation Sources (IAEA Safety Series No. 120) and International Basic Safety Standards for Protection against Ionizing, Radiation and for the Safety of Radiation Sources (IAEA Safety Series No. 115), both of which were jointly sponsored by the Food and Agriculture Organization of the United Nations, the IAEA, the International Labour Organisation, the OECD Nuclear Energy Agency, the Pan American Health Organization and the World Health Organization. These publications set out, respectively, the objectives and principles for radiation safety and the requirements to be met to apply the principles and to achieve the objectives. The establishment of safety requirements and guidance on occupational radiation protection is a major component of the support for radiation safety provided by the IAEA to its Member States. The objective of the IAEA's occupational protection programme is to promote an internationally harmonized approach to the optimization of occupational radiation protection, through the development and application of guidelines for restricting radiation exposures and applying current radiation protection techniques in the workplace. Guidance on meeting the requirements of the Basic Safety Standards for occupational protection is provided in three interrelated Safety Guides, one giving general guidance on the development of occupational radiation protection programmes and two giving more detailed guidance on the monitoring and assessment of workers' exposure due to external radiation sources and from intakes of radionuclides, respectively. These Safety

  20. Probabilistic safety analysis procedures guide

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Bari, R.A.; Buslik, A.J.

    1984-01-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of tissues affecting reactor safety. This guide addresses the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant and from loss of offsite electric power. The scope includes analyses of problem-solving (cognitive) human errors, a determination of importance of the various core damage accident sequences, and an explicit treatment and display of uncertainties for the key accident sequences. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance) and the risk associated with external accident initiators, as consensus is developed regarding suitable methodologies in these areas. This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are essential for regulatory decision making. Methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study

  1. Guide On Safety Tests

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-05-15

    This book tells US important things to do safety tests, which includes basic caution for experiment treatment of used materials such as ignition substance inflammables, explosive substance and toxic substance, handling of used equipment like inflammable device, machine, high pressure device, high pressure gas, and high energy device, first aid such as addiction by drug, flame, radiation exposure, and heart massage treatment of waste in laboratory like cautions on general treatment, handling of inorganic waste, organic waste and waste treatment with disposal facilities.

  2. Guide On Safety Tests

    International Nuclear Information System (INIS)

    1987-05-01

    This book tells US important things to do safety tests, which includes basic caution for experiment treatment of used materials such as ignition substance inflammables, explosive substance and toxic substance, handling of used equipment like inflammable device, machine, high pressure device, high pressure gas, and high energy device, first aid such as addiction by drug, flame, radiation exposure, and heart massage treatment of waste in laboratory like cautions on general treatment, handling of inorganic waste, organic waste and waste treatment with disposal facilities.

  3. Seismic safety in nuclear-waste disposal

    International Nuclear Information System (INIS)

    Carpenter, D.W.; Towse, D.

    1979-01-01

    Seismic safety is one of the factors that must be considered in the disposal of nuclear waste in deep geologic media. This report reviews the data on damage to underground equipment and structures from earthquakes, the record of associated motions, and the conventional methods of seismic safety-analysis and engineering. Safety considerations may be divided into two classes: those during the operational life of a disposal facility, and those pertinent to the post-decommissioning life of the facility. Operational hazards may be mitigated by conventional construction practices and site selection criteria. Events that would materially affect the long-term integrity of a decommissioned facility appear to be highly unlikely and can be substantially avoided by conservative site selection and facility design. These events include substantial fault movement within the disposal facility and severe ground shaking in an earthquake epicentral region. Techniques need to be developed to address the question of long-term earthquake probability in relatively aseismic regions, and for discriminating between active and extinct faults in regions where earthquake activity does not result in surface ruptures

  4. Seismic safety in nuclear-waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, D.W.; Towse, D.

    1979-04-26

    Seismic safety is one of the factors that must be considered in the disposal of nuclear waste in deep geologic media. This report reviews the data on damage to underground equipment and structures from earthquakes, the record of associated motions, and the conventional methods of seismic safety-analysis and engineering. Safety considerations may be divided into two classes: those during the operational life of a disposal facility, and those pertinent to the post-decommissioning life of the facility. Operational hazards may be mitigated by conventional construction practices and site selection criteria. Events that would materially affect the long-term integrity of a decommissioned facility appear to be highly unlikely and can be substantially avoided by conservative site selection and facility design. These events include substantial fault movement within the disposal facility and severe ground shaking in an earthquake epicentral region. Techniques need to be developed to address the question of long-term earthquake probability in relatively aseismic regions, and for discriminating between active and extinct faults in regions where earthquake activity does not result in surface ruptures.

  5. Role of seismic PRA in seismic safety decisions of nuclear power plants

    International Nuclear Information System (INIS)

    Ravindra, M.K.; Kennedy, R.P.; Sues, R.H.

    1985-01-01

    This paper highlights the important roles that seismic probabilistic risk assessments (PRAs) can play in the seismic safety decisions of nuclear power plants. If a seismic PRA has been performed for a plant, its results can be utilized to evaluate the seismic capability beyond the safe shutdown event (SSE). Seismic fragilities of key structures and equipment, fragilities of dominant plant damage states and the frequencies of occurrence of these plant damage states are reviewed to establish the seismic safety of the plant beyond the SSE level. Guidelines for seismic margin reviews and upgrading may be developed by first identifying the generic classes of structures and equipment that have been shown to be dominant risk contributors in the completed seismic PRAs, studying the underlying causes for their contribution and examining why certain other items (e.g., piping) have not proved to be high-risk-contributors

  6. Seismic safety of building structures of NPP Kozloduy III

    International Nuclear Information System (INIS)

    Varbanov, G.I.; Kostov, M.K.; Stefanov, D.D.; Kaneva, A.D.

    2005-01-01

    In the proposed paper is presented a general summary of the analyses carried out to evaluate the dynamic behavior and to assess the seismic safety of some safety related building structures of NPP Kozloduy. The design seismic loads for the site of Kozloduy NPP has been reevaluated and increased during and after the construction of investigated Units 5 and 6. Deterministic and probabilistic approaches are applied to assess the seismic vulnerability of the investigated structures, taking into account the newly defined seismic excitations. The presented results show sufficient seismic safety for the studied critical structures and good efficiency of the seismic upgrading. The applicability of the investigated structures at sites with some higher seismic activities is discussed. The presented study is dealing mainly with the civil structures of the Reactor building, Turbine hall, Diesel Generator Station and Water Intake Structure. (authors)

  7. 41 CFR 128-1.8006 - Seismic Safety Program requirements.

    Science.gov (United States)

    2010-07-01

    ... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false Seismic Safety Program requirements. 128-1.8006 Section 128-1.8006 Public Contracts and Property Management Federal Property Management Regulations System (Continued) DEPARTMENT OF JUSTICE 1-INTRODUCTION 1.80-Seismic Safety Program...

  8. 41 CFR 128-1.8004 - Seismic Safety Coordinators.

    Science.gov (United States)

    2010-07-01

    ... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false Seismic Safety Coordinators. 128-1.8004 Section 128-1.8004 Public Contracts and Property Management Federal Property Management Regulations System (Continued) DEPARTMENT OF JUSTICE 1-INTRODUCTION 1.80-Seismic Safety Program...

  9. Safety guides development process in Spain

    International Nuclear Information System (INIS)

    Butragueno, J.L.; Perello, M.

    1979-01-01

    Safety guides have become a major factor in the licensing process of nuclear power plants and related nuclear facilities of the fuel cycle. As far as the experience corroborates better and better engineering methodologies and procedures, the results of these are settled down in form of standards, guides, and similar issues. This paper presents the actual Spanish experience in nuclear standards and safety guides development. The process to develop a standard or safety guide is shown. Up to date list of issued and on development nuclear safety guides is included and comments on the future role of nuclear standards in the licensing process are made. (author)

  10. International contributions of JNES on seismic safety areas

    International Nuclear Information System (INIS)

    Ebisawa, Katsumi; Uchiyama, Yuichi; Yamada, Hiroyuki

    2010-01-01

    JNES actively promotes the international cooperation in seismic safety areas, aiming to play a role as the important international hub for it. To meet this purpose, JNES is now mainly focusing on the increased support of the international organizations including IAEA and the technological improvement in the seismic related assessment of Asian countries. This paper summarizes these efforts made by JNES. (author)

  11. Seismic safety margin assessment program (Annual safety research report, JFY 2010)

    International Nuclear Information System (INIS)

    Suzuki, Kenichi; Iijima, Toru; Inagaki, Masakatsu; Taoka, Hideto; Hidaka, Shinjiro

    2011-01-01

    Seismic capacity test data, analysis method and evaluation code provided by Seismic Safety Margin Assessment Program have been utilized for the support of seismic back-check evaluation of existing plants. The summary of the program in 2010 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. Many seismic capacity tests of various snubbers were conducted and quantitative seismic capacities were evaluated. One of the emergency diesel generator partial-model seismic capacity tests was conducted and quantitative seismic capacity was evaluated. Some of the analytical evaluations of piping-system seismic capacities were conducted. 2. Analysis method for minute evaluation of component seismic response. The difference of seismic response of large components such as primary containment vessel and reactor pressure vessel when they were coupled with 3-dimensional FEM building model or 1-dimensional lumped mass building model, was quantitatively evaluated. 3. Evaluation code for quantitative evaluation of seismic safety margin of systems, structures and components. As the example, quantitative evaluation of seismic safety margin of systems, structures and components were conducted for the reference plant. (author)

  12. Use of the t-distribution to construct seismic hazard curves for seismic probabilistic safety assessments

    Energy Technology Data Exchange (ETDEWEB)

    Yee, Eric [KEPCO International Nuclear Graduate School, Dept. of Nuclear Power Plant Engineering, Ulsan (Korea, Republic of)

    2017-03-15

    Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered.

  13. Use of the t-distribution to construct seismic hazard curves for seismic probabilistic safety assessments

    International Nuclear Information System (INIS)

    Yee, Eric

    2017-01-01

    Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered

  14. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    This publication supports the Safety Requirements on the Safety of Nuclear Power Plants: Design. This Safety Guide was prepared on the basis of a systematic review of all the relevant publications including the Safety Fundamentals, Safety of Nuclear Power Plants: Design, current and ongoing revisions of other Safety Guides, INSAG reports and other publications that have addressed the safety of nuclear power plants. This Safety Guide also provides guidance for Contracting Parties to the Convention on Nuclear Safety in meeting their obligations under Article 14 on Assessment and Verification of Safety. The Safety Requirements publication entitled Safety of Nuclear Power Plants: Design states that a comprehensive safety assessment and an independent verification of the safety assessment shall be carried out before the design is submitted to the regulatory body. This publication provides guidance on how this requirement should be met. This Safety Guide provides recommendations to designers for carrying out a safety assessment during the initial design process and design modifications, as well as to the operating organization in carrying out independent verification of the safety assessment of new nuclear power plants with a new or already existing design. The recommendations for performing a safety assessment are suitable also as guidance for the safety review of an existing plant. The objective of reviewing existing plants against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The methods and the recommendations of this Safety Guide can also be used by regulatory bodies for the conduct of the regulatory review and assessment. Although most recommendations of this Safety Guide are general and applicable to all types of nuclear reactors, some specific recommendations and examples apply mostly to water cooled reactors. Terms such as 'safety assessment', 'safety analysis' and 'independent

  15. Seismic safety in conducting large-scale blasts

    Science.gov (United States)

    Mashukov, I. V.; Chaplygin, V. V.; Domanov, V. P.; Semin, A. A.; Klimkin, M. A.

    2017-09-01

    In mining enterprises to prepare hard rocks for excavation a drilling and blasting method is used. With the approach of mining operations to settlements the negative effect of large-scale blasts increases. To assess the level of seismic impact of large-scale blasts the scientific staff of Siberian State Industrial University carried out expertise for coal mines and iron ore enterprises. Determination of the magnitude of surface seismic vibrations caused by mass explosions was performed using seismic receivers, an analog-digital converter with recording on a laptop. The registration results of surface seismic vibrations during production of more than 280 large-scale blasts at 17 mining enterprises in 22 settlements are presented. The maximum velocity values of the Earth’s surface vibrations are determined. The safety evaluation of seismic effect was carried out according to the permissible value of vibration velocity. For cases with exceedance of permissible values recommendations were developed to reduce the level of seismic impact.

  16. Seismic site evaluation practice and seismic design guide for NPP in Continent of China

    Energy Technology Data Exchange (ETDEWEB)

    Yuxian, Hu [State Seismological Bureau, Beijing, BJ (China). Inst. of Geophysics

    1997-03-01

    Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

  17. Seismic site evaluation practice and seismic design guide for NPP in Continent of China

    International Nuclear Information System (INIS)

    Hu Yuxian

    1997-01-01

    Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

  18. Seismic qualification of multiple interconnected safety-related cabinets in a high seismic zone

    International Nuclear Information System (INIS)

    Khan, M.R.; Chen, W.H.W.; Wang, T.Y.

    1993-01-01

    Certain safety-related multiple, interconnected electrical cabinets and the devices contained therein are required to perform their intended safety functions during and after a design basis seismic event. In general, seismic testing is performed to ensure the structural integrity of the cabinets and the functionality of their associated devices. Constrained by the shake table capacity, seismic testing is usually performed only for a limited number of interconnected cabinets. Also, original shake table tests performed usually did not provide detailed response information at various locations inside the cabinets. For operational and maintenance purposes, doors and panels of some cabinets may need to be opened while the adjacent cabinets are required to remain functional. In addition, in-cabinet response spectra need to be generated for the seismic qualification of new devices and the replacement parts. Consequently, seismic analysis of safety-related multiple, interconnected cabinets is frequently required for configurations which are different from the original tested conditions. This paper presents results of seismic tests of three interconnected safety-related cabinets and finite element analyses performed to compare the analytical results with those obtained from the cabinet seismic tests. Parametric analyses are performed to determine how many panels and doors can be opened while the adjacent cabinets still remain functional. The study indicates that for cabinets located in a high seismic zone, the critical damping of the cabinet is significantly higher than 5% to 7% typically used in qualifying electrical equipment. For devices mounted on the cabinet doors to performed their intended safety function, it requires stiffening of doors and that these doors be properly bolted to the cabinet frame. It also shows that even though doors and panels bolted to the cabinet frame are the primary seismic resistant element of the cabinet, opening of a limited number of them

  19. The reevaluation of seismic safety of existing nuclear power plant

    International Nuclear Information System (INIS)

    Kitagawa, Hiroshi; Tominaga, Shohei; Kumagai, Chiyoshi; Koshiba, Koremutsu; Kono, Tomonori; Agawa, Kazuyoshi; Kuwata, Kenichiro

    2003-01-01

    We have carried out additional geological surveys in order to enrich our database on geological faults in the vicinity of Shimane Nuclear Power Plant (NPP). Prior to additional geological surveys, given the social importance of nuclear power plants, we hypothetically assumed that almost the whole length of an area covered by surveys would be an active fault that must be considered in seismic design, and tried to reevaluate the seismic safety of the NPP by applying an input earthquake ground motion larger than the level at the design stage. As a result, we have confirmed that seismic safety of the NPP can be maintained. This paper describes the method that we employed to reevaluate the seismic safety of Shimane NPP. (author)

  20. Historical development of the seismic requirements for construction of nuclear power plants in the U.S. and worldwide and their current impact on cost and safety

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    2003-01-01

    The following topics are described and discussed: Historical development of NPP seismic design requirements: Peak ground acceleration; Response spectra and damping; Floor or amplified response spectra; Effective high frequency response spectra; Seismic modeling procedures; Impact on cost (site preparation and foundations; site seismic response and generation of site dependent spectra). Potential use of indirect earthquake experience data in design and construction of NPP. Seismic contribution to safety. The following facts are summarized in two Appendices: Seismic intensity scales, and GRS safety codes and guides. (P.A.)

  1. Seismic safety margins research program. Phase I final report - Overview

    International Nuclear Information System (INIS)

    Smith, P.D.; Dong, R.G.; Bernreuter, D.L.; Bohn, M.P.; Chuang, T.Y.; Cummings, G.E.; Johnson, J.J.; Mensing, R.W.; Wells, J.E.

    1981-04-01

    The Seismic Safety Margins Research Program (SSMRP) is a multiyear, multiphase program whose overall objective is to develop improved methods for seismic safety assessments of nuclear power plants, using a probabilistic computational procedure. The program is being carried out at the Lawrence Livermore National Laboratory and is sponsored by the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. Phase I of the SSMRP was successfully completed in January 1981: A probabilistic computational procedure for the seismic risk assessment of nuclear power plants has been developed and demonstrated. The methodology is implemented by three computer programs: HAZARD, which assesses the seismic hazard at a given site, SMACS, which computes in-structure and subsystem seismic responses, and SEISIM, which calculates system failure probabilities and radioactive release probabilities, given (1) the response results of SMACS, (2) a set of event trees, (3) a family of fault trees, (4) a set of structural and component fragility descriptions, and (5) a curve describing the local seismic hazard. The practicality of this methodology was demonstrated by computing preliminary release probabilities for Unit 1 of the Zion Nuclear Power Plant north of Chicago, Illinois. Studies have begun aimed at quantifying the sources of uncertainty in these computations. Numerous side studies were undertaken to examine modeling alternatives, sources of error, and available analysis techniques. Extensive sets of data were amassed and evaluated as part of projects to establish seismic input parameters and to produce the fragility curves. (author)

  2. Seismic safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Gurpinar, A.; Godoy, A.

    1995-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in WWER type nuclear power plants during the past five years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on B enchmark study for the seismic analysis and testing of WWER type nuclear power plants . These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  3. The SISIFO project: Seismic Safety at High Schools

    Science.gov (United States)

    Peruzza, Laura; Barnaba, Carla; Bragato, Pier Luigi; Dusi, Alberto; Grimaz, Stefano; Malisan, Petra; Saraò, Angela; Mucciarelli, Marco

    2014-05-01

    For many years, the Italian scientific community has faced the problem of the reduction of earthquake risk using innovative educational techniques. Recent earthquakes in Italy and around the world have clearly demonstrated that seismic codes alone are not able to guarantee an effective mitigation of risk. After the tragic events of San Giuliano di Puglia (2002), where an earthquake killed 26 school children, special attention was paid in Italy to the seismic safety of schools, but mainly with respect to structural aspects. Little attention has been devoted to the possible and even significant damage to non-structural elements (collapse of ceilings, tipping of cabinets and shelving, obstruction of escape routes, etc..). Students and teachers trained on these aspects may lead to a very effective preventive vigilance. Since 2002, the project EDURISK (www.edurisk.it) proposed educational tools and training programs for schools, at primary and middle levels. More recently, a nationwide campaign aimed to adults (www.iononrischio.it) was launched with the extensive support of civil protection volounteers. There was a gap for high schools, and Project SISIFO was designed to fill this void and in particular for those schools with technical/scientific curricula. SISIFO (https://sites.google.com/site/ogssisifo/) is a multidisciplinary initiative, aimed at the diffusion of scientific culture for achieving seismic safety in schools, replicable and can be structured in training the next several years. The students, helped by their teachers and by experts from scientific institutions, followed a course on specialized training on earthquake safety. The trial began in North-East Italy, with a combination of hands-on activities for the measurement of earthquakes with low-cost instruments and lectures with experts in various disciplines, accompanied by specifically designed teaching materials, both on paper and digital format. We intend to raise teachers and students knowledge of the

  4. Criticality safety basics, a study guide

    Energy Technology Data Exchange (ETDEWEB)

    V. L. Putman

    1999-09-01

    This document is a self-study and classroom guide, for criticality safety of activities with fissile materials outside nuclear reactors. This guide provides a basic overview of criticality safety and criticality accident prevention methods divided into three parts: theory, application, and history. Except for topic emphasis, theory and history information is general, while application information is specific to the Idaho National Engineering and Environmental Laboratory (INEEL). Information presented here should be useful to personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. However, the guide's primary target audience is fissile material handler candidates.

  5. Criticality safety basics, a study guide

    International Nuclear Information System (INIS)

    Putman, V.L.

    1999-01-01

    This document is a self-study and classroom guide, for criticality safety of activities with fissile materials outside nuclear reactors. This guide provides a basic overview of criticality safety and criticality accident prevention methods divided into three parts: theory, application, and history. Except for topic emphasis, theory and history information is general, while application information is specific to the Idaho National Engineering and Environmental Laboratory (INEEL). Information presented here should be useful to personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. However, the guide's primary target audience is fissile material handler candidates

  6. Seismic Safety Margins Research Program: a concluding look

    International Nuclear Information System (INIS)

    Cummings, G.E.

    1984-01-01

    The Seismic Safety Margins Research Program (SSMRP) was started in 1978 with the goal of developing tools and data bases to compute the probability of earthquake - caused radioactive release from commercial nuclear power plants. These tools and data bases were to help NRC to assess seismic safety at nuclear plants. The methodology to be used was finalized in 1982 and applied to the Zion Nuclear Power Station. The SSMRP will be completed this year with the development of a more simplified method of analysis and a demonstration of its use on Zion. This simplified method is also being applied to a boiling-water-reactor, LaSalle

  7. Seismic qualification of existing safety class manipulators

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Moran, T.J.

    1992-01-01

    There are two bridge type electromechanical manipulators within a nuclear fuel handling facility which were constructed over twenty-five years ago. At that time, there were only minimal seismic considerations. These manipulators together with the facility are being reactivated. Detailed analyses have shown that the manipulators will satisfy the requirements of ANSI/AISC N690-1984 when they are subjected to loadings including the site specific design basis earthquake. 4 refs

  8. Collection and accumulation of seismic safety research findings, and considerations for information dissemination

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Seismic Safety Division of JNES is collecting and analyzing the findings of seismic safety research, and is developing a system to organize and disseminate the information internally and internationally. These tasks have been conducted in response to the lessons learned from Fukushima Daiichi NPP accident. The overview of the tasks is as follows; 1) Collection of the knowledge and findings from seismic safety research. JNES collects information on seismic safety researches including the 2011 off the Pacific coast of Tohoku Earthquake. The information is analyzed whether it is important for regulation to increase seismic safety of NPP. 2) Constructing database of seismic safety research. JNES collects information based on documents published by committee and constructs database of active faults around NPP sites in order to incorporate in the seismic safety review. 3) Dissemination of information related to seismic safety. JNES disseminates outcomes of own researches internally and internationally. (author)

  9. Collection and accumulation of seismic safety research findings, and considerations for information dissemination

    International Nuclear Information System (INIS)

    2013-01-01

    Seismic Safety Division of JNES is collecting and analyzing the findings of seismic safety research, and is developing a system to organize and disseminate the information internally and internationally. These tasks have been conducted in response to the lessons learned from Fukushima Daiichi NPP accident. The overview of the tasks is as follows; 1) Collection of the knowledge and findings from seismic safety research. JNES collects information on seismic safety researches including the 2011 off the Pacific coast of Tohoku Earthquake. The information is analyzed whether it is important for regulation to increase seismic safety of NPP. 2) Constructing database of seismic safety research. JNES collects information based on documents published by committee and constructs database of active faults around NPP sites in order to incorporate in the seismic safety review. 3) Dissemination of information related to seismic safety. JNES disseminates outcomes of own researches internally and internationally. (author)

  10. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.M.; Ketcham, D.R.

    1989-01-01

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table testing which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its ''Generic Safety Evaluation Report'' approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the United States and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluating program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  11. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.; Ketcham, D.

    1989-01-01

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table tested which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its Generic Safety Evaluation Report approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the US and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective approach developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluation program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  12. A progressive methodology for seismic safety evaluation of gravity dams

    International Nuclear Information System (INIS)

    Ghrib, F.; Leger, P.; Tinawi, R.; Lupien, R.; Veilleux, M.

    1995-01-01

    A progressive methodology for the seismic safety evaluation of existing concrete gravity dams was described. The methodology was based on five structural analysis levels with increasing complexity to represent inertia forces, dam-foundation and dam-interaction mechanisms, as well as concrete cracking. The five levels were (1) preliminary screening, (2) pseudo-static method, (3) pseudo-dynamic method, (4) linear time history analysis, and (5) non-linear history analysis. The first four levels of analysis were applied for the seismic safety evaluation of Paugan gravity dam (Quebec). Results showed that internal forces from pseudo-dynamic, response spectra and transient finite element analyses could be used to interpret the dynamic stability of dams from familiar strength-based criteria. However, as soon as the base was cracked, the seismically induced forces were modified, and level IV analyses proved more suitable to handle rationally these complexities. 8 refs., 7 figs., 1 tab

  13. Regulatory control of radiation sources. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The basic requirements for the protection of persons against exposure to ionizing radiation and for the safety of radiation sources were established in the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (the Basic Safety Standards), jointly sponsored by the Food and Agriculture Organization of the United Nations (FAO), the International Atomic Energy Agency (IAEA), the International Labour Organization (ILO), the OECD Nuclear Energy Agency (OECD/ NEA), the Pan American Health Organization (PAHO) and the World Health Organization (WHO) (the Sponsoring Organizations). The application of the Basic Safety Standards is based on the presumption that national infrastructures are in place to enable governments to discharge their responsibilities for radiation protection and safety. Requirements relating to the legal and governmental infrastructure for the safety of nuclear facilities and sources of ionizing radiation, radiation protection, the safe management of radioactive waste and the safe transport of radioactive material are established in the Safety Requirements on Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety, Safety Standards Series No. GS-R-1. This Safety Guide, which is jointly sponsored by the FAO, the IAEA, the International Labour Office, the PAHO and the WHO, gives detailed guidance on the key elements for the organization and operation of a national regulatory infrastructure for radiation safety, with particular reference to the functions of the national regulatory body that are necessary to ensure the implementation of the Basic Safety Standards. The Safety Guide is based technically on material first published in IAEA-TECDOC-10671, which was jointly sponsored by the FAO, the IAEA, the OECD/NEA, the PAHO and the WHO. The requirements established in GS-R-1 have been taken into account. The Safety Guide is oriented towards national

  14. Nuclear safety guide. TID-7016, Revision 2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1978-01-01

    The Nuclear Safety Guide was first issued in 1956 as classified AEC report LA-2063 and was reprinted the next year, unclassified, as TID-7016. Revision 1, published in 1961, extended the scope and refined the guiding information. The present revision of the Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams, experienced in the field. The reader will find a significant change in the character of information presented in this version. Nuclear Criticality Safety has matured in the past twelve years. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the formerGuides by employing appropriate safety factors. In fact, it becomes incumbent on the Criticality Safety Specialist to necessarily impose safety factors consistent with the possible normal and abnormal credible contingencies of an operation as revealed by his evaluation. In its present form the Guide easily becomes a suitable module in any compendium or handbook tailored for internal use by organizations. It is hoped the Guide will continue to serve immediate needs and will encourage continuing and more comprehensive efforts toward organizing nuclear criticality safety information

  15. Nuclear safety guide. TID-7016, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, J T [ed.

    1978-05-01

    The Nuclear Safety Guide was first issued in 1956 as classified AEC report LA-2063 and was reprinted the next year, unclassified, as TID-7016. Revision 1, published in 1961, extended the scope and refined the guiding information. The present revision of the Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams, experienced in the field. The reader will find a significant change in the character of information presented in this version. Nuclear Criticality Safety has matured in the past twelve years. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the formerGuides by employing appropriate safety factors. In fact, it becomes incumbent on the Criticality Safety Specialist to necessarily impose safety factors consistent with the possible normal and abnormal credible contingencies of an operation as revealed by his evaluation. In its present form the Guide easily becomes a suitable module in any compendium or handbook tailored for internal use by organizations. It is hoped the Guide will continue to serve immediate needs and will encourage continuing and more comprehensive efforts toward organizing nuclear criticality safety information.

  16. 41 CFR 128-1.8005 - Seismic safety standards.

    Science.gov (United States)

    2010-07-01

    ... the model building codes that the Interagency Committee on Seismic Safety in Construction (ICSSC...) Uniform Building Code (UBC); (2) The 1992 Supplement to the Building Officials and Code Administrators International (BOCA) National Building Code (NBC); and (3) The 1992 Amendments to the Southern Building Code...

  17. Seismic requalification of a safety class crane

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Moran, T.J.

    1991-01-01

    A remotely operated 5-ton crane within a nuclear fuel handling facility was designed and constructed over 25 years ago. At that time, less severe design criteria, particularly on seismic loadings, were in use. This crane is being reactivated and requalified under new design criteria with loads including a site specific design basis earthquake. Detailed analyses of the crane show that the maximum stress coefficient is less than 90% of the code allowable, indicating that this existing crane is able to withstand loadings including those from the design basis earthquake. 3 refs., 8 figs., 2 tabs

  18. Development of Seismic Safety Assessment Technology for Containment Structure

    Energy Technology Data Exchange (ETDEWEB)

    Jang, J.B.; Suh, Y.P.; Lee, J.R. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    This final report is made based on the research results of seismic analysis and seismic margin assessment field, carried out during 3rd stage ('01.4.1{approx}'02.3.31) under financial support of MOST(Ministry of Science and Technology). The objective of this research is to develop the soil - structure interaction analysis technique with high reliability, the main research subjects, performed during 3rd stage are as follows. 1) Preparation of user's guide manual for SSI analysis with high accuracy. 2) Sensitivity analysis of effective shear strain and seismic input motion. 3) Database construction of Hualien earthquake recorded data. (author). 21 refs., 27 figs., 2 tabs.

  19. Nuclear safety guide TID-7016 Revision 2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1980-01-01

    The present revision of TID-7016 Nuclear Safety Guide is discussed. This Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams, experienced in the field. The reader will find a significant change in the character of information presented in this version. Nuclear Criticality Safety has matured in the past twelve years. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available, information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the former Guides by employing appropriate safety factors. In fact, it becomes incumbent on the Criticality Safety Specialist to necessarily impose safety factors consistent with the possible normal and abnormal credible contingencies of an operation as revealed by his evaluation. In its present form the Guide easily becomes a suitable module in any compendium or handbook tailored for internal use by organizations. It is hoped the Guide will continue to serve immediate needs and will encourage continuing and more comprehensive efforts toward organizing nuclear criticality safety information

  20. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  1. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  2. A Guide to Laser Safety

    Energy Technology Data Exchange (ETDEWEB)

    Davies, W M

    1998-09-01

    In one of the few volumes dedicated to laser safety to appear since the 'bible' of Sliney and Wolbarsht, Roy Henderson sets out to provide the reader with a practical account of both the philosophy and practice across contemporary application of lasers. The book is split into three sections. The first section is essentially a non-mathematical review of lasers, optical hazards and laser safety. It is intended as an easily digestible introduction to the subject, conveying the primary concepts of laser safety without the camouflage of equations. This piece of text is manifestly readable by all who have interest in the topic. The second section introduces more meat onto the bones introduced in the first section and some of the practical mathematics necessary to determine optical irradiance in simple laser beams. The book is not intended as a scientific treatise and rigorous treatment of laser physics is left (for the better) to other texts. Laser hazard assessment and safety management are covered in sufficient detail to allow the reader to understand what precautions are necessary to mitigate the risks of laser use. The final section takes a brief look at laser safety in a number of specific industrial applications. These include industrial processing, medicine, telecommunications and entertainment. These should be taken in the context of the second section and are not stand-alone text. With few typographical errors, and packed with practical hints, this book serves as an excellent text for any educational course on laser safety and provides a quick and easy reference for laser safety officers. (book review: A Roy Henderson - ISBN: 0 412 72940 7)

  3. The regulatory requirements, design bases, researches and assessments in the field of Ukrainian NPP's seismic safety

    International Nuclear Information System (INIS)

    Mykolaychuk, O.; Mayboroda, O.; Krytskyy, V.; Karnaukhov, O.

    2001-01-01

    State Nuclear Regulatory Authority of Ukraine (SNRA) pays large attention to problem of nuclear installations seismic stability. As a result the seismic design regulatory guides is revised, additional seismic researches of NPP sites are conducted, seismic reassessment of NPP designs were begun. The experts involved address all seismic related factors under close contact with the staff of NPP, design institutes and research organizations. This document takes stock on the situation and the research programs. (author)

  4. Nuclear safety guide: TID--7016, Revision 2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1978-01-01

    The Nuclear Safety Guide was first issued in 1956 as classified AEC report LA-2063 and was reprinted the next year, unclassified, as TID-7016. Revision 1, published in 1961, extended the scope and refined the guiding information. Revision 2 of the Guide differs significantly from its predecessor in that the latter was intentionally conservative in its recommendations. Firmly based on experimental evidence of criticality, the original Guide and the first revision were considered to be of most value to organizations whose activities with fissionable materials were not extensive and, secondarily, that it would serve as a point of departure for members of established nuclear safety teams experienced in the field. The advance of calculational capability has permitted validated calculations to extend and substitute for experimental data. The broadened data base has enabled better interpolation, extension, and understanding of available information, especially in areas previously addressed by undefined but adequate factors of safety. The content has been thereby enriched in qualitative guidance. The information inherently contains, and the user can recapture, the quantitative guidance characteristic of the former Guides by employing appropriate safety factors

  5. 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

  6. New French basic safety rule on seismic input ground motions

    International Nuclear Information System (INIS)

    Forner, Sophie; Boulaigue, Yves

    2002-01-01

    French regulatory practice requires that the main safety functions of a land-based major nuclear facility, in particular in accordance with its specific characteristics, safe shutdown, cooling and containment of radioactive substances, be assured during and/or after earthquake events that can plausibly occur at the site where the installation is located. This rule specifies an acceptable method for determining the seismic motion to be taken into account when designing a facility to address the seismic risk. In regions where deformation factors are low, such as in metropolitan France, the intervals between strong earthquakes are long and it can be difficult to associate some earthquakes with known faults. In addition, despite substantial progress in recent years, it is difficult, given the French seismotectonic situation, to identify potentially seismogenic faults and determine the characteristics of the earthquakes that are liable to occur. Therefore, the approach proposed in this Basic Safety Rule is intended to avoid this difficulty by allowing for all direct and indirect influences that can play a role in the occurrence of earthquakes, as well as all seismic knowledge. Furthermore, as concerns calculation of seismic motion, the low number of records of strong motion in metropolitan France makes it necessary to use data from other regions of the world

  7. Safety design guide for safety related systems for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A.C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new.

  8. Safety design guide for safety related systems for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new

  9. Social acceptance for seismic safety of nuclear installations

    International Nuclear Information System (INIS)

    Oiso, Shinichi

    2010-01-01

    The social acceptance of seismic safety of the nuclear installations was considered based on the situation that people's concern and anxieties for it having risen by earthquake suffering of the Kashiwazaki Kariwa facility in 2007, etc. It aimed mainly to extract a social awareness (acknowledgment and evaluation) which is peculiar to the earthquake in the field of nuclear power generation, and to show the attention point concerning the public relations of seismic safety of the nuclear power plant. As a result, it was suggested that we should explain based on the opinion of the third party which has a high trust such as specialist scholars, and emphasize that the severe examinations of outside third parties such as committee of the prefecture are conducted. (author)

  10. Handbook of nuclear power plant seismic fragilities, Seismic Safety Margins Research Program

    International Nuclear Information System (INIS)

    Cover, L.E.; Bohn, M.P.; Campbell, R.D.; Wesley, D.A.

    1983-12-01

    The Seismic Safety Margins Research Program (SSMRP) has a gola to develop a complete fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor were made. These calculations required a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most would be usable for other sites as well

  11. Regulatory Control of Radiation Sources. Safety Guide

    International Nuclear Information System (INIS)

    2009-01-01

    This Safety Guide is intended to assist States in implementing the requirements established in Safety Standards Series No. GS-R-1, Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety, for a national regulatory infrastructure to regulate any practice involving radiation sources in medicine, industry, research, agriculture and education. The Safety Guide provides advice on the legislative basis for establishing regulatory bodies, including the effective independence of the regulatory body. It also provides guidance on implementing the functions and activities of regulatory bodies: the development of regulations and guides on radiation safety; implementation of a system for notification and authorization; carrying out regulatory inspections; taking necessary enforcement actions; and investigating accidents and circumstances potentially giving rise to accidents. The various aspects relating to the regulatory control of consumer products are explained, including justification, optimization of exposure, safety assessment and authorization. Guidance is also provided on the organization and staffing of regulatory bodies. Contents: 1. Introduction; 2. Legal framework for a regulatory infrastructure; 3. Principal functions and activities of the regulatory body; 4. Regulatory control of the supply of consumer products; 5. Functions of the regulatory body shared with other governmental agencies; 6. Organization and staffing of the regulatory body; 7. Documentation of the functions and activities of the regulatory body; 8. Support services; 9. Quality management for the regulatory system.

  12. Fire Safety. Managing School Facilities, Guide 6.

    Science.gov (United States)

    Department for Education and Employment, London (England). Architects and Building Branch.

    This booklet discusses how United Kingdom schools can manage fire safety and minimize the risk of fire. The guide examines what legislation school buildings must comply with and covers the major risks. It also describes training and evacuation procedures and provides guidance on fire precautions, alarm systems, fire fighting equipment, and escape…

  13. Department of Energy Construction Safety Reference Guide

    Energy Technology Data Exchange (ETDEWEB)

    1993-09-01

    DOE has adopted the Occupational Safety and Health Administration (OSHA) regulations Title 29 Code of Federal Regulations (CFR) 1926 ``Safety and Health Regulations for Construction,`` and related parts of 29 CFR 1910, ``Occupational Safety and Health Standards.`` This nonmandatory reference guide is based on these OSHA regulations and, where appropriate, incorporates additional standards, codes, directives, and work practices that are recognized and accepted by DOE and the construction industry. It covers excavation, scaffolding, electricity, fire, signs/barricades, cranes/hoists/conveyors, hand and power tools, concrete/masonry, stairways/ladders, welding/cutting, motor vehicles/mechanical equipment, demolition, materials, blasting, steel erection, etc.

  14. Geotechnical aspects of site evaluation and foundations for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2003-01-01

    This publication is a revision of the former safety standards of IAEA Safety Series No. 50-SG-S8. The scope has been extended to cover not only foundations but also design questions related to geotechnical science and engineering, such as the bearing capacity of foundations, design of earth structures and design of buried structures. Seismic aspects also play an important role in this field, and consequently the Safety Guide on Evaluation of Seismic Hazards for Nuclear Power Plants, Safety Standards Series No. NS-G-3.3, which discusses the determination of seismic input motion, is referenced on several occasions. The present Safety Guide provides an interpretation of the Safety Requirements on Site Evaluation for Nuclear Installations and guidance on how to implement them. It is intended for the use of safety assessors or regulators involved in the licensing process as well as the designers of nuclear power plants, and it provides them with guidance on the methods and procedures for analyses to support the assessment of the geotechnical aspects of the safety of nuclear power plants

  15. Geotechnical aspects of site evaluation and foundations for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    This publication is a revision of the former safety standards of IAEA Safety Series No. 50-SG-S8. The scope has been extended to cover not only foundations but also design questions related to geotechnical science and engineering, such as the bearing capacity of foundations, design of earth structures and design of buried structures Seismic aspects also play an important role in this field, and consequently the Safety Guide on Evaluation of Seismic Hazards for Nuclear Power Plants, Safety Standards Series No. NS-G-3.3, which discusses the determination of seismic input motion, is referenced on several occasions. The present Safety Guide provides an interpretation of the Safety Requirements on Site Evaluation for Nuclear Installations and guidance on how to implement them. It is intended for the use of safety assessors or regulators involved in the licensing process as well as the designers of nuclear power plants, and it provides them with guidance on the methods and procedures for analyses to support the assessment of the geotechnical aspects of the safety of nuclear power plants

  16. Seismic safety margin research program. Program plan, Revision I

    International Nuclear Information System (INIS)

    Smith, P.D.; Tokarz, F.J.; Bernreuter, D.L.; Cummings, G.E.; Chou, C.K.; Vagliente, V.N.

    1978-01-01

    The overall objective of the SSMRP is to develop mathematical models that realistically predict the probability of radioactive releases from seismically induced events in nuclear power plants. These models will be used for four purposes: (1) To perform sensitivity studies to determine the weak links in seismic methodology. The weak links will then be improved by research and development. (2) To estimate the probability of release for a plant. It is believed that the major difficulty in the program will be to obtain acceptably small confidence limits on the probability of release. (3) To estimate the conservatisms in the Standard Review Plan (SRP) seismic design methodology. This will be done by comparing the results of the SRP methodology and the methodology resulting from the research and development in (1). (4) To develop an improved seismic design methodology based on probability. The Phase I objective proposed in this report is to develop mathematical models which will accomplish the purposes No. 1 and No. 2 with simplified assumptions such as linear elastic analysis, limited assessment on component fragility (considering only accident sequences leading to core melt), and simplified safety system

  17. Radiation Safety in Industrial Radiography. Specific Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2013-01-01

    This Safety Guide provides recommendations for ensuring radiation safety in industrial radiography used in non-destructive testing. This includes industrial radiography work that utilizes X ray and gamma sources, both in shielded facilities that have effective engineering controls and in outside shielded facilities using mobile sources. Contents: 1. Introduction; 2. Duties and responsibilities; 3. Safety assessment; 4. Radiation protection programme; 5. Training and qualification; 6. Individual monitoring of workers; 7. Workplace monitoring; 8. Control of radioactive sources; 9. Safety of industrial radiography sources and exposure devices; 10. Radiography in shielded enclosures; 11. Site radiography; 12. Transport of radioactive sources; 13. Emergency preparedness and response; Appendix: IAEA categorization of radioactive sources; Annex I: Example safety assessment; Annex II: Overview of industrial radiography sources and equipment; Annex III: Examples of accidents in industrial radiography

  18. Radiation Safety in Industrial Radiography. Specific Safety Guide

    International Nuclear Information System (INIS)

    2011-01-01

    This Safety Guide provides recommendations for ensuring radiation safety in industrial radiography used in non-destructive testing. This includes industrial radiography work that utilizes X ray and gamma sources, both in shielded facilities that have effective engineering controls and in outside shielded facilities using mobile sources. Contents: 1. Introduction; 2. Duties and responsibilities; 3. Safety assessment; 4. Radiation protection programme; 5. Training and qualification; 6. Individual monitoring of workers; 7. Workplace monitoring; 8. Control of radioactive sources; 9. Safety of industrial radiography sources and exposure devices; 10. Radiography in shielded enclosures; 11. Site radiography; 12. Transport of radioactive sources; 13. Emergency preparedness and response; Appendix: IAEA categorization of radioactive sources; Annex I: Example safety assessment; Annex II: Overview of industrial radiography sources and equipment; Annex III: Examples of accidents in industrial radiography.

  19. Radiation Safety in Industrial Radiography. Specific Safety Guide (French Edition)

    International Nuclear Information System (INIS)

    2013-01-01

    This Safety Guide provides recommendations for ensuring radiation safety in industrial radiography used in non-destructive testing. This includes industrial radiography work that utilizes X ray and gamma sources, both in … shielded facilities that have effective engineering controls and in outside shielded facilities using mobile sources. Contents: 1. Introduction; 2. Duties and responsibilities; 3. Safety assessment; 4. Radiation protection programme; 5. Training and qualification; 6. Individual monitoring of workers; 7. Workplace monitoring; 8. Control of radioactive sources; 9. Safety of industrial radiography sources and exposure devices; 10. Radiography in shielded enclosures; 11. Site radiography; 12. Transport of radioactive sources; 13. Emergency preparedness and response; Appendix: IAEA categorization of radioactive sources; Annex I: Example safety assessment; Annex II: Overview of industrial radiography sources and equipment; Annex III: Examples of accidents in industrial radiography

  20. Radiation Safety in Industrial Radiography. Specific Safety Guide (Arabic Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide provides recommendations for ensuring radiation safety in industrial radiography used in non-destructive testing. This includes industrial radiography work that utilizes X ray and gamma sources, both in shielded facilities that have effective engineering controls and outside shielded facilities using mobile sources. Contents: 1. Introduction; 2. Duties and responsibilities; 3. Safety assessment; 4. Radiation protection programme; 5. Training and qualification; 6. Individual monitoring of workers; 7. Workplace monitoring; 8. Control of radioactive sources; 9. Safety of industrial radiography sources and exposure devices; 10. Radiography in shielded enclosures; 11. Site radiography; 12. Transport of radioactive sources; 13. Emergency preparedness and response; Appendix: IAEA categorization of radioactive sources; Annex I: Example safety assessment; Annex II: Overview of industrial radiography sources and equipment; Annex III: Examples of accidents in industrial radiography.

  1. Seismic qualification of non-safety class equipment whose failure would damage safety class equipment

    International Nuclear Information System (INIS)

    LaSalle, F.R.

    1991-01-01

    Both Code of Federal Regulations, Title 10, Part 50, and US Department of Energy Order 6340.1A have requirements to assess the interaction of non-safety and safety class structures and equipment during a seismic event to maintain the safety function. At the Hanford Site, a cost effective program has been developed to perform the evaluation of non-safety class equipment. Seismic qualification is performed by analysis, test, or upgrading of the equipment to ensure the integrity of safety class structures and equipment. This paper gives a brief overview and synopsis that address design analysis guidelines including applied loading, damping values, component anchorage, allowable loads, and stresses. Test qualification of equipment and walkdown acceptance criteria for heating ampersand ventilation (H ampersand V) ducting, conduit, cable tray, missile zone of influence, as well as energy criteria are presented

  2. Seismic design and performance of nuclear safety related RC structures based on new seismic design principle

    International Nuclear Information System (INIS)

    Murugan, R.; Sivathanu Pillai, C.; Chattopadhyaya, S.; Sundaramurthy, C.

    2011-01-01

    Full text: Seismic design of safety related Reinforced Concrete (RC) structures of Nuclear power plants (NPP) in India as per the present AERB codal procedures tries to ensure predominantly elastic behaviour under OBE so that the features of Nuclear Power Plant (NPP) necessary for continued safe operation are designed to remain functional and prevent accident (collapse) of NPP under SSE for which certain Structures, Systems and Components (SSCs) those are necessary to ensure the capability to shut down the reactor safely, are designed to remain functional. While the seismic design principles of non safety related structures as per Indian code (IS 1893-2002) are ensuring elastic behaviour under DBE and inelastic behaviour under MCE by utilizing ductility and energy dissipation capacity of the structure effectively. The design principle of AERB code is ensuring elastic behaviour under OBE and is not enlightening much inference about the overall structural behaviour under SSE (only ensuring the capability of certain SSCs required for safe shutdown of reactor). Various buildings and structures of Indian Nuclear power plant are classified from the basis of associated safety functions in a descending order in according with their roles in preventions and mitigation of an accident or support functions for prevention. This paper covers a comprehensive seismic analysis and design methodology based on the AERB codal provisions followed for safety related RC structure taking Diesel Generator Building of PFBR as a case study and study and investigates its performance under OBE and SSE by carrying out Non-linear static Pushover analysis. Based on the analysis, observed variations, recommendations are given for getting the desired performance level so as to implement performance based design in the future NPP design

  3. Seismic safety margin research program. Program plan, Revision II

    International Nuclear Information System (INIS)

    Smith, P.D.; Tokarz, F.J.; Bernreuter, D.L.; Cummings, G.E.; Chou, C.K.; Vagliente, V.N.; Johnson, J.J.; Dong, R.G.

    1978-01-01

    The document has been prepared pursuant to the second meeting of the Senior Research Review Group of the Seismic Safety Margin Research Program (SSMRP), which was held on June 15, 16, 1978. The major portion of the material contained in the document is descriptions of specific subtasks to be performed on the SSMRP. This is preceded by a brief discussion of the objective of the SSMRP and the approach to be used. Specific subtasks to be performed in Phase I of the SSMRP are as follows: (1) plant/site selection, (2) seismic input, (3) soil structure interaction, (4) structural building response, (5) structural sub-system response, (6) fragility, (7) system analysis, and (8) Phase II task definition

  4. Modifications to nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA's programme for safety standards for nuclear power plants. It supplements Section 7 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation, which establishes the safety requirements for the modification of nuclear power plants. Reasons for carrying out modifications to nuclear power plants may include: (1) maintaining or strengthening existing safety provisions and thus maintaining consistency with or improving on the current design. (2) recovering from plant faults. (3) improving the thermal performance or increasing the power rating of the plant. (4) increasing the maintainability of the plant, reducing the radiation exposure of personnel or reducing the costs of plant maintenance. And (5) extending the design life of the plant. Most modifications, made on the basis of operating experience, are intended to improve on the design or to improve operational performance and flexibility. Some are rendered necessary by new regulatory requirements, ageing of the plant or obsolescence of equipment. However, the benefits of regularly updating the plant design can be jeopardized if modifications are not kept under rigorous control throughout the lifetime of the plant. The need to reduce costs and improve efficiency, in combination with changes to the structure of the electricity generation sector of the economy in many countries, has led many companies to make changes in the structure of the operating organization for nuclear power plants. Whatever the reason for such organizational changes, consideration should be given to the effects of those changes with the aim of ensuring that they would have no impacts that would compromise the safety of the plant. The objective of this Safety Guide is to provide guidance and recommendations on controlling activities relating to modifications at nuclear power plants in order to reduce risk and to ensure that the configuration of the plant is at all times under

  5. Modifications to nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2007-01-01

    This Safety Guide was prepared under the IAEA's programme for safety standards for nuclear power plants. It supplements Section 7 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation, which establishes the safety requirements for the modification of nuclear power plants. Reasons for carrying out modifications to nuclear power plants may include: (1) maintaining or strengthening existing safety provisions and thus maintaining consistency with or improving on the current design. (2) recovering from plant faults. (3) improving the thermal performance or increasing the power rating of the plant. (4) increasing the maintainability of the plant, reducing the radiation exposure of personnel or reducing the costs of plant maintenance. And (5) extending the design life of the plant. Most modifications, made on the basis of operating experience, are intended to improve on the design or to improve operational performance and flexibility. Some are rendered necessary by new regulatory requirements, ageing of the plant or obsolescence of equipment. However, the benefits of regularly updating the plant design can be jeopardized if modifications are not kept under rigorous control throughout the lifetime of the plant. The need to reduce costs and improve efficiency, in combination with changes to the structure of the electricity generation sector of the economy in many countries, has led many companies to make changes in the structure of the operating organization for nuclear power plants. Whatever the reason for such organizational changes, consideration should be given to the effects of those changes with the aim of ensuring that they would have no impacts that would compromise the safety of the plant. The objective of this Safety Guide is to provide guidance and recommendations on controlling activities relating to modifications at nuclear power plants in order to reduce risk and to ensure that the configuration of the plant is at all times under

  6. Construction for Nuclear Installations. Specific Safety Guide

    International Nuclear Information System (INIS)

    2015-01-01

    This Safety Guide provides recommendations and guidance based on international good practices in the construction of nuclear installations, which will enable construction to proceed with high quality. It can be applied to support the development, implementation and assessment of construction methods and procedures and the identification of good practices for ensuring the quality of the construction to meet the design intent and ensure safety. It will be a useful tool for regulatory bodies, licensees and new entrant countries for nuclear power plants and other nuclear installations

  7. Environment, health and safety guiding principles

    International Nuclear Information System (INIS)

    1997-06-01

    The Canadian Energy Pipeline Association (CEPA) has taken a leadership role in promoting responsible planning, management and work practices that meet the pipeline industry's environment, health and safety objectives. This brochure contains CEPA's environment, health and safety statement. It lists the guiding principles developed and endorsed by CEPA and its member companies in support of protecting the environment and the health and safety of its employees and the public. The 11 CEPA member companies are: Alberta Natural Gas Company Ltd., ATCO Gas Services Ltd., Foothills Pipe Lines Ltd., Interprovincial Pipe Line Inc., NOVA Gas Transmission Limited, TransGas Limited, Trans Mountain Pipe Line Company Ltd., Trans-Northern Pipelines Inc., Trans Quebec and Maritimes Pipeline Inc., and Westcoast Energy Inc

  8. Consideration of vertical seismic response spectrum in nuclear safety review

    International Nuclear Information System (INIS)

    Sun Zaozhan; Huang Bingchen

    2011-01-01

    The basic requirements for civil nuclear installation are introduced in the article. Starting from the basic concept of seismic response spectrum, the authors analyze the site seismic response spectrum and the design seismic response spectrum that desire much consideration. By distinguishing the absolute seismic response spectrum and relative seismic response spectrum, the authors analyze the difference and relationship between the vertical seismic response spectrum and horizontal seismic response spectrum. The authors also bring forward some suggestions for determining the site vertical seismic response spectrum by considering the fact in our country. (authors)

  9. For safety in procurement, follow the guide!

    CERN Multimedia

    HSE Unit

    2014-01-01

    At one time or another, whether as part of a project or for an activity or service, you may find that you have to write a technical specification before placing an order for equipment or machinery. In all cases, when specifying what you need, you must make sure that aspects linked to safety and, in some cases, radiation protection and the protection of the environment, are taken into account in your invitation to tender/price enquiry.   In order to help you with this, the HSE Unit has just published Safety Guideline GS 0-0-1: “27 Key Questions to Ensure that Safety Aspects are Integrated into Invitations to Tender". This guide, available on EDMS under document number 1334815, has been drawn up after the verification of safety aspects of over 300 invitations to tender recently issued by CERN. It collates the most commonly received comments and remarks concerning safety in a question-and-answer format, so you will find plenty of explanations and points to include in your doc...

  10. 41 CFR 128-1.8009 - Review of Seismic Safety Program.

    Science.gov (United States)

    2010-07-01

    ... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false Review of Seismic Safety Program. 128-1.8009 Section 128-1.8009 Public Contracts and Property Management Federal Property Management Regulations System (Continued) DEPARTMENT OF JUSTICE 1-INTRODUCTION 1.80-Seismic Safety Program...

  11. A framework of risk-informed seismic safety evaluation of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Kondo, S.; Sakagami, M.; Hirano, M.; Shiba, M.

    2001-01-01

    A framework of risk-informed seismic design and safety evaluation of nuclear power plants is under consideration in Japan so as to utilize the progress in the seismic probabilistic safety assessment methodology. Issues resolved to introduce this framework are discussed after the concept, evaluation process and characteristics of the framework are described. (author)

  12. Classification of Radioactive Waste. General Safety Guide

    International Nuclear Information System (INIS)

    2009-01-01

    This publication is a revision of an earlier Safety Guide of the same title issued in 1994. It recommends revised waste management strategies that reflect changes in practices and approaches since then. It sets out a classification system for the management of waste prior to disposal and for disposal, driven by long term safety considerations. It includes a number of schemes for classifying radioactive waste that can be used to assist with planning overall national approaches to radioactive waste management and to assist with operational management at facilities. Contents: 1. Introduction; 2. The radioactive waste classification scheme; Appendix: The classification of radioactive waste; Annex I: Evolution of IAEA standards on radioactive waste classification; Annex II: Methods of classification; Annex III: Origin and types of radioactive waste

  13. Classification of Radioactive Waste. General Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-11-15

    This publication is a revision of an earlier Safety Guide of the same title issued in 1994. It recommends revised waste management strategies that reflect changes in practices and approaches since then. It sets out a classification system for the management of waste prior to disposal and for disposal, driven by long term safety considerations. It includes a number of schemes for classifying radioactive waste that can be used to assist with planning overall national approaches to radioactive waste management and to assist with operational management at facilities. Contents: 1. Introduction; 2. The radioactive waste classification scheme; Appendix: The classification of radioactive waste; Annex I: Evolution of IAEA standards on radioactive waste classification; Annex II: Methods of classification; Annex III: Origin and types of radioactive waste.

  14. Seismic Safety Program: Ground motion and structural response

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    In 1964, John A. Blume & Associates Research Division (Blume) began a broad-range structural response program to assist the Nevada Operations Office of the US Atomic Energy Commission (AEC) in ensuring the continued safe conduct of underground nuclear detonation testing at the Nevada Test Site (NTS) and elsewhere. Blume`s long experience in earthquake engineering provided a general basis for the program, but much more specialized knowledge was required for the AEC`s purposes. Over the next 24 years Blume conducted a major research program to provide essential understanding of the detailed nature of the response of structures to dynamic loads such as those imposed by seismic wave propagation. The program`s results have been embodied in a prediction technology which has served to provide reliable advanced knowledge of the probable effects of seismic ground motion on all kinds of structures, for use in earthquake engineering and in building codes as well as for the continuing needs of the US Department of Energy`s Nevada Operations Office (DOE/NV). This report is primarily an accounting of the Blume work, beginning with the setting in 1964 and the perception of the program needs as envisioned by Dr. John A. Blume. Subsequent chapters describe the structural response program in detail and the structural prediction procedures which resulted; the intensive data acquisition program which, as is discussed at some length, relied heavily on the contributions of other consultant-contractors in the DOE/NV Seismic Safety Support Program; laboratory and field studies to provide data on building elements and structures subjected to dynamic loads from sources ranging from testing machines to earthquakes; structural response activities undertaken for testing at the NTS and for off-NTS underground nuclear detonations; and concluding with an account of corollary studies including effects of natural forces and of related studies on building response.

  15. Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

    2012-03-15

    JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

  16. Seismic analysis of control and safety rod drive mechanism

    International Nuclear Information System (INIS)

    Meher Prasad, A.; Jaya, K.P.; Chellapandi, P.; Rajan Babu, V.; Selvaraj, T.

    2003-01-01

    Control rod and its driving mechanism for a Fast Breeder Reactor is to facilitate safe shutdown of the reactor in case of emergency. A theoretical study on the seismic qualification of control and safety rod driving mechanism is carried out. Earthquake excitations under Operational Basis (ORE) and Safe Shutdown condition (SSE) are considered. The time required for the control rod to reach the bottom position in order to shut down the reaction under excited condition is traced out. The maximum displaced positions and extreme stresses in various parts of the system under excitations are evaluated. The system modeled using beam elements. The connections between different parts are modeled through rigid elements. The interaction between various parts are modeled using GAP elements. (author)

  17. Effects of relay chatter in seismic probabilistic safety analysis

    International Nuclear Information System (INIS)

    Reed, J.W.; Shiu, K.K.

    1985-01-01

    In the Zion and Indian Point Probabilistic Safety Studies, relay chatter was dismissed as a credible event and hence was not formally included in the analyses. Although little discussion is given in the Zion and Indian Point PSA documentation concerning the basis for this decision, it has been expressed informally that it was assumed that the operators will be able to reset all relays in a timely manner. Currently, it is the opinion of many professionals that this may be an oversimplification. The three basic areas which must be considered in addressing relay chatter include the fragility of the relays per se, the reliability of the operators to reset the relays and finally the systems response aspects. Each of these areas is reviewed and the implications for seismic PSA are discussed. Finally, recommendations for future research are given

  18. Seismic safety of the LLL plutonium facility (Building 332)

    International Nuclear Information System (INIS)

    Torkarz, F.J.; Shaw, G.

    1980-01-01

    This report states the basis for the Lawrence Livermore Laboratory's assurance to the public that the plutonium operations at the Laboratory pose essentially no risk to anyone's health or safety, either under normal circumstances or in the event of an earthquake or a fire. The report is intended for a general audience, and so for the most part it is not highly technical. It summarizes the steps taken to ensure the seismic safety of the plutonium facility (Bldg. 332). It describes plutonium and its potential hazard and how the facility copes with that hazard. It recounts the geologic investigations and interpretations that led to the design-basis earthquake (DBE) for the Livermore site, and presents a summary analysis of the facility structure in relation to the DBE. An appendix presents a quantitative calculation of the health risk to the public associated with the worst-case hypothetical fire. The document supports the conclusions that the facility will continue to function safely after the maximum earthquake ground motion to which it may be subjected and that there is no evidence of a potential for surface offset under it

  19. Laboratory Safety Guide for Arkansas K-12 Schools.

    Science.gov (United States)

    Arkansas State Dept. of Education, Little Rock.

    This document presents laboratory safety rules for Arkansas K-12 schools which were developed by the Arkansas Science Teachers Association (ASTA) and the Arkansas Department of Education (ADE). Contents include: (1) "Laboratory Safety Guide for Arkansas K-12 Schools"; (2) "Safety Considerations"; (3) "Safety Standards for Science Laboratories";…

  20. Summary report on the Seismic Safety Margins Research Program

    International Nuclear Information System (INIS)

    Cummings, G.E.

    1986-01-01

    The Seismic Safety Margins Research Program (SSMRP) was a program to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. The SSMRP was the first effort to trace seismically induced failure modes in a reactor system down to the individual component level, and to take into account common-cause earthquake-induced failures at the component level. This report summarizes methods and results generated by SSMRP. The SSMRP method makes use of three computer codes, HAZARD, SMACS and SEISIM to calculate ground motion acceleration time histories, structure and component responses and failure, and radioactive release probabilities. To demonstrate the methodology, an analysis was done of the Zion Nuclear Power Plant. The median frequency of core melt was computed to be 3E-5 per year, with upper (90%) and lower (10%) bounds of 8E-4 and 6E-7 per year. The main contribution to risk came from earthquakes about 2 through 4 times the design basis earthquake level. Risk was dominated by structural and inter-building piping failures and loss of off-site power. Sensitivity studies were undertaken to test assumptions and modeling procedures relative to soil-structure interaction effects, feed-and-bleed cooling, and structural failures. Assumptions made could have an order-of-magnitude effect on core melt frequency. Also, guidelines were developed for simplifying the SSMRP method, and importance rankings were generated based on the Zion analysis. 56 refs., 6 figs

  1. Proposal for a seismic facility for reactor safety research

    International Nuclear Information System (INIS)

    Anderson, C.A.; Dove, R.C.; Rhorer, R.L.

    1976-07-01

    Certain problem areas in the seismic analysis and design of nuclear reactors are enumerated and the way in which an experimental program might contribute to each area is examined. The use of seismic simulation testing receives particular attention, especially with regard to the verification of structural response analysis. The importance of scale modeling used in conjunction with seismic simulation is also stressed. The capabilities of existing seismic simulators are summarized, and a proposed facility is described which would considerably extend the ability to conduct, with confidence, confirmatory experiments on the behavior of reactor components when subjected to seismic excitation. Particular applications to gas-cooled and other reactor types are described

  2. Seismic qualification of safety class components in non-reactor nuclear facilities at Hanford site

    International Nuclear Information System (INIS)

    Ocoma, E.C.

    1989-01-01

    This paper presents the methods used during the walkdowns to compile as-built structural information to seismically qualify or verify the seismic adequacy of safety class components in the Plutonium Finishing Plant complex. The Plutonium finishing Plant is a non-reactor nuclear facility built during the 1950's and was designed to the Uniform Building Code criteria for both seismic and wind events. This facility is located at the US Department of Energy Hanford Site near Richland, Washington

  3. Guided Seismic Waves: Possible Diagnostics for Hot Plumes in the Mantle

    Science.gov (United States)

    Evans, J. R.; Julian, B. R.; Foulger, G. R.

    2005-12-01

    Seismic waves potentially provide by far the highest resolution view of the three-dimensional structure of the mantle, and the hope of detecting wave-speed anomalies caused by hot or compositionally buoyant mantle plumes has been a major incentive to the development of tomographic seismic techniques. Seismic tomography is limited, however, by the uneven geographical distribution of earthquakes and seismometers, which can produce artificial tomographic wave-speed anomalies that are difficult to distinguish from real structures in the mantle. An alternate approach may be possible, because hot plumes and possibly some compositional upwellings would have low seismic-wave speeds and would act as efficient waveguides over great depth ranges in the mantle. Plume-guided waves would be little affected by bends or other geometric complexities in the waveguides (analogously to French horns and fiber-optic cables), and their dispersion would make them distinctive on seismograms and would provide information on the size and structure of the waveguide. The main unanswered question is whether guided waves in plumes could be excited sufficiently to be observable. Earthquakes do not occur in the deep mantle, but at least two other possible sources of excitation can be imagined: (1) shallow earthquakes at or near plume-fed hotspots; and (2) coupling of plume-guided waves to seismic body waves near the bottom of the mantle. In the first case, downward-traveling guided waves transformed to seismic body waves at the bottom of the waveguide would have to be detected at teleseismic distances. In the second case, upward-traveling guided waves generated by teleseismic body waves would be detected on seismometers at hotspots. Qualitative reasoning based on considerations of reciprocity suggests that the signals in these two situations should be similar in size and appearance. The focusing of seismic core phases at caustics would amplify plume waves excited by either mechanism (1) or (2) at

  4. Safety guide on fire protection in nuclear power plants

    International Nuclear Information System (INIS)

    1976-01-01

    The purpose of the Safety Guide is to give specific design and operational guidance for protection from fire and explosion in nuclear power plants, based on the general guidance given in the relevant sections of the 'Safety Code of Practice - Design' and the 'Safety Code of Practice - Operation' of the International Atomic Energy Agency. The guide will confine itself to fire protection of safety systems and items important to safety, leaving the non-safety matters of fire protection in nuclear power plants to be decided upon the basis of the various available national and international practices and regulations. (HP) [de

  5. Seismic qualification of equipment in operating nuclear power plants: Unresolved Safety Issue A-46

    International Nuclear Information System (INIS)

    Chang, T.Y.

    1987-02-01

    The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform their intended safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants must be reassessed to determine whether requalification is necessary. The objective of technical studies performed under the Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring qualification to the current criteria that are applied to new plants. This report summarizes the work accomplished on USI A-46. In addition, the collection and review of seismic experience data and existing seismic test data are presented. Staff assessment of work accomplished under USI A-46 leads to the conclusion that the use of seismic experience data provides the most reasonable alternative to current qualification criteria. Consideration of seismic qualification by use of experience data was a specific task in USI A-46. Several other A-46 tasks serve to support the use of an experienced data base. The principal technical finding of USI A-46 is that seismic experience data, supplemented by existing seismic test data, applied in accordance with the guidelines developed, can be used to verify the seismic adequacy of mechanical and electrical equipment in operating nuclear plants. Explicit seismic qualification should be required only if seismic experience data or existing test data on similar components cannot be shown to apply

  6. The roles of the seismic safety and monitoring systems in the PEC fast reactor

    International Nuclear Information System (INIS)

    Masoni, P.; Di Tullio, E.M.; Massa, B.; Martelli, A.; Sano, T.

    1988-01-01

    Two different seismic systems are foreseen in the case of PEC: the seismic safety system, that provides the automatic scram, and the seismic monitoring system. During earthquake, three triaxial seismic switches are triggered if a threshold value of the ground acceleration is exceeded. In this case, the signals from the seismic switches are processed by the safety system (with a 2/3 logic) and the shutdown system is triggered. Peak acceleration is the parameter used by the safety system to quantify the seismic event. This way, however, no information is obtained with regard to earthquake frequency content. Thus, reactor safety is guaranteed by adopting a threshold considerably lower than the Z.P.A. of the Design Basis Earthquake. Furthermore, in the case of significant earthquakes, the seismic motion is measured by about 20 triaxial accelerometers, located both in the free field and on the plant's structures. Data are digitazed and recordered by the seismic monitoring system. This system also elaborates the recordered time-histories providing floor response spectra and compares such spectra to the design values. The above-mentioned elaborations and comparisons are performed in short time for two triaxial measuring positions, thus allowing the Operator to immediately get a more complete information on the seismic event. The complete set of data recorded by the seismic monitoring system also allows the actual dynamic response of the plant to be determined and compared to the design values. On the basis of this comparison the necessary safety analysis can be carried out to verify whether the design limits of the plant were respected: in the positive case the reactor can be restarted. (author)

  7. Seismic Safety Margins Research Program (Phase I). Project VII. Systems analysis specification of computational approach

    International Nuclear Information System (INIS)

    Wall, I.B.; Kaul, M.K.; Post, R.I.; Tagart, S.W. Jr.; Vinson, T.J.

    1979-02-01

    An initial specification is presented of a computation approach for a probabilistic risk assessment model for use in the Seismic Safety Margin Research Program. This model encompasses the whole seismic calculational chain from seismic input through soil-structure interaction, transfer functions to the probability of component failure, integration of these failures into a system model and thereby estimate the probability of a release of radioactive material to the environment. It is intended that the primary use of this model will be in sensitivity studies to assess the potential conservatism of different modeling elements in the chain and to provide guidance on priorities for research in seismic design of nuclear power plants

  8. Seismic structural fragility investigation for the Zion Nuclear Power Plant. Seismic safety margins research program (phase 1)

    International Nuclear Information System (INIS)

    Wesley, D.A.; Hashimoto, P.S.

    1981-10-01

    An evaluation of the seismic capacity of the essential structures for the Zion Nuclear Power Plant in Zion, Illinois, was conducted as part of the Seismic Safety Margins Research Program (SSMRP). The structures included the reactor containment building, the turbine/auxiliary building, and the crib house (intake structure). The evaluation was devoted to seismically induced failures rather than those resulting from combined Loss of Coolant Accident (LOCA) or other extreme load combinations. The seismic loads used in the investigation were based on elastic analyses. The loads for the reactor containment and turbine/auxiliary buildings were developed by Lawrence Livermore Laboratory using time history analyses. The loads used for the crib house were the original seismic design loads developed by Sargent and Lundy. No non-linear seismic analyses were conducted. The seismic capacity of the structures accounted for the actual concrete and steel material properties including the aging of the concrete. Median centered properties were used throughout the evaluation including levels of damping considered appropriate for structures close to collapse as compared to the more conservative values used for design. The inelastic effects were accounted for using ductility modified response spectrum techniques based on system ductility ratios expected for structures near collapse. Sources of both inherent randomness and uncertainties resulting from lack of knowledge or approximations in analytical modelling were considered in developing the dispersion of the structural dynamic characteristics. Coefficients of variation were developed assuming lognormal distributions for all variables. The earthquake levels for many of the seismically induced failure modes are so high as to be considered physically incredible. (author)

  9. Recommended research program for improving seismic safety of light-water nuclear power plants

    International Nuclear Information System (INIS)

    1979-04-01

    Recommendations are presented for research areas concerned with seismic safety. These recommendations are based on an analysis of the answers to a questionnaire which was sent to over 80 persons working in the area of seismic safety of nuclear power plants. In addition to the answers of the 55 questionnaires which were received, the recommendations are based on ideas expressed at a meeting of an ad hoc group of professionals formed by Sandia, review of literature, current research programs, and engineering judgement

  10. Unresolved Safety Issue A-46 - seismic qualification of equipment in operating plants

    International Nuclear Information System (INIS)

    Anderson, N.

    1985-01-01

    Seismic Qualification of Equipment in Operating Plants was designated as an Unresolved Safety Issue (USI) in December, 1980. The USI A-46 program was developed in early 1981 to investigate the adequacy of mechanical and electrical equipment in operating plants to withstand a safe shutdown earthquake. The approach taken was to develop viable, cost effective alternatives to current seismic qualification licensing requirements which could be applied to operating nuclear power plants. The tasks investigated include: (1) identification of seismic sensitive systems and equipment; (2) assessment of adequacy of existing seismic qualification methods; (3) development and assessment of in-situ test procedures to assist in qualification of equipment; (4) seismic qualification of equipment using seismic experience data; and (5) development of methods to generate generic floor response spectra. Progress to date and plans for completion of resolution are reported

  11. Japan's international cooperation programs on seismic safety of nuclear power plants

    International Nuclear Information System (INIS)

    Sanada, Akira

    1997-01-01

    MITI is promoting many international cooperation programs on nuclear safety area. The seismic safety of nuclear power plants (NPPs) is a one of most important cooperation areas. Experts from MITI and related organization join the multilateral cooperation programs carried out by international organization such as IAEA, OECD/NEA etc. MITI is also promoting bilateral cooperation programs such as information exchange meetings, training programs and seminars on nuclear safety with several countries. Concerning to the cooperation programs on seismic safety of NPPs such as information exchange and training, MITI shall continue and expand these programs. (J.P.N.)

  12. Japan`s international cooperation programs on seismic safety of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Sanada, Akira [Agency of Natural Resources and Energy, Tokyo (Japan)

    1997-03-01

    MITI is promoting many international cooperation programs on nuclear safety area. The seismic safety of nuclear power plants (NPPs) is a one of most important cooperation areas. Experts from MITI and related organization join the multilateral cooperation programs carried out by international organization such as IAEA, OECD/NEA etc. MITI is also promoting bilateral cooperation programs such as information exchange meetings, training programs and seminars on nuclear safety with several countries. Concerning to the cooperation programs on seismic safety of NPPs such as information exchange and training, MITI shall continue and expand these programs. (J.P.N.)

  13. Seismic safety review mission for the follow-up of the seismic upgrading of Kozloduy NPP (Units 1-4). Sofia, Bulgaria, 16-20 November 1992

    International Nuclear Information System (INIS)

    David, M.; Shibata, H.; Stevenson, J.D.; Godoy, A.; Gurpinar, A.

    1992-11-01

    A Seismic Safety Review Mission for the follow-up of the design and implementation of the seismic upgrading of Kozloduy NPP was performed in Sofia from 16-20 November 1992. This mission continued the second task of the follow-up activities of the design and implementation of the seismic upgrading (Phases 1 and 2), which is being carried out in Units 1 and 2 of the NPP. Thus the objectives of the mission was to assist the Bulgarian authorities in the technical evaluation of the design tasks defined for Phases 1 and 2 item HB of WANO 6 Month Programme, as follows: anchorage upgrades of low seismic capacity components; list of seismic safety related systems and components; detailed walkdown to assess seismic capacity of components and define priorities for the upgrading; determination of seismic structural capacity of pump house, diesel generator building and turbine building and design of required upgrades; liquefaction potential evaluation. Tabs

  14. Safety design guides for fire protection for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide establishes design requirements to ensure the radiological risk to the public due to fire is acceptable and operating personnel are adequately protected from the hazards of fires. This safety design guide also specifies the safety criteria for fire protection to be applied to mitigate fires and recommends the fire protection program to be established to initiate, coordinate and document the design activities associated with fire protection. The requirements for fire protection outlined in this safety design guide shall be satisfied in the design stage and the change status of the regulatory requirements, code and standards should be traced and incorporated into this safety design guide accordingly. 1 fig., (Author) .new

  15. Seismic simulation and functional performance evaluation of a safety related, seismic category I control room emergency air cleaning system

    International Nuclear Information System (INIS)

    Manley, D.K.; Porco, R.D.; Choi, S.H.

    1985-01-01

    Under a nuclear contract MSA was required to design, manufacture, seismically test and functionally test a complete Safety Related, Seismic Category I, Control Room Emergency Air Cleaning System before shipment to the Yankee Atomic Electric Company, Yankee Nuclear Station in Rowe, Massachusetts. The installation of this system was required to satisfy the NRC requirements of NUREG-0737, Section III, D.3.4, ''Control Room Habitability''. The filter system tested was approximately 3 ft. wide by 8 ft. high by 18 ft. long and weighed an estimated 8300 pounds. It had a design flow rate of 3000 SCFM and contained four stages of filtration - prefilters, upstream and downstream HEPA filters and Type II sideload charcoal adsorber cells. The filter train design followed the guidelines set forth by ANSI/ASME N509-1980. Seismic Category I Qualification Testing consisted of resonance search testing and triaxial random multifrequency testing. In addition to ANSI/ASME N510-1980 testing, triaxial response accelerometers were placed at specific locations on designated prefilters, HEPA filters, charcoal adsorbers and test canisters along with accelerometers at the corresponding filter seal face locations. The purpose of this test was to demonstrate the integrity of the filters, filter seals, and monitor seismic response levels which is directly related to the system's ability to function during a seismic occurrence. The Control Room Emergency Air Cleaning System demonstrated the ability to withstand the maximum postulated earthquake for the plant site by remaining structurally sound and functional

  16. Countermeasures that work : a highway safety countermeasure guide for state highway safety offices : eighth edition : 2015

    Science.gov (United States)

    2015-11-01

    The guide is a basic reference to assist State Highway Safety Offices in selecting effective, evidence- based : countermeasures for traffic safety problem areas. These areas include: : - Alcohol-and Drug-Impaired Driving; : - Seat Belts and Child Res...

  17. Ensuring seismic safety of Blahutovice nuclear power plant

    International Nuclear Information System (INIS)

    Bartak, V.; David, M.; Hrabe, T.; Simunek, P.

    1989-01-01

    The results are presented of the seismic and geological survey of the Blahutovice nuclear power plant site. The variants are discussed of laying foundations and securing earthquake protection of the reactor building. The calculations made show that all variants are suitable with respect to seismic effects because the acceleration of seismic vibrations at the foundation slab level reaches the maximum intensity of 8deg MSK 64. The variant envisaging that the reactor building should be supported on spring insulators with viscous dampers is considered most advanced. (J.B.). 8 figs., 1 tab

  18. IAEA code and safety guides on quality assurance

    International Nuclear Information System (INIS)

    Raisic, N.

    1980-01-01

    In the framework of its programme in safety standards development, the IAEA has recently published a Code of Practice on Quality Assurance for Safety in Nuclear Power Plants. The Code establishes minimum requirements for quality assurance which Member States should use in the context of their own nuclear safety requirements. A series of 10 Safety Guides which describe acceptable methods of implementing the requirements of specific sections of the Code are in preparation. (orig.)

  19. Methodology and results of the seismic probabilistic safety assessment of Krsko nuclear power plant

    International Nuclear Information System (INIS)

    Vermaut, M.K.; Monette, P.; Campbell, R.D.

    1995-01-01

    A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Krsko plant. The methodology adopted is the seismic PSA (Probabilistic Safety Assessment). The Krsko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of the site hazard, calculation of plant structures response including soil-structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Also relay chatter analysis and soil stability studies were performed. The seismic PSA described here is limited to the analysis of CDF (level I PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Krsko NPP but are not further described in this paper. The results of the seismic PSA study indicate that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable to that of most US and Western Europe NPPs. (author)

  20. Instrumentation and control systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. It supplements Safety Standards Series No. NS-R-1: Safety of Nuclear Power Plants: Design (the Requirements for Design), which establishes the design requirements for ensuring the safety of nuclear power plants. This Safety Guide describes how the requirements should be met for instrumentation and control (I and C) systems important to safety. This publication is a revision and combination of two previous Safety Guides: Safety Series Nos 50-SG-D3 and 50-SG-D8, which are superseded by this new Safety Guide. The revision takes account of developments in I and C systems important to safety since the earlier Safety Guides were published in 1980 and 1984, respectively. The objective of this Safety Guide is to provide guidance on the design of I and C systems important to safety in nuclear power plants, including all I and C components, from the sensors allocated to the mechanical systems to the actuated equipment, operator interfaces and auxiliary equipment. This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety. It expands on paragraphs of Ref in the area of I and C systems important to safety. This publication is intended for use primarily by designers of nuclear power plants and also by owners and/or operators and regulators of nuclear power plants. This Safety Guide provides general guidance on I and C systems important to safety which is broadly applicable to many nuclear power plants. More detailed requirements and limitations for safe operation specific to a particular plant type should be established as part of the design process. The present guidance is focused on the design principles for systems important to safety that warrant particular attention, and should be applied to both the design of new I and C systems and the modernization of existing systems. Guidance is provided on how design

  1. Advances in crosshole seismic instrumentation for dam safety monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Anderlini, G.; Anderlini, C. [BC Hydro, Burnaby, BC (Canada); Taylor, R. [RST Instruments Ltd., Coquitlam, BC (Canada)

    2009-07-01

    Since 1996, crosshole shear wave velocity measurements have been performed annually at the WAC Bennett Dam in order to monitor the performance of the dam core and integrity of the 1997 sinkhole repairs. As the testing showed to be responsive to embankment conditions and capable of detecting subtle changes, the testing program was expanded to include the development of an electrical shear wave source capable of carrying out crosshole seismic testing in Mica and Revelstoke Dams over distances of 100 metres and depths of 250 metres. This paper discussed the development and capabilities of the crosshole seismic instrumentation and presented preliminary results obtained during initial testing. Specific topics that were discussed included conventional crosshole seismic equipment; design basics; description of new crosshole seismic equipment; and automated in-situ crosshole seismic system (ACSS) system description and operation. It was concluded that the ACSS and accompanying electrical shear wave source, developed as part of the project, has advanced and improved on traditional crosshole seismic equipment. 7 refs., 9 figs.

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

    International Nuclear Information System (INIS)

    2014-01-01

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

  3. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  4. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  5. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  6. Categorization of Radioactive Sources. Safety Guide

    International Nuclear Information System (INIS)

    2009-01-01

    The IAEA's Statute authorizes the Agency to establish safety standards to protect health and minimize danger to life and property - standards which the IAEA must use in its own operations, and which a State can apply by means of its regulatory provisions for nuclear and radiation safety. A comprehensive body of safety standards under regular review, together with the IAEA's assistance in their application, has become a key element in a global safety regime. In the mid-1990s, a major overhaul of the IAEA's safety standards programme was initiated, with a revised oversight committee structure and a systematic approach to updating the entire corpus of standards. The new standards that have resulted are of a high calibre and reflect best practices in Member States. With the assistance of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its safety standards. Safety standards are only effective, however, if they are properly applied in practice. The IAEA's safety services - which range in scope from engineering safety, operational safety, and radiation, transport and waste safety to regulatory matters and safety culture in organizations - assist Member States in applying the standards and appraise their effectiveness. These safety services enable valuable insights to be shared and I continue to urge all Member States to make use of them. Regulating nuclear and radiation safety is a national responsibility, and many Member States have decided to adopt the IAEA's safety standards for use in their national regulations. For the Contracting Parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions. The standards are also applied by designers, manufacturers and operators around the world to enhance nuclear and radiation safety in power generation, medicine, industry, agriculture, research and education

  7. Seismic safety margins research program. Project I SONGS 1 AFWS Project

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Smith, P.D.; Dong, R.G.; Bernreuter, D.L.; Bohn, M.P.; Cummings, G.E.; Wells, J.E.

    1981-01-01

    The seismic qualification requirements of auxiliary feedwater systems (AFWS) of Pressurized Water Reactors (PWR) were developed over a number of years. These are formalized in the publication General Design Criteria (Appendix A to 10CFR50). The full recognition of the system as an engineered safety feature did not occur until publication of the Standard Review Plan (1975). Efforts to determine how to backfit seismic requirements to earlier plants has been undertaken primarily in the Systematic Evaluation Program (SEP) for a limited number of operating reactors. Nuclear Reactor Research (RES) and NRR have requested LLNL to perform a probabilistic study on the AFWS of San Onofre Nuclear Generating Station (SONGS) Unit 1 utilizing the tools developed by the Seismic Safety Margins Research Program (SSMRP). The main objectives of this project are to: identify the weak links of AFWS; compare the failure probabilities of SONGS 1 and Zion 1 AFWS: and compare the seismic responses due to different input spectra and design values

  8. Seismic analysis of the safety related piping and PCLS of the WWER-440 NPP

    International Nuclear Information System (INIS)

    Berkovski, A.M.; Kostarev, V.V.; Schukin, A.J.; Boiadjiev, Z.; Kostov, M.

    2001-01-01

    This paper presents the results of seismic analysis of Safety Related Piping Systems of the typical WWER-440 NPP. The methodology of this analysis is based on WANO Terms of Reference and ASME BPVC. The different possibilities for seismic upgrading of Primary Coolant Loop System (PCLS) were considered. The first one is increasing of hydraulic snubber units and the second way is installation of limited number of High Viscous Dampers (HVD). (author)

  9. 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

  10. 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

  11. The U.S. Nuclear Regulatory Commission seismic safety research program

    International Nuclear Information System (INIS)

    Kenneally, R.M.; Guzy, D.J.; Murphy, A.J.

    1988-01-01

    The seismic safety research program sponsored by the U.S. Nuclear Regulatory Commission is directed toward improving the evaluation of potential earthquake effects on nuclear power plant operations. The research has been divided into three major program areas: earth sciences, seismic design margins, and fragilities and response. A major thrust of this research is to assess plant behavior for seismic events more severe and less probable than those considered in design. However, there is also research aimed at improving the evaluation of earthquake input and plant response at plant design levels

  12. Seismic Safety Margins Research Program (Phase I). Project IV. Structural building response; Structural Building Response Review

    International Nuclear Information System (INIS)

    Healey, J.J.; Wu, S.T.; Murga, M.

    1980-02-01

    As part of the Phase I effort of the Seismic Safety Margins Research Program (SSMRP) being performed by the University of California Lawrence Livermore Laboratory for the US Nuclear Regulatory Commission, the basic objective of Subtask IV.1 (Structural Building Response Review) is to review and summarize current methods and data pertaining to seismic response calculations particularly as they relate to the objectives of the SSMRP. This material forms one component in the development of the overall computational methodology involving state of the art computations including explicit consideration of uncertainty and aimed at ultimately deriving estimates of the probability of radioactive releases due to seismic effects on nuclear power plant facilities

  13. Safety, Health, and Environmental Auditing A Practical Guide

    CERN Document Server

    Pain, Simon Watson

    2010-01-01

    A practical guide to environmental, safety, and occupational health audits. It allows organizations and business to avoid expensive external auditors and retain the knowledge and learning 'in-house'. It allows any competent manager or safety/environmental officer to undertake in-house audits in a competent and reproducible fashion.

  14. Efforts toward enhancing seismic safety at Kashiwazaki Kariwa Nuclear Power Station

    International Nuclear Information System (INIS)

    Yamashita, Kazuhiko

    2009-01-01

    It has been two years since the Niigata-ken Chuetsu-oki Earthquake (NCOE) occurred in 2007. The earthquake brought a major disaster for Kashiwazaki, Kariwa, and the neighboring areas. First of all, we would like to give condolences to people in the devastated area and to pray for the immediate recovery. Our Kashiwazaki Kariwa Nuclear Power Station located in the same area was naturally caught up in the earthquake. The station was hit by a big tremor more than its intensity assumed to be valid at the station design stage. In spite of unexpected tremor, preventive functions for the station safety worked as expected as it designed. Critical facilities designed as high seismic class were not damaged, though considerable damages were seen in outside-facilities designed as low seismic class. We currently make efforts to inspect and recover damages. While we carefully carry out inspection and assessment to make sure the station integrity, we are also going forward restoration as well as construction for seismic safety enhancement in turn. This report introduces details of the following accounts, these are an outline of guidelines for seismic design evaluation that was revised in 2006, a situation at Kashiwazaki Kariwa Nuclear Power Station in the aftermath of the earthquake, and efforts toward enhancing seismic safety that the Tokyo Electric Power Company (TEPCO) has made since the seismic disaster, and our approach to evaluation of facility integrity. (author)

  15. Subsystem response review. Seismic safety margins research program

    International Nuclear Information System (INIS)

    Kennedy, R.P.; Campbell, R.D.; Wesley, D.A.; Kamil, H.; Gantayat, A.; Vasudevan, R.

    1981-07-01

    A study was conducted to document the state of the art in seismic qualification of nuclear power plant components and subsystems by analysis and testing and to identify the sources and magnitude of the uncertainties associated with analysis and testing methods. The uncertainties are defined in probabilistic terms for use in probabilistic seismic risk studies. Recommendations are made for the most appropriate subsystem response analysis methods to minimize response uncertainties. Additional studies, to further quantify testing uncertainties, are identified. Although the general effect of non-linearities on subsystem response is discussed, recommendations and conclusions are based principally on linear elastic analysis and testing models. (author)

  16. Technical guidelines for the seismic safety re-evaluation at Eastern European NPPs

    International Nuclear Information System (INIS)

    Godoy, A.R.; Guerpinar, A.

    2001-01-01

    The paper describes one of the outcomes of the Engineering Safety Review Services (ESRS) that the IAEA provides as an element of the Agency's national, regional and interregional technical assistance and co-operation programmes and other extrabudgetary programmes to assess the safety of nuclear facilities. This refers to the establishment of detailed guidelines for conducting the seismic safety re-evaluation of existing nuclear power plants in Eastern European countries in line with updated criteria and current international practice. (author)

  17. Decommissioning of nuclear fuel cycle facilities. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    The objective of this Safety Guide is to provide guidance to regulatory bodies and operating organizations on planning and provision for the safe management of the decommissioning of non-reactor nuclear fuel cycle facilities. While the basic safety considerations for the decommissioning of nuclear fuel cycle facilities are similar to those for nuclear power plants, there are important differences, notably in the design and operating parameters for the facilities, the type of radioactive material and the support systems available. It is the objective of this Safety Guide to provide guidance for the shutdown and eventual decommissioning of such facilities, their individual characteristics being taken into account

  18. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  19. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  20. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  1. Probabilistic safety analysis procedures guide, Sections 8-12. Volume 2, Rev. 1

    International Nuclear Information System (INIS)

    McCann, M.; Reed, J.; Ruger, C.; Shiu, K.; Teichmann, T.; Unione, A.; Youngblood, R.

    1985-08-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of issues affecting reactor safety. The first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. This second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study. For external events, more explicit guidance is given

  2. Probabilistic safety analysis procedures guide. Sections 1-7 and appendices. Volume 1, Revision 1

    International Nuclear Information System (INIS)

    Bari, R.A.; Buslik, A.J.; Cho, N.Z.

    1985-08-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. It will be revised as comments are received, and as experience is gained from its use. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of issues affecting reactor safety. This first volume of the guide describes the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant (i.e., intrinsic to plant operation) and from loss of off-site electric power. The scope includes human reliability analysis, a determination of the importance of various core damage accident sequences, and an explicit treatment and display of uncertainties for key accident sequences. The second volume deals with the treatment of the so-called external events including seismic disturbances, fires, floods, etc. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance). This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are valuable for regulatory decision making. For internal events, methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study. For external events, more explicit guidance is given

  3. The IAEA International Seismic Safety Centre and IAEA safety standards for site evaluation and design of NPPs

    International Nuclear Information System (INIS)

    Godoy, A.; Sollogoub, P; )

    2009-01-01

    This presentation covers the following topics: 'Lessons learned' from the occurrence of strong natural events, (tsunamis, earthquakes, hurricanes, etc.) The International Seismic Safety Centre as a global focal point for the nuclear engineering community in those fields. A need for international cooperation, openness and transparency – Sharing of experience

  4. Motorcycle Safety Education. A Curriculum Guide.

    Science.gov (United States)

    Ohio State Board of Education, Columbus.

    This curriculum guide was produced to assist instructors of educational programs for novice motorcycle operators, automobile drivers, and all highway users. An introductory section discusses program implementation concerns, such as public relations, legal considerations, scheduling, staff, students, facilities, motorcycles, insurance, financial…

  5. IAEA codes and guides for safety of nuclear power plants

    International Nuclear Information System (INIS)

    Raisic, N.

    1980-01-01

    The objectives and scope of the Agency's programme of nuclear safety standards are described and the role of these documents in regulation of nuclear power im Member States is discussed. For each of the five areas of safety standards development, i.e. siting, design, operation, quality assurance and governmental organization, a set of principles underlying requirements and recommendations contained in the Code of Practice and Safety Guides will be presented. Safety Guides in each of the five areas will be reviewed in respect of the scope and content. A consideration will be given to the future development of the safety standards and to the revision and updating of the published documents. (orig./RW)

  6. Approaches that use seismic hazard results to address topics of nuclear power plant seismic safety, with application to the Charleston earthquake issue

    International Nuclear Information System (INIS)

    Sewell, R.T.; McGuire, R.K.; Toro, G.R.; Stepp, J.C.; Cornell, C.A.

    1990-01-01

    Plant seismic safety indicators include seismic hazard at the SSE (safe shut-down earthquake) acceleration, seismic margin, reliability against core damage, and reliability against offsite consequences. This work examines the key role of hazard analysis in evaluating these indicators and in making rational decisions regarding plant safety. The paper outlines approaches that use seismic hazard results as a basis for plant seismic safety evaluation and applies one of these approaches to the Charleston earthquake issue. This approach compares seismic hazard results that account for the Charleston tectonic interpretation, using the EPRI-Seismicity Owners Group (SOG) methodology, with hazard results that are consistent with historical tectonic interpretations accepted in regulation. Based on hazard results for a set of 21 eastern U.S. nuclear power plant sites, the comparison shows that no systematic 'plant-to-plant' increase in hazard accompanies the Charleston hypothesis; differences in mean hazards for the two interpretations are generally insignificant relative to current uncertainties in seismic hazard. (orig.)

  7. Comparison of evaluation guidelines for life-safety seismic hazards

    International Nuclear Information System (INIS)

    Wyllie, L.A.; Love, R.J.

    1989-01-01

    The guidelines presented in Design Evaluation guidelines for Department of Energy Facilities Subjected to natural Phenomena Hazards (UCRL 15910 Draft; May 1989) include evaluation criteria for existing Department of Energy buildings subjected to earthquakes. These criteria were developed at the Lawrence Livermore National Laboratory for use in both the seismic design of new structures and the evaluation of existing structures. ATC-14: Evaluating The Seismic Resistance of Existing Buildings developed by the Applied Technology Council, consists of guidelines and criteria for identifying the buildings or building components that present unacceptable risk to human lives. This paper compares and contrasts the two evaluation guidelines for existing buildings using a prototype building as an example. The prototype building is a seven story, concrete shear wall building assuming a General Use Occupancy

  8. Armenian nuclear power plant: US NRC assistance programme for seismic upgrade and safety analysis

    International Nuclear Information System (INIS)

    Simos, N.; Perkins, K.; Jo, J.; Carew, J.; Ramsey, J.

    2003-01-01

    This paper summarizes the U.S. Nuclear Regulatory Commission's (US NRC) technical support program activities associated with the Armenian Nuclear Power Plant (ANPP) safety upgrade. The US NRC program, integrated within the overall IAEA-led initiative for safety re-evaluation of the WWER plants, has as its main thrust the technical support to the Armenian Nuclear Regulatory Authority (ANRA) through close collaboration with the scientific staff at Brookhaven National Laboratory (BNL). Several major technical areas of support to ANRA form the basis of the NRC program. These include the seismic re-evaluation and upgrade of the ANPP, safety evaluation of critical systems, and the generation of the Safety Analysis Report (SAR). Specifically, the seismic re-evaluation of the ANPP is part of a broader activity that involves the re-assessment of the seismic hazard at the site, the identification of the Safe Shutdown Equipment at the plant and the evaluation of their seismic capacity, the detailed modeling and analysis of the critical facilities at ANPP, and the generation of the Floor Response Spectra (FRS). Based on the new spectra that incorporate all new findings (hazard, site soil, structure, etc.), the overall capacity of the main structures and the seismic capacity of the critical systems are being re-evaluated. In addition, analyses of critical safe shutdown systems and safe shutdown processes are being performed to ensure both the capabilities of the operating systems and the enhancement of safety due to system upgrades. At present, one of the principal goals of the US NRC's regulatory assistance activities with ANRA is enhancing ANRA's regulatory oversight of high-priority safety issues (both generic and plant-specific) associated with operation of the ANPP. As such, assisting ANRA in understanding and assessing plant-specific seismic and other safety issues associated with the ANPP is a high priority given the ANPP's being located in a seismically active area

  9. Seismic investigations of the HDR Safety Program. Summary report

    International Nuclear Information System (INIS)

    Malcher, L.; Schrammel, D.; Steinhilber, H.; Kot, C.A.

    1994-08-01

    The primary objective of the seismic investigations, performed at the HDR facility in Kahl/Main, FRG was to validate calculational methods for the seismic evaluation of nuclear-reactor systems, using experimental data from an actual nuclear plant. Using eccentric mass shaker excitation the HDR soil/structure system was tested to incipient failure, exhibiting highly nonlinear response and demonstrating that structures not seismically designed can sustain loads equivalent to a design basin earthquake (DBE). Load transmission from the structure to piping/equipment indicated significant response amplifications and shifts to higher frequencies, while the response of tanks/vessels depended mainly on their support conditions. The evaluation of various piping support configurations demonstrated that proper system design (for a given spectrum) rather than number of supports or system stiffness is important to limiting pipe greens. Piping at loads exceeding the DBE eightfold still had significant margins and failure is improbable inspite of multiple support failures. The mean value for pipe damping, even under extreme loads, was found to be about 4%. Comparison of linear and nonlinear computational results with piping response measurements showed that predictions have a wide scatter and do not necessarily yield conservative responses underpredicting, in particular, peak support forces. For the soil/structure system the quality of the predictions did not depend so much on the complexity of the modeling, but rather on whether the model captured the salient features and nonlinearities of the system

  10. Safety guide of the Tandar accelerator

    International Nuclear Information System (INIS)

    1987-01-01

    The safety standards that the installations of the Tandar accelerator have to comply with are presented here. In order to maintain the safety, the knowledge and the accomplishment of these standards are mandatory for all persons. The risks of external irradiation and of contamination are pointed out. The risks at the Tandar are: the calibration standards used at the premises and the irradiation produced by the activity of the accelerator, which can be primary, secondary, induced or X rays. The identification of the different areas of installation are given with their corresponding classification; the rules concerning the manipulation of radioactive materials and the movement of persons in areas of reglamentary access are established. Finally conventional safety and rules for evacuation and fires are presented. (M.E.L.) [es

  11. Seismic test for safety evaluation of low level radioactive wastes containers

    International Nuclear Information System (INIS)

    Ohoka, Makoto; Horikiri, Morito

    1998-08-01

    Seismic safety of three-piled container system used in Tokai reprocessing center was confirmed by seismic test and computational analysis. Two types of container were evaluated, for low level noninflammable radioactive solid wastes, and for used filters wrapped by large plastic bags. Seismic integrity of three-piled containers was confirmed by evaluating response characteristics such as acceleration and displacement under the design earthquake condition S1, which is the maximum earthquake expected at the stored site during the storage time. Computational dynamic analysis was also performed, and several conclusions described below were made. (1) Response characteristics of the bottom board and the side board were different. The number of pile did not affect the response characteristics of the bottom board of each container. They behaved as a rigid body. (2) The response of the side board was larger than that of the bottom board. (3) The response depended on the direction in each board, either side or bottom. The response acceleration became larger to the seismic wave perpendicular to the plane which has the entrance for fork lift and the radioactive warning mark. (4) The maximum horizontal response displacement under the S1 seismic wave was approximately 10 mm. It is so small that it does not affect the seismic safety. (5) The stoppers to prevent fall down had no influence to the response acceleration. (6) There was no fall down to the S1 seismic wave and 2 times of S1 seismic wave, which was the maximum input condition of the test. (7) The response of the bottom board of the containers, which are main elements of fall down, had good agreements both in the test and in the computational analysis. (author)

  12. Guide for understanding and evaluation of safety culture

    International Nuclear Information System (INIS)

    2008-01-01

    This report was the guide of understanding and evaluation of safety culture. Operator's activities for enhancement of safety culture in nuclear installations became an object of safety regulation in the management system. Evaluation of operator's activities (including top management's involvement) to prevent degradation of safety culture and organization climate in daily works needed understanding of safety culture and diversity of operator's activities. This guide was prepared to check indications of degradation of safety culture and organization climate in operator's activities in daily works and encourage operator's activities to enhance safety culture improvement and good practice. Comprehensive evaluation of operator's activities to prevent degradation of safety culture and organization climate would be performed from the standpoints of 14 safety culture elements such as top management commitment, clear plan and implementation of upper manager, measures to avoid wrong decision making, questioning attitude, reporting culture, good communications, accountability and openness, compliance, learning system, activities to prevent accidents or incidents beforehand, self-assessment or third party evaluation, work management, change management and attitudes/motivation. Element-wise examples and targets for evaluation were attached with evaluation check tables. (T. Tanaka)

  13. Safety design guide for pipe rupture protection for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for pipe rupture protection identifies high-energy systems in which pipe ruptures must be postulated to occur, as well as systems that must be protected from the dynamic effects of such ruptures. Dynamic effects considered in this SDG consist of pipe whip (including missiles generated by pipe ruptures, if any) and jet impingement, Requirements for protection against the dynamic effects of a postulated pipe rupture and method of protection of essential structures, systems and components are specified for these effects. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 5 refs. (Author) .new

  14. Differences in safety margins between nuclear and conventional design standards with regards to seismic hazard definition and design criteria

    International Nuclear Information System (INIS)

    Elgohary, M.; Saudy, A.; Orbovic, N.; Dejan, D.

    2006-01-01

    With the surging interest in new build nuclear all over the world and a permanent interest in earthquake resistance of nuclear plants, there is a need to quantify the safety margins in nuclear buildings design in comparison to conventional buildings in order to increase the public confidence in the safety of nuclear power plants. Nuclear (CAN3-N289 series) and conventional (NBCC 2005) seismic standards have different approaches regarding the design of civil structures. The origin of the differences lays in the safety philosophy behind the seismic nuclear and conventional standards. Conventional seismic codes contain the minimal requirement destined primarily to safeguard against major structural failure and loss of life. It doesn't limit damage to a certain acceptable degree or maintain function. Nuclear seismic code requires that structures, systems and components important to safety, withstand the effects of earthquakes. The requirement states that for equipment important to safety, both integrity and functionality should be ascertained. The seismic hazard is generally defined on the basis of the annual probability of exceedence (return period). There is a major difference on the return period and the confidence level for design earthquakes between the conventional and the nuclear seismic standards. The seismic design criteria of conventional structures are based on the use of Force Modification Factors to take into account the energy dissipation by incursion in non-elastic domain and the reserve of strength. The use of such factors to lower intentionally the seismic input is consistent with the safety philosophy of the conventional seismic standard which is the 'non collapse' rather than the integrity and/or the operability of the structures or components. Nuclear seismic standard requires that the structure remain in the elastic domain; energy dissipation by incursion in non-elastic domain is not allowed for design basis earthquake conditions. This is

  15. Seismic analysis of safety class 1 incinerator glovebox in building 232-Z 200 W Area

    International Nuclear Information System (INIS)

    Ocoma, E.C.

    1994-09-01

    This report documents the seismic evaluation for the existing safety class 1 incinerator glovebox in 232Z Building. The glovebox is no longer in use and most of the internal mechanical equipment have been removed. However, the insulation firebricks are still in the glovebox for proper disposal

  16. Safety design guides for environmental qualification for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide describes the safety philosophy and requirements for the environmental qualification of safety related systems and components for CANDU 9. The environmental qualification program identifies the equipments to be qualified and conditions to be used for qualification and provides comprehensive set of documentation to ensure that the qualification is complete and can be maintained for the life of the plant. A summary of the system, components and structures requiring environmental qualification is provided in the table for the guidance of the system design, and this table will be subject to change or confirmation by the environmental qualification program. Also, plant ares subject to harsh environment is provided in the figure. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 1 tab., 5 figs. (Author) .new

  17. Decommissioning of Medical, Industrial and Research Facilities. Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of medical, industrial and research facilities where radioactive materials and sources are produced, received, used and stored. It is intended to provide guidance to national authorities and operating organizations, particularly to those in developing countries (as such facilities are predominant in these countries), for the planning and safe management of the decommissioning of such facilities. The Safety Guide has been prepared through a series of Consultants meetings and a Technical Committee meeting

  18. Decommissioning of medical, industrial and research facilities. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of medical, industrial and research facilities where radioactive materials and sources are produced, received, used and stored. It is intended to provide guidance to national authorities and operating organizations, particularly to those in developing countries (as such facilities are predominant in these countries), for the planning and safe management of the decommissioning of such facilities. The Safety Guide has been prepared through a series of Consultants meetings and a Technical Committee meeting

  19. Regulatory Control of Radiation Sources. Safety Guide (Arabic Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide is intended to assist States in implementing the requirements established in Safety Standards Series No. GS-R-1, Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety, for a national regulatory infrastructure to regulate any practice involving radiation sources in medicine, industry, research, agriculture and education. The Safety Guide provides advice on the legislative basis for establishing regulatory bodies, including the effective independence of the regulatory body. It also provides guidance on implementing the functions and activities of regulatory bodies: the development of regulations and guides on radiation safety; implementation of a system for notification and authorization; carrying out regulatory inspections; taking necessary enforcement actions; and investigating accidents and circumstances potentially giving rise to accidents. The various aspects relating to the regulatory control of consumer products are explained, including justification, optimization of exposure, safety assessment and authorization. Guidance is also provided on the organization and staffing of regulatory bodies. Contents: 1. Introduction; 2. Legal framework for a regulatory infrastructure; 3. Principal functions and activities of the regulatory body; 4. Regulatory control of the supply of consumer products; 5. Functions of the regulatory body shared with other governmental agencies; 6. Organization and staffing of the regulatory body; 7. Documentation of the functions and activities of the regulatory body; 8. Support services; 9. Quality management for the regulatory system.

  20. Subsystem response analysis for the Seismic Safety Margins Research Program

    International Nuclear Information System (INIS)

    Chuang, T.Y.

    1981-01-01

    A review of the state-of-the-art of seismic qualification methods of subsystem has been completed. This task assesses the accuracy of seismic analysis techniques to predict dynamic response, and also identifies and quantifies sources of random and modeling undertainty in subsystem response determination. The subsystem has been classified as two categories according to the nature of support: multiply supported subsystems (e.g., piping systems) and singly supported subsystems (e.g., pumps, turbines, electrical control panels, etc.). The mutliply supported piping systems are analyzed by multisupport input time history method. The input motions are the responses of major structures. The dynamic models of the subsystems identified by the event/fault tree are created. The responses calculated by multisupport input time history method are consistent with the fragility parameters. These responses are also coordinated with the event/fault tree description. The subsystem responses are then evaluated against the fragility curves of components and systems and incorporated in the event/fault tree analysis. (orig./HP)

  1. Predisposal management of high level radioactive waste. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Radioactive waste is generated in the generation of electricity in nuclear power plants and in the use of radioactive material in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized. The principles and requirements that govern the safety of the management of radioactive waste are presented in 'The Principles of Radioactive Waste Management', 'Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety' and 'Predisposal Management of Radioactive Waste, Including Decommissioning'. The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established in Refs for the predisposal management of HLW. This Safety Guide applies to the predisposal management of HLW. For liquid HLW arising from the reprocessing of spent fuel the recommendations of this Safety Guide apply from when liquid waste from the first extraction process is collected for storage and subsequent processing. Recommendations and guidance on the storage of spent fuel, whether or not declared as waste, subsequent to its removal from the storage facility of a reactor are provided in Refs. For spent fuel declared as waste this Safety Guide applies to all activities subsequent to its removal from the storage facility of a reactor and prior to its disposal. Requirements pertaining to the transport of spent fuel, whether or not declared as waste, and of all forms of HLW are established. This Safety Guide provides recommendations on the safety aspects of managing HLW, including the planning, design, construction, commissioning, operation and decommissioning of equipment or facilities for the predisposal management of HLW. It addresses the following elements: (a) The characterization and processing (i.e. pretreatment

  2. Non-compliance with agrochemical safety guides and associated ...

    African Journals Online (AJOL)

    Although several occupational health hazards are associated with farming, cocoa farmers could be exposed to more health hazards through use of agrochemicals. The objective of this study was to analyze the effect of non-compliance with agrochemical safety guides on health risks of farmers. The data were collected from ...

  3. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

  4. End of mission report on seismic safety review mission for Belene NPP site

    International Nuclear Information System (INIS)

    Gurpinar, A.; Mohammadioun, B.; Schneider, H.; Serva, L.

    1995-01-01

    Upon the invitation of the Bulgarian government through the Committee for the Peaceful Uses of Atomic Energy and within the framework of the implementation of the Technical Cooperation project BUL/9/012 related to site and seismic of NPPs, a mission visited Sofia 3 - 7 July 1995. The mission constituted a follow-up of the interim review of subjects related to tectonic stability and seismic hazard characterization of the site which was performed in September 1993. The main objective of the mission was the final review of the subjects already reviewed in September 1993 as well as issues related to geotechnical engineering and foundation safety. The main terms of reference of the present mission was to verify the implementation of the recommendations of the Site Safety Review Mission of June 1990. This document gives findings on geology-tectonics, seismology and foundation safety. In the end conclusions and recommendations of the mission are presented

  5. Criticality Safety in the Handling of Fissile Material. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-05-15

    This Safety Guide provides guidance and recommendations on how to meet the relevant requirements for ensuring subcriticality when dealing with fissile material and for planning the response to criticality accidents. The guidance and recommendations are applicable to both regulatory bodies and operating organizations. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences of this if it were to occur. The Safety Guide makes recommendations on how to ensure subcriticality in systems involving fissile materials during normal operation, anticipated operational occurrences, and, in the case of accident conditions, within design basis accidents, from initial design through commissioning, operation, and decommissioning and disposal.

  6. Radiation Safety of Gamma, Electron and X Ray Irradiation Facilities. Specific Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    The objective of this Safety Guide is to provide recommendations on how to meet the requirements of the BSS with regard to irradiation facilities. This Safety Guide provides specific, practical recommendations on the safe design and operation of gamma, electron and X ray irradiators for use by operating organizations and the designers of these facilities, and by regulatory bodies. SCOPE. The facilities considered in this publication include five types of irradiator, whether operated on a commercial basis or for research and development purposes. This publication is concerned with radiation safety issues and not with the uses of irradiators, nor does it cover the irradiation of product or its quality management. The five types of irradiator are: - Panoramic dry source storage irradiators; - Underwater irradiators, in which both the source and the product being irradiated are under water; - Panoramic wet source storage irradiators; - Electron beam irradiation facilities, in which irradiation is performed in an area that is potentially accessible to personnel, but that is kept inaccessible during the irradiation process; - X ray irradiation facilities, in which irradiation is performed in an area that is potentially accessible to personnel, but that is kept inaccessible during the irradiation process. Consideration of non-radiation-related risks and of the benefits resulting from the operation of irradiators is outside the scope of this Safety Guide. The practices of radiotherapy and radiography are also outside the scope of this Safety Guide. Category I gamma irradiators (i.e. 'self-shielded' irradiators) are outside the scope of this Safety Guide

  7. Assessment of NPP safety taking into account seismic and engineering-geological factors

    International Nuclear Information System (INIS)

    Yakovlev, E.A.

    1990-01-01

    Consideration is given to the problem of probabilistic analysis of NPP safety with account of risk of destructive effect of earthquakes and the danger of accidental geological processes (diapirism, karst etc.) under NPP operation. It is shown that account of seismic and engineering-geological (engineering-seismological) risk factors in probabilistic analysis of safety enables to perform anticipatory analysis of behaviour of principle plant objects and to improve safety of their operation by revealing the most unstable elements of geotechnical system forming the main contribution to the total NPP risk

  8. Seismic Safety Margins Research Program. Phase 1. Project V. Structural sub-system response: subsystem response review

    International Nuclear Information System (INIS)

    Fogelquist, J.; Kaul, M.K.; Koppe, R.; Tagart, S.W. Jr.; Thailer, H.; Uffer, R.

    1980-03-01

    This project is directed toward a portion of the Seismic Safety Margins Research Program which includes one link in the seismic methodology chain. The link addressed here is the structural subsystem dynamic response which consists of those components and systems whose behavior is often determined decoupled from the major structural response. Typically the mathematical model utilized for the major structural response will include only the mass effects of the subsystem and the main model is used to produce the support motion inputs for subsystem seismic qualification. The main questions addressed in this report have to do with the seismic response uncertainty of safety-related components or equipment whose seismic qualification is performed by (a) analysis, (b) tests, or (c) combinations of analysis and tests, and where the seismic input is assumed to have no uncertainty

  9. The role of IAEA in the seismic assessment and upgrading of existing NPPs. Seismic safety of nuclear power plants in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Guerpinar, A; Godoy, A [International Atomic Energy Agency, Vienna (IAEA). Div. of Nuclear Installation Safety

    1997-03-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on `Benchmark study for the seismic analysis and testing of WWER type nuclear power plants`. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  10. The role of IAEA in the seismic assessment and upgrading of existing NPPs. Seismic safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Guerpinar, A.; Godoy, A.; . Div. of Nuclear Installation Safety)

    1997-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  11. Safety design guides for grouping and separation for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    This safety design guide for grouping and separation describes the philosophy of physical and functional separation for systems, structures and components in CANDU 9 plants and provides the requirements for the implementation of the philosophy in the detailed plant design. The separation of the safety systems is to ensure that common cause events and functional interconnections between systems do not impair the capability to perform the required safety functions for accident conditions. The separation requirements are also applied to the design by grouping the plant systems into two basic groups. Group 1 includes the power production systems and Group 2 includes the safety related systems required for the mitigation of serious process failure. The Group 2 is further separated into subgroups to ensure that events that could cause failure of a special safety system in one subgroup can be mitigated by the other subgroup. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 6 figs. (Author) .new.

  12. Safety design guides for grouping and separation for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for grouping and separation describes the philosophy of physical and functional separation for systems, structures and components in CANDU 9 plants and provides the requirements for the implementation of the philosophy in the detailed plant design. The separation of the safety systems is to ensure that common cause events and functional interconnections between systems do not impair the capability to perform the required safety functions for accident conditions. The separation requirements are also applied to the design by grouping the plant systems into two basic groups. Group 1 includes the power production systems and Group 2 includes the safety related systems required for the mitigation of serious process failure. The Group 2 is further separated into subgroups to ensure that events that could cause failure of a special safety system in one subgroup can be mitigated by the other subgroup. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 6 figs. (Author) .new

  13. Re-assessment of seismic loads in conjunction with periodic safety review

    International Nuclear Information System (INIS)

    Jonczyk, Josef

    2002-01-01

    The objective of this paper is the fundamental consideration of a safeguard-aim-oriented approach for use in the re-assessment of seismic events with regard to the periodic safety review (PSR) of nuclear power plants (NPP). The re-assessment aspects of site-specific design earthquakes (DEQ), specially the procedure for seismic hazard analysis, will not, however, be considered in detail here. The proposed assessment concept clearly presents a general approach for safety assessments. The approach is based on a successive screening review of components that are considered sufficiently earthquake-resistant. In this respect, the principle of maximum practical application of the design documentation has been considered in the re-assessment process. On the other hand, the safeguard-aim-oriented evaluation will also be applied with regard to whether the requirements of the safety regulations are fulfilled with respect to the safety goals. The review in conjunction with PSR does not, however, attempt to perform this under all technical aspects. Moreover, it is possible to make extensive use of experimental knowledge and engineering judgement with regard to the structural capacity behaviour in case of a seismic event. Compared with design procedures, however, this proposed approach differs from the one applied in licensing procedures, in which such assessment freedom will not usually be exhausted. (author)

  14. Seismic qualification of equipment in operating nuclear power plants. Unresolved safety issue A-46, draft report for comment

    International Nuclear Information System (INIS)

    Chang, T.Y.

    1985-08-01

    The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform their intended safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants should be reassessed to determine whether requalification is necessary. The objective of technical studies performed under the Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring qualification to the current criteria that are applied to new plants. This report summarizes the work accomplished on USI A-46 by the Nuclear Regulatory Commission staff and its contractors, Idaho National Engineering Laboratory, Southwest Research Institute, Brookhaven National Laboratory, and Lawrence Livermore National Laboratory. In addition, the collection and review of seismic experience data by the Seismic Qualification Utility Group and the review and recommendations of a group of seismic consultants, the Senior Seismic Review Advisory Panel, are presented. Staff assessment of work accomplished under USI A-46 leads to the conclusion that the use of seismic experience data provides the most reasonable alternative to current qualification criteria. Consideration of seismic qualification by use of experience data was a specific task in USI A-46. Several other A-46 tasks serve to support the use of an experience data base

  15. Methods used to seismically upgrade. The safety related components of Belgian plants

    International Nuclear Information System (INIS)

    Lafaille, J.P.

    1993-01-01

    Belgian nuclear power amounts to about 6,000 MW, generated by seven plants that started operation as early as 1967. The latest plant started in 1985. Some of these plants were designed with no seismic requirements whatsoever. Even for those that had seismic requirements at the design stage, seismic demand was raised after design had been frozen (late during construction or at the 10 years revision). As a consequence all the plants had to undergo, to a variable extent, a seismic reevaluation and/or backfitting. Civil structures were concerned as well as electro-mechanical equipment and piping systems. The present paper deals with the mechanical aspect of the problem (equipment and piping). In order to minimize hardware modifications, advanced analytical techniques were used throughout the process, starting with the elaboration of a site specific spectrum, and using a full soil-structure interaction in order to get as 'realistic' as possible floor response spectra. In some instances, non linear elasto-plastic time history analysis was performed on piping-systems in order to qualify them without hardware modifications. In other cases a 'Load Coefficient Method' was used. Sometimes stresses or displacements taken from the original stress reports and scaled by comparison of applicable spectra, allowed to assess the seismic validity of the system under investigation. Seismic acceptability of installed active equipment is more difficult to demonstrate, as this is usually done by testing. This problem is a generic issue in the US, identified under the label USI-A-46 (Unresolved Safety Issue). It is treated by. a group of Utilities (SQUG = Seismic Qualification Utilities Group). The Belgian Utility is member of that group since 1985. The application of this program is starting in the US. SQUG methodology has been applied to three Belgian plants starting in 1988 and is now completed. The required fixes are being implemented. Experience gained in the process has been applied

  16. Seismic performance assessment of base-isolated safety-related nuclear structures

    Science.gov (United States)

    Huang, Y.-N.; Whittaker, A.S.; Luco, N.

    2010-01-01

    Seismic or base isolation is a proven technology for reducing the effects of earthquake shaking on buildings, bridges and infrastructure. The benefit of base isolation has been presented in terms of reduced accelerations and drifts on superstructure components but never quantified in terms of either a percentage reduction in seismic loss (or percentage increase in safety) or the probability of an unacceptable performance. Herein, we quantify the benefits of base isolation in terms of increased safety (or smaller loss) by comparing the safety of a sample conventional and base-isolated nuclear power plant (NPP) located in the Eastern U.S. Scenario- and time-based assessments are performed using a new methodology. Three base isolation systems are considered, namely, (1) Friction Pendulum??? bearings, (2) lead-rubber bearings and (3) low-damping rubber bearings together with linear viscous dampers. Unacceptable performance is defined by the failure of key secondary systems because these systems represent much of the investment in a new build power plant and ensure the safe operation of the plant. For the scenario-based assessments, the probability of unacceptable performance is computed for an earthquake with a magnitude of 5.3 at a distance 7.5 km from the plant. For the time-based assessments, the annual frequency of unacceptable performance is computed considering all potential earthquakes that may occur. For both assessments, the implementation of base isolation reduces the probability of unacceptable performance by approximately four orders of magnitude for the same NPP superstructure and secondary systems. The increase in NPP construction cost associated with the installation of seismic isolators can be offset by substantially reducing the required seismic strength of secondary components and systems and potentially eliminating the need to seismically qualify many secondary components and systems. ?? 2010 John Wiley & Sons, Ltd.

  17. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  18. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  19. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  20. Seismic analysis for safety related structures of 900MWe PWR NPP

    International Nuclear Information System (INIS)

    Liu Wei

    2002-01-01

    Nuclear Power Plant aseismic design becomes more and more important in China due to the fact that China is a country where earthquakes occur frequently and most of plants arc unavoidably located in seismic regions. Therefore, Chinese nuclear safety authority and organizations have worked out a series of regulations and codes related to NPP anti-seismic design taking account of local conditions. The author presents here an example of structural anti-seismic design of 90GM We PWR NPP which is comprised of: ground motion input, including the principles for ground motion determination and time history generation; soil and upper-structure modelling, presenting modeling procedures and typical models of safety related buildings such as Reactor Building, Nuclear Auxiliary Building and Fuel Building; soil-structure interaction analysis; and in-structure response analysis and floor response spectrum generation. With this example, the author intends to give an overview of Chinese practice in NPP structure anti-seismic design such as the main procedures to be followed and the codes and regulations to be respected. (author)

  1. Safety inspection guide, Mod III (a systematic approach to conducting a safety inspection)

    International Nuclear Information System (INIS)

    Davidson, J.E.

    1977-06-01

    This guide was developed as a comprehensive/systematic approach to the problem of performing a safety inspection. Five basic sections (categories) are considered in the guide: physical work place; machines/mechanical equipment; hazardous materials/processes/environments; energy sources; and management hazard . control factors. The basic concept is that one starts evaluating hazard potentials from the physical work place and continues considering other elements as they are added to the physical work place. This approach provides a better understanding of the interfaces of each section to the entire group. The guide is supported by an Area Safety Inspection Result form to record defects or conditions found, the evaluation (best estimate) of the urgency or priority for correcting deficiencies or areas of noncompliance, and the status of corrective action. Additionally, the guide serves as an educational tool in accident prevention for supervisors and employees

  2. Seismic Safety Margins Research Program. Phase I. Interim definition of terms

    International Nuclear Information System (INIS)

    Smith, P.D.; Dong, R.G.

    1980-01-01

    This report documents interim definitions of terms in the Seismic Safety Margins Research Program (SSMRP). Intent is to establish a common-based terminology integral to the probabilistic methods that predict more realistically the behavior of nuclear power plants during an earthquake. These definitions are a response to a request by the Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards at its meeting held November 15-16, 1979

  3. 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

  4. Major structural response methods used in the seismic safety margins research program

    International Nuclear Information System (INIS)

    Chou, C.K.; Lo, T.; Vagliente, V.

    1979-01-01

    In order to evaluate the conservatisms in present nuclear power plant seismic safety requirements, a probabilistic based systems model is being developed. This model will also be used to develop improved requirements. In Phase I of the Seismic Safety Margins Research Program (SSMRP), this methodology will be developed for a specific nuclear power plant and used to perform probabilistic sensitivity studies to gain engineering insights into seismic safety requirements. Random variables in the structural response analysis area, or parameters which cause uncertainty in the response, are discussed and classified into three categories; i.e., material properties, structural dynamic characteristics and related modeling techniques, and analytical methods. The sensitivity studies are grouped into two categories; deterministic and probabilistic. In a system analysis, transfer functions in simple form are needed since there are too many responses which have to be calculated in a Monte Carlo simulation to use the usual straightforward calculation approach. Therefore, the development of these simple transfer functions is one of the important tasks in SSMRP. Simplified as well as classical transfer functions are discussed

  5. Report on the seismic safety examination of nuclear facilities based on the 1995 Hyogoken-Nanbu earthquake

    International Nuclear Information System (INIS)

    2001-01-01

    Just after the Hyogoken-Nanbu Earthquake occurred, Nuclear Safety Commission of Japan established a committee to examine the validity or related guidelines on the seismic design to be used for the safety examination. After the 8 months study, the committee confirmed that the validity of guidelines regulating the seismic design of nuclear facilities is not impaired even though on the basis of the Hyogoken-Nanbu earthquake. This report is the outline of the Committee's study results. (author)

  6. Seismic safety review mission to assist in the evaluation of the design of seismic upgrading for Kozloduy NPP. Sofia, Bulgaria, 19-23 October 1992

    International Nuclear Information System (INIS)

    Ma, D.; Prato, C.; Godoy, A.

    1992-10-01

    A seismic Safety Review Mission to assist in the evaluation of the design of seismic upgrading for Kozloduy NPP was performed in Sofia from 19-23 October 1992. The objectives of the mission were to assist the Bulgarian authorities in: the evaluation of the floor response spectra of the main buildings of units 1-4 at Kozloduy NPP, calculated for the new defined seismic parameters at site (Review Level Earthquake - RLE); the evaluation of the remedial and strengthening measures proposed for the seismic upgrading of the pump house and diesel generator buildings to the new defined RLE. This mission completed the scope of previous IAEA mission - BUL/9/012-18b - (see Report 3262) performed from 3-7 August 1992, with regard to tasks which were not evaluated at that time because they had not been finished. 2 tabs

  7. Research on Safety Factor of Dam Slope of High Embankment Dam under Seismic Condition

    Directory of Open Access Journals (Sweden)

    Li Bin

    2015-01-01

    Full Text Available With the constant development of construction technology of embankment dam, the constructed embankment dam becomes higher and higher, and the embankment dam with its height over 200m will always adopt the current design criteria of embankment dam only suitable for the construction of embankment dam lower than 200m in height. So the design criteria of high embankment dam shall be improved. We shall calculate the stability and safety factors of dam slope of high embankment dam under different dam height, slope ratio and different seismic intensity based on ratio of safety margin, and clarify the change rules of stability and safety factors of dam slope of high embankment dam with its height over 200m. We calculate the ratio of safety margin of traditional and reliable method by taking the stable, allowable and reliability index 4.2 of dam slope of high embankment dam with its height over 200m as the standard value, and conduct linear regression for both. As a result, the conditions, where 1.3 is considered as the stability and safety factors of dam slope of high embankment dam with its height over 200m under seismic condition and 4.2 as the allowable and reliability index, are under the same risk control level.

  8. Probabilistic seismic safety study of an existing nuclear power plant

    International Nuclear Information System (INIS)

    Kennedy, R.P.; Cornell, C.A.; Kaplan, S.; Perla, H.F.

    1980-01-01

    This study was conducted as part of an overall safety study of the Oyster Creek nuclear power plant. The earthquake hazard was considered as an initiating event that could result in radioactive release from the site as a result of core melt. The probability of earthquake initiated releases were compared with the probability of releases due to other initiating events. Three steps are necessary to evaluate the probability of earthquake initiated core melt. (1) estimate the ground motion (peak ground acceleration) and uncertainty in this estimate as functions of annual probability of occurrence; (2) estimate the conditional probability of failure and its uncertainty for structures, equipment, piping, controls, etc., as functions of ground acceleration; and (3) combine these estimates to obtain probabilities of earthquake induced failure and uncertainties in such estimates to be used in event trees, system models, and fault trees for evaluating the probability of earthquake induced core melt. This paper concentrates on the first two steps with emphasis on step 2. The major difference between the work presented and previous papers is the development and use of uncertainty estimates for both the ground motion probability estimates and the conditional probability of failure estimates. (orig.)

  9. Seismic Safety Margins Research Program: Phase II program plan (FY 83-FY 84)

    International Nuclear Information System (INIS)

    Bohn, M.P.; Bernreuter, D.L.; Cover, L.E.; Johnson, J.J.; Shieh, L.C.; Shukla, S.N.; Wells, J.E.

    1982-01-01

    The Seismic Safety Margins Research Program (SSMRP) is an NRC-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its goal is to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from a commercial nuclear power plant. The analysis procedure is based upon a state-of-the-art evaluation of the current seismic analysis and design process and explicitly includes the uncertainties inherent in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. As currently planned, the SSMRP will be completed in September, 1984. This document presents the program plan for work to be done during the remainder of the program. In Phase I of the SSMRP, the necessary tools (both computer codes and data bases) for performing a detailed seismic risk analysis were identified and developed. Demonstration calculations were performed on the Zion Nuclear Power Plant. In the remainder of the program (Phase II) work will be concentrated on developing a simplified SSMRP methodology for routine probabilistic risk assessments, quantitative validation of the tools developed and application of the simplified methodology to a Boiling Water Reactor. (The Zion plant is a pressurized water reactor.) In addition, considerable effort will be devoted to making the codes and data bases easily accessible to the public

  10. Proceedings of the Specialist Meeting on the Seismic Probabilistic Safety Assessment of Nuclear Facilities

    International Nuclear Information System (INIS)

    2007-01-01

    The main objectives of the Meeting were to review recent advances in the methodology of Seismic Probabilistic Safety Assessment (SPSA), to discuss practical applications, to review the current state of the art, and to identify methodological issues where further research would be beneficial in enhancing the usefulness of the methodology. Applications of the Seismic Margin Assessment methodology (SMA), a methodology related to SPSA, were also discussed. One specific objective was to compare the situation today with the situation at the time of the 1999 Tokyo workshop, and to develop a set of findings and recommendations that would update those from that earlier workshop. There was a consensus at the Specialists Meeting that SPSA is now in widespread use throughout the nuclear-power industry worldwide, by the operating nuclear power plants (NPPs) themselves, by the various national regulatory agencies, and by the designers of new NPPs. It was also widely agreed that it can systematically accomplish several very important objectives; specifically, it can contribute: - To understanding the seismic risk arising from NPPs. - To understanding the safety significance of seismic design shortfalls. - To prioritizing seismic safety improvements. - To evaluating and improving seismic regulations. - To modifying the seismic regulatory/licensing basis of an individual NPP. Compared to the situation in 1999, when the first Workshop was held in Tokyo, there have been significant expansions in the use of SPSA in many different areas. Some countries provided detailed information on their regulatory framework for using seismic PSA. Many other countries also provided some information in their papers as background for conducting SPSA. During the Meeting, a small number of important weaknesses in SPSA methodology were identified. None of these are new, all having been widely recognized for many years. However, for some of the weaknesses, extensive discussions during the Meeting provided

  11. Proceedings of the Specialist Meeting on the Seismic Probabilistic Safety Assessment of Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-11-14

    The main objectives of the Meeting were to review recent advances in the methodology of Seismic Probabilistic Safety Assessment (SPSA), to discuss practical applications, to review the current state of the art, and to identify methodological issues where further research would be beneficial in enhancing the usefulness of the methodology. Applications of the Seismic Margin Assessment methodology (SMA), a methodology related to SPSA, were also discussed. One specific objective was to compare the situation today with the situation at the time of the 1999 Tokyo workshop, and to develop a set of findings and recommendations that would update those from that earlier workshop. There was a consensus at the Specialists Meeting that SPSA is now in widespread use throughout the nuclear-power industry worldwide, by the operating nuclear power plants (NPPs) themselves, by the various national regulatory agencies, and by the designers of new NPPs. It was also widely agreed that it can systematically accomplish several very important objectives; specifically, it can contribute: - To understanding the seismic risk arising from NPPs. - To understanding the safety significance of seismic design shortfalls. - To prioritizing seismic safety improvements. - To evaluating and improving seismic regulations. - To modifying the seismic regulatory/licensing basis of an individual NPP. Compared to the situation in 1999, when the first Workshop was held in Tokyo, there have been significant expansions in the use of SPSA in many different areas. Some countries provided detailed information on their regulatory framework for using seismic PSA. Many other countries also provided some information in their papers as background for conducting SPSA. During the Meeting, a small number of important weaknesses in SPSA methodology were identified. None of these are new, all having been widely recognized for many years. However, for some of the weaknesses, extensive discussions during the Meeting provided

  12. Seismic PSA implementation standards by AESJ and the utilization of the advanced safety examination guideline for seismic design for nuclear power plant

    International Nuclear Information System (INIS)

    Ebisawa, Katsumi; Hibino, Kenta

    2008-01-01

    The Advanced Safety Examination Guideline for Seismic Design for Nuclear Power Plant (the advanced safety examination guideline) was worked out on September 19, 2006. In this paper, a summary of the method of probability theory in the advanced safety examination guideline and the Seismic PSA Implementation Standards is stated. On utilization of the probability theory for the advanced safety examination guideline, the uncertainty resulting from the process of the decision of the basic design earthquake ground motion (Ss) is stated to be considered using the proper method. The references of the extra probability for evaluation of earthquake hazard and combination of the working load and the earthquake load are stated. Definition, evaluation method and effort to lower the 'residual risks', and relation between the residual risks and the extra probability of Ss are described. A summary of the earthquake-resistant design for nuclear power facilities is explained by the old guideline. (S.Y.)

  13. Safety goals for seismic and tsunami risks: Lessons learned from the Fukushima Daiichi disaster

    International Nuclear Information System (INIS)

    Saji, Genn

    2014-01-01

    Highlights: • Reviewed why the Fukushima disaster was not anticipated among seismologists. • Reviewed Fukushima Daiichi's preparedness against the earthquake and tsunami. • There was a large “cliff edge” in radiological consequences from the design basis tsunami. • By including earthquakes as an “external event” resulted in insufficient “defense in depths”. • Proposes a new probabilistic seismic and tsunami safety goal be developed. - Abstract: This paper first reviews why the potential occurrence of the Tohoku-oki earthquake with momentum magnitude M w of 9.0 earthquake was not anticipated by Japanese seismologists, and to clarify our limitations in predicting rare but severe earthquakes at our current knowledge in the field of geosciences. Although there was a large volume of historical records related to earthquakes and tsunamis, generally this data infer high plate coupling in regions where earthquakes were known to have already occurred, with only partial or even no coupling from the Japan Trench to a point approximately midway between the trench and the coastline—precisely the region where the 2011 Tohoku-Oki earthquake occurred. This phenomenon has been explained as a “silent earthquake” or a fault creep as observed at the San Andreas Faults in the US. Considering the large uncertainties in seismic events, nuclear power plants should be conservatively designed with adequate safety margins. TEPCO's preparedness against seismic and tsunami hazards were reviewed in order to clarify why the established safety margin was not sufficient during the Fukushima Daiichi. It was found that the plant incorporated the necessary safety margins against seismic oscillation however, there was a large “cliff edge” in which the radiological consequences surged by several orders of magnitude from the design basis tsunami. Since the tsunami's height was greater than the ground level of the turbine hall, a large amount of the tsunami

  14. Safety goals for seismic and tsunami risks: Lessons learned from the Fukushima Daiichi disaster

    Energy Technology Data Exchange (ETDEWEB)

    Saji, Genn, E-mail: sajig@bd5.so-net.ne.jp

    2014-12-15

    Highlights: • Reviewed why the Fukushima disaster was not anticipated among seismologists. • Reviewed Fukushima Daiichi's preparedness against the earthquake and tsunami. • There was a large “cliff edge” in radiological consequences from the design basis tsunami. • By including earthquakes as an “external event” resulted in insufficient “defense in depths”. • Proposes a new probabilistic seismic and tsunami safety goal be developed. - Abstract: This paper first reviews why the potential occurrence of the Tohoku-oki earthquake with momentum magnitude M{sub w} of 9.0 earthquake was not anticipated by Japanese seismologists, and to clarify our limitations in predicting rare but severe earthquakes at our current knowledge in the field of geosciences. Although there was a large volume of historical records related to earthquakes and tsunamis, generally this data infer high plate coupling in regions where earthquakes were known to have already occurred, with only partial or even no coupling from the Japan Trench to a point approximately midway between the trench and the coastline—precisely the region where the 2011 Tohoku-Oki earthquake occurred. This phenomenon has been explained as a “silent earthquake” or a fault creep as observed at the San Andreas Faults in the US. Considering the large uncertainties in seismic events, nuclear power plants should be conservatively designed with adequate safety margins. TEPCO's preparedness against seismic and tsunami hazards were reviewed in order to clarify why the established safety margin was not sufficient during the Fukushima Daiichi. It was found that the plant incorporated the necessary safety margins against seismic oscillation however, there was a large “cliff edge” in which the radiological consequences surged by several orders of magnitude from the design basis tsunami. Since the tsunami's height was greater than the ground level of the turbine hall, a large amount of the

  15. Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants. Specific Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    The objective of this Safety Guide is to provide recommendations for meeting the IAEA safety requirements in performing or managing a level 2 probabilistic safety assessment (PSA) project for a nuclear power plant; thus it complements the Safety Guide on level 1 PSA. One of the aims of this Safety Guide is to promote a standard framework, standard terms and a standard set of documents for level 2 PSAs to facilitate regulatory and external peer review of their results. It describes all elements of the level 2 PSA that need to be carried out if the starting point is a fully comprehensive level 1 PSA. Contents: 1. Introduction; 2. PSA project management and organization; 3. Identification of design aspects important to severe accidents and acquisition of information; 4. Interface with level 1 PSA: Grouping of sequences; 5. Accident progression and containment analysis; 6. Source terms for severe accidents; 7. Documentation of the analysis: Presentation and interpretation of results; 8. Use and applications of the PSA; Annex I: Example of a typical schedule for a level 2 PSA; Annex II: Computer codes for simulation of severe accidents; Annex III: Sample outline of documentation for a level 2 PSA study.

  16. 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.

  17. 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.

  18. Decision making with epistemic uncertainty under safety constraints: An application to seismic design

    Science.gov (United States)

    Veneziano, D.; Agarwal, A.; Karaca, E.

    2009-01-01

    The problem of accounting for epistemic uncertainty in risk management decisions is conceptually straightforward, but is riddled with practical difficulties. Simple approximations are often used whereby future variations in epistemic uncertainty are ignored or worst-case scenarios are postulated. These strategies tend to produce sub-optimal decisions. We develop a general framework based on Bayesian decision theory and exemplify it for the case of seismic design of buildings. When temporal fluctuations of the epistemic uncertainties and regulatory safety constraints are included, the optimal level of seismic protection exceeds the normative level at the time of construction. Optimal Bayesian decisions do not depend on the aleatory or epistemic nature of the uncertainties, but only on the total (epistemic plus aleatory) uncertainty and how that total uncertainty varies randomly during the lifetime of the project. ?? 2009 Elsevier Ltd. All rights reserved.

  19. Seismic safety review mission Almaty WWR 10 MW research reactor Almaty, Kazakhstan. Final report

    International Nuclear Information System (INIS)

    Gurpinar, A.; Slemmons, D.B.; David, M.; Masopust, R.

    1995-06-01

    On the request of the government of Kazakhstan and within the scope of the TC project KAZ/0/004, a seismic safety review mission was conducted in Almaty, 8-19 May 1995 for the WWR 10 Mw research reactor. This review followed the fact finding mission which visited Almaty in November 1993 together with an INSARR mission. At that time some information regarding the seismotectonic setting of the site as well as the seismic capacity of the facility was obtained. This document presents the results of further work carried out on both the issues. It discusses technical session findings on geology, seismology, structures and equipments. In the end conclusions and recommendations of the mission are given. 4 refs, figs, tabs, 18 photos

  20. Radiation Safety in Industrial Radiography. Specific Safety Guide (French Edition); Surete radiologique en radiographie industrielle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-05-15

    This Safety Guide provides recommendations for ensuring radiation safety in industrial radiography used in non-destructive testing. This includes industrial radiography work that utilizes X ray and gamma sources, both in Horizontal-Ellipsis shielded facilities that have effective engineering controls and in outside shielded facilities using mobile sources. Contents: 1. Introduction; 2. Duties and responsibilities; 3. Safety assessment; 4. Radiation protection programme; 5. Training and qualification; 6. Individual monitoring of workers; 7. Workplace monitoring; 8. Control of radioactive sources; 9. Safety of industrial radiography sources and exposure devices; 10. Radiography in shielded enclosures; 11. Site radiography; 12. Transport of radioactive sources; 13. Emergency preparedness and response; Appendix: IAEA categorization of radioactive sources; Annex I: Example safety assessment; Annex II: Overview of industrial radiography sources and equipment; Annex III: Examples of accidents in industrial radiography.

  1. Radiation Safety in Industrial Radiography. Specific Safety Guide (Spanish Edition); Seguridad radiologica en la radiografia industrial

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-12-15

    This Safety Guide provides recommendations for ensuring radiation safety in industrial radiography used in non-destructive testing. This includes industrial radiography work that utilizes X ray and gamma sources, both in shielded facilities that have effective engineering controls and in outside shielded facilities using mobile sources. Contents: 1. Introduction; 2. Duties and responsibilities; 3. Safety assessment; 4. Radiation protection programme; 5. Training and qualification; 6. Individual monitoring of workers; 7. Workplace monitoring; 8. Control of radioactive sources; 9. Safety of industrial radiography sources and exposure devices; 10. Radiography in shielded enclosures; 11. Site radiography; 12. Transport of radioactive sources; 13. Emergency preparedness and response; Appendix: IAEA categorization of radioactive sources; Annex I: Example safety assessment; Annex II: Overview of industrial radiography sources and equipment; Annex III: Examples of accidents in industrial radiography.

  2. The Management System for Nuclear Installations Safety Guide

    International Nuclear Information System (INIS)

    2009-01-01

    This Safety Guide is applicable throughout the lifetime of a nuclear installation, including any subsequent period of institutional control, until there is no significant residual radiation hazard. For a nuclear installation, the lifetime includes site evaluation, design, construction, commissioning, operation and decommissioning. These stages in the lifetime of a nuclear installation may overlap. This Safety Guide may be applied to nuclear installations in the following ways: (a)To support the development, implementation, assessment and improvement of the management system of those organizations responsible for research, site evaluation, design, construction, commissioning, operation and decommissioning of a nuclear installation; (b)As an aid in the assessment by the regulatory body of the adequacy of the management system of a nuclear installation; (c)To assist an organization in specifying to a supplier, via contractual documentation, any specific element that should be included within the supplier's management system for the supply of products. This Safety Guide follows the structure of the Safety Requirements publication on The Management System for Facilities and Activities, whereby: (a)Section 2 provides recommendations on implementing the management system, including recommendations relating to safety culture, grading and documentation. (b)Section 3 provides recommendations on the responsibilities of senior management for the development and implementation of an effective management system. (c)Section 4 provides recommendations on resource management, including guidance on human resources, infrastructure and the working environment. (d)Section 5 provides recommendations on how the processes of the installation can be specified and developed, including recommendations on some generic processes of the management system. (e)Section 6 provides recommendations on the measurement, assessment and improvement of the management system of a nuclear installation. (f

  3. The Management System for Nuclear Installations. Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This Safety Guide is applicable throughout the lifetime of a nuclear installation, including any subsequent period of institutional control, until there is no significant residual radiation hazard. For a nuclear installation, the lifetime includes site evaluation, design, construction, commissioning, operation and decommissioning. These stages in the lifetime of a nuclear installation may overlap. This Safety Guide may be applied to nuclear installations in the following ways: (a) To support the development, implementation, assessment and improvement of the management system of those organizations responsible for research, site evaluation, design, construction, commissioning, operation and decommissioning of a nuclear installation; (b) As an aid in the assessment by the regulatory body of the adequacy of the management system of a nuclear installation; (c) To assist an organization in specifying to a supplier, via contractual documentation, any specific element that should be included within the supplier's management system for the supply of products. This Safety Guide follows the structure of the Safety Requirements publication on The Management System for Facilities and Activities, whereby: (a) Section 2 provides recommendations on implementing the management system, including recommendations relating to safety culture, grading and documentation. (b) Section 3 provides recommendations on the responsibilities of senior management for the development and implementation of an effective management system. (c) Section 4 provides recommendations on resource management, including guidance on human resources, infrastructure and the working environment. (d) Section 5 provides recommendations on how the processes of the installation can be specified and developed, including recommendations on some generic processes of the management system. (e) Section 6 provides recommendations on the measurement, assessment and improvement of the management system of a nuclear

  4. The safety evaluation guide for laboratories and plants a tool for enhancing safety

    International Nuclear Information System (INIS)

    Lhomme, Veronique; Daubard, Jean-Paul

    2013-01-01

    The Institute for Radioprotection and Nuclear Safety (IRSN) acts as technical support for the French government Authorities competent in nuclear safety and radiation protection for civil and defence activities. In this frame, the Institute's performs safety assessments of the safety cases submitted by operators to these Authorities for each stage in the life cycle of a nuclear facility, including dismantling operations, which is subjected to a licensing procedure. In the fuel cycle field, this concerns a large variety of facilities. Very often, depending on facilities and on safety cases, safety assessment to be performed is multidisciplinary and involves the supervisor in charge of the facility and several safety experts, particularly to cover the whole set of risks (criticality, exposure to radiation, fire, handling, containment, human and organisational factors...) encountered during facility's operations. Taking these into account, and in order to formalize the assessment process of the fuel cycle facilities, laboratories, irradiators, particle accelerators, under-decommissioning reactors and radioactive waste management, the 'Plants, Laboratories, Transports and Waste Safety' Division of IRSN has developed an internal guide, as a tool: - To present the methodological framework, and possible specificities, for the assessment according to the 'Defence in Depth Concept' (Part 1); - To provide key questions associated to the necessary contradictory technical review of the safety cases (Part 2); - To capitalise on experience on the basis of technical examples (coming from incident reports, previous safety assessments...) demonstrating the questioning (Part 3). The guide is divided in chapters, each dedicated to a type of risk (dissemination of radioactive material, external or internal exposure from ionising radiation, criticality, radiolysis mechanisms, handling operations, earthquake, human or organisational factors...) or to a type

  5. Implementation guidelines for seismic PSA

    International Nuclear Information System (INIS)

    Coman, Ovidiu; Samaddar, Sujit; Hibino, Kenta; )

    2014-01-01

    The presentation was devoted to development of guidelines for implementation of a seismic PSA. If successful, these guidelines can close an important gap. ASME/ANS PRA standards and the related IAEA Safety Guide (IAEA NS-G-2.13) describe capability requirements for seismic PSA in order to support risk-informed applications. However, practical guidance on how to meet these requirements is limited. Such guidelines could significantly contribute to improving risk-informed safety demonstration, safety management and decision making. Extensions of this effort to further PSA areas, particularly to PSA for other external hazards, can enhance risk-informed applications

  6. Regulatory Guide 1.122: Development of floor design response spectra for seismic design of floor-supported equipment or components

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    ''Reactor Site Criteria,'' requires, in part, that safety-related structures, systems, and components remain functional in the event of a Safe Shutdown Earthquake (SSE). It specifies the use of a suitable dynamic analysis as one method of ensuring that the structures, systems, and components can withstand the seismic loads. Similarly, paragraph (a)(2) of Section VI of the same appendix requires, in part, that the structures, systems, and components necessary for continued operation without undue risk to the health and safety of the public remain functional in the event of an Operating Basis Earthquake (OBE). Again, the use of suitable dynamic analysis is specified as one method of ensuring that the structures, systems, and components can withstand the seismic loads. This guide describes methods acceptable to the NRC staff for developing two horizontal and one vertical floor design response spectra at various floors or other equipment-support locations of interest from the time-history motions resulting from the dynamic analysis of the supporting structure. These floor design response spectra are needed for the dynamic analysis of the systems or equipment supported at various locations of the supporting structure

  7. A cost summary applicable to seismic construction and maintenance of nuclear safety related piping

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-01-01

    This paper presents a summary of costs applicable to nuclear power plant piping for an earthquake defined as 0.2 SSE-PGA as a function of three eras of initial construction: 1967--1974, 1974--1981 and 1981--1990. Costs have been presented for both new construction and maintenance in operating plants using both the original PSAR-FSAR design criteria and current SRP requirements. It is recommended that the cost information contained in this report be considered in evaluating the cost benefit relationships associated with current and proposed future changes in seismic design procedures applicable to safety-related piping systems

  8. Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Ebisawa, K.; Abe, K.; Muramatsu, K.; Itoh, M.; Kohno, K.; Tanaka, T.

    1994-01-01

    This paper presents a method for evaluating 'response factors' of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design resonse to actual response. This method has the following characteristic features: (1) The components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components. This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups. (orig.)

  9. Structure soil structure interaction effects: Seismic analysis of safety related collocated concrete structures

    International Nuclear Information System (INIS)

    Joshi, J.R.

    2000-01-01

    The Process, Purification and Stack Buildings are collocated safety related concrete shear wall structures with plan dimensions in excess of 100 feet. An important aspect of their seismic analysis was the determination of structure soil structure interaction (SSSI) effects, if any. The SSSI analysis of the Process Building, with one other building at a time, was performed with the SASSI computer code for up to 50 frequencies. Each combined model had about 1500 interaction nodes. Results of the SSSI analysis were compared with those from soil structure interaction (SSI) analysis of the individual buildings, done with ABAQUS and SASSI codes, for three parameters: peak accelerations, seismic forces and the in-structure floor response spectra (FRS). The results may be of wider interest due to the model size and the potential applicability to other deep soil layered sites. Results obtained from the ABAQUS analysis were consistently higher, as expected, than those from the SSI and SSSI analyses using the SASSI. The SSSI effect between the Process and Purification Buildings was not significant. The Process and Stack Building results demonstrated that under certain conditions a massive structure can have an observable effect on the seismic response of a smaller and less stiff structure

  10. Potential seismic structural failure modes associated with the Zion Nuclear Plant. Seismic safety margins research program (Phase I). Project VI. Fragilities

    International Nuclear Information System (INIS)

    1979-10-01

    The Zion 1 and 2 Nuclear Power Plant consists of a number of structures. The most important of these from the viewpoint of safety are the containment buildings, the auxiliary building, the turbine building, and the crib house (or intake structure). The evaluation of the potential seismic failure modes and determination of the ultimate seismic capacity of the structures is a complex undertaking which will require a large number of detailed calculations. As the first step in this evaluation, a number of potential modes of structural failure have been determined and are discussed. The report is principally directed towards seismically induced failure of structures. To some extent, modes involving soil foundation failures are discussed in so far as they affect the buildings. However, failure modes involving soil liquefaction, surface faulting, tsunamis, etc., are considered outside the scope of this evaluation

  11. The art of appropriate evaluation : a guide for highway safety program managers

    Science.gov (United States)

    2008-08-01

    The guide, updated from its original release in 1999, is intended for project managers who will oversee the evaluation of traffic safety programs. It describes the benefits of evaluation and provides an overview of the steps involved. The guide inclu...

  12. Seismic Safety Margins Research Program. Phase I final report - Subsystem response (Project V)

    International Nuclear Information System (INIS)

    Shieh, L.C.; Chuang, T.Y.; O'Connell, W.J.

    1981-10-01

    This document reports on (1) the computation of the responses of subsystems, given the input subsystem support motion for components and systems whose failure can lead to an accident sequence (radioactive release), and (2) the results of a sensitivity study undertaken to determine the contributions of the several links in the seismic methodology chain (SMC) - seismic input (SI), soil-structure interaction (SSI), structure response (STR), and subsystem response (SUB) - to the uncertainty in subsystem response. For the singly supported subsystems (e.g., pumps, turbines, electrical control panels, etc.), we used the spectral acceleration response of the structure at the point where the subsystem components were mounted. For the multiple supported subsystems, we developed 13 piping models of five safety-related systems, and then used the pseudostatic-mode method with multisupport input motion to compute the response parameters in terms of the parameters used in the fragility descriptions (i.e., peak resultant accelerations for valves and peak resultant moments for piping). Damping and frequency were varied to represent the sources of modeling and random uncertainty. Two codes were developed: a modified version of SAPIV which assembles the piping supports into groups depending on the support's location relative to the attached structure, and SAPPAC a stand-alone modular program from which the time-history analysis module is extracted. On the basis of our sensitivity study, we determined that the variability in the combined soil-structure interaction, structural response, and subsystem response areas contribute more to uncertainty in subsystem response than does the variability in the seismic input area, assuming an earthquake within the limited peak ground acceleration range, i.e., 0.15 to 0.30g. The seismic input variations were in terms of different earthquake time histories. (author)

  13. Safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    Iansiti, E.

    1976-01-01

    The Codes of Practice and Safety Guides that are being developed by the International Atomic Energy Agency are divided in five topical areas: Governmental Organization, Siting, Design, Operation and Quality Assurance. In each area, a scientific secretary is responsible for developing the documents and five Technical Review Committees composed of 10 to 12 experts from various Members Countries revise the drafts at different stages. A Senior Advisory Group supervises the entire programme and revises the document. A scientific co-ordinator is responsible for the co-ordination within the programme with other sections of the IAEA, and with other international organizations. In preparing a document, information on the practice adopted by Member States is collected, a group of experts is convened for preparing a preliminary draft on the basis of this material and the draft is then reviewed by the appropriate Technical Review Committee. The document is translated into various languages, reviewed by the Senior Advisory Group and sent to Member States for comments. After the comments of Member States have been received, the Technical Review Committee and then the Senior Advisory Group are convened again for the final revision of the document. Some 25 drafts, are in different stages of development. The preparation of a document in its final form takes about two years. The programme started in 1975 and to date most of the safety codes and a few safety guides have been sent to Member States for comments. These documents will have gone through the entire development procedure by early 1977. The Senior Advisory Groups and the Technical Review Committees meet on the average four times a year for a week at a time. Until now these meetings have been mainly concerned with the development of new documents or with that part of the procedure which precedes the transmission of the draft to Member States for comments. The next series of meetings will deal with the revisions needed to

  14. Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    The IAEA's Statute authorizes the Agency to 'establish or adopt' standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

  15. Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report. Specific Safety Guide

    International Nuclear Information System (INIS)

    2011-01-01

    The IAEA's Statute authorizes the Agency to 'establish or adopt' standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

  16. Development of a Seismic Setpoint Calculation Methodology Using a Safety System Approach

    International Nuclear Information System (INIS)

    Lee, Chang Jae; Baik, Kwang Il; Lee, Sang Jeong

    2013-01-01

    The Automatic Seismic Trip System (ASTS) automatically actuates reactor trip when it detects seismic activities whose magnitudes are comparable to a Safe Shutdown Earthquake (SSE), which is the maximum hypothetical earthquake at the nuclear power plant site. To ensure that the reactor is tripped before the magnitude of earthquake exceeds the SSE, it is crucial to reasonably determine the seismic setpoint. The trip setpoint and allowable value for the ASTS for Advanced Power Reactor (APR) 1400 Nuclear Power Plants (NPPs) were determined by the methodology presented in this paper. The ASTS that trips the reactor when a large earthquake occurs is categorized as a non safety system because the system is not required by design basis event criteria. This means ASTS has neither specific analytical limit nor dedicated setpoint calculation methodology. Therefore, we developed the ASTS setpoint calculation methodology by conservatively considering that of PPS. By incorporating the developed methodology into the ASTS for APR1400, the more conservative trip setpoint and allowable value were determined. In addition, the ZPA from the Operating Basis Earthquake (OBE) FRS of the floor where the sensor module is located is 0.1g. Thus, the allowance of 0.17g between OBE of 0.1 g and ASTS trip setpoint of 0.27 g is sufficient to prevent the reactor trip before the magnitude of the earthquake exceeds the OBE. In result, the developed ASTS setpoint calculation methodology is evaluated as reasonable in both aspects of the safety and performance of the NPPs. This will be used to determine the ASTS trip setpoint and allowable for newly constructed plants

  17. Outline of the report on the seismic safety examination of nuclear facilities based on the 1995 Hyogoken-Nanbu earthquake (tentative translation) - September 1995

    International Nuclear Information System (INIS)

    2003-01-01

    From the standpoint of thoroughly confirming the seismic safety of nuclear facilities, Nuclear Safety Commission established an Examination Committee on the Seismic Safety of Nuclear Power Reactor Facilities (hereinafter called Seismic Safety Examination Committee) based on the 1995 Hyogoken-Nanbu Earthquake on January 19, 1995, two days after the occurrence of the earthquake, in order to examine the validity of related guidelines on the seismic design to be used for the safety examination. This report outlines the results of the examinations by the Seismic Safety Examination Committee: basic principle of examinations at the seismic safety examination committee, overview on the related guidelines of the seismic design, information and knowledge obtained on the 1995 Hyogoken-Nanbu earthquake, examination of validity of the guidelines based on various information of the Hyogoken-Nanbu earthquake. The Seismic Design Examination Committee surveyed the related guidelines on seismic design, selected the items to be examined, and examined on those items based on the knowledge obtained from the Hyogoken-Nanbu Earthquake. As a result, the Committee confirmed that the validity of the guidelines regulating the seismic design of nuclear facilities is not impaired even though on the basis of the Hyogoken-Nanbu Earthquake. However, the people related to the nuclear facilities may not be content with the above result, but continuously put efforts in doing the following matters to improve furthermore the reliability of seismic design of nuclear facilities by always reflecting the latest knowledge on the seismic design. 1) - The people related to nuclear facilities must seriously accept the fact that valuable knowledge could be obtained from the Hyogoken-Nanbu Earthquake, try to study and analyze the obtained data, and reflect the results of investigations, studies, and examinations conducted appropriately to the seismic design of nuclear facilities referring to the investigations

  18. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  19. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  20. Current status of ground motions evaluation in seismic design guide for nuclear power facilities. Investigation on IAEA and US.NRC

    International Nuclear Information System (INIS)

    Nakajima, Masato; Ito, Hiroshi; Hirata, Kazuta

    2009-01-01

    Recently, IAEA (International Atomic Energy Agency) and US.NRC (US. Nuclear Regulatory Commission) published several standards and technical reports on seismic design and safety evaluation for nuclear power facilities. This report summarizes the current status of the international guidelines on seismic design and safety evaluation for nuclear power facilities in order to explore the future research topics. The main results obtained are as follows: 1 IAEA: (1) In the safety standard series, two levels are defined as seismic design levels, and design earthquake ground motion is determined corresponding to each seismic design level. (2) A new framework on seismic design which consists of conventional deterministic method and risk-based method is discussed in the technical report although the framework is not adopted in the safety guidelines. 2 USA: (1) US.NRC discusses a performance-based seismic design framework which has been originally developed by the private organization (American Society of Civil Engineers). (2) Design earthquakes and earthquake ground motion are mainly evaluated and determined based on probabilistic seismic hazard evaluations. 3 Future works: It should be emphasized that IAEA and US.NRC have investigated the implementation of risk-based concept into seismic design. The implementation of risk-based concept into regulation and seismic design makes it possible to consider various uncertainties and to improve accountability. Therefore, we need to develop the methods for evaluating seismic risk of structures, and to correlate seismic margin and seismic risk quantitatively. Moreover, the probabilistic method of earthquake ground motions, that is required in the risk-based design, should be applied to sites in Japan. (author)

  1. CARES-ESTSC, Seismic Structure Safety Analysis for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Costantino, C.J.; Miller, C.A.; Heymsfield, E.; Yang, A.

    1999-01-01

    1 - Description of program or function: CARES, Computer Analysis for Rapid Evaluation of Structures, was developed for NRC staff use to determine the validity and accuracy of the analysis methods used by various utilities for structural safety evaluations of nuclear power plants. CARES is organized in a modular format with the basic modules of the system performing static, seismic, and nonlinear analysis. In this release, only the seismic module is implemented. This module defines the design seismic criteria at a given site, evaluates the free-field motion, and computes the structural response and floor response spectra including soil-structure interaction. The eight options in CARES currently are: a general manager for the seismic module, deconvolution analysis, structural data preparation for soil-structure interaction (SSI) analysis, input motion preparation for SSI analysis, SSI analysis, earthquake simulations/data, PSD (Power Spectral Density) related acceleration time history/spectra analysis, and plot generation. 2 - Method of solution: The seismic module works in the frequency domain. Earthquake motion simulation is based on the fundamental property that any periodic function can be expanded in a series of sinusoidal waves. The computer uses a random number generator to produce strings of phase angles with uniform distribution in the 0-2 pi range. Then, a linear correction procedure due to Scanlon and Sacks is employed to derive an adjusted array of amplitudes. The acceleration ensemble is subsequently modified by a deterministic intensity function composed of three segments: an initial buildup, a stationary duration, and exponential steady decay. A parabolic correction procedure outlined by Jennings and Housner is applied to the acceleration ensemble to bring the end velocity of the ground motion to zero. The soil-structure system is represented by a three-dimensional lumped parameter type model. The structural model is built up from three

  2. Operational limits and conditions and operating procedures for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared as part of the Agency's programme for establishing safety standards relating to nuclear power plants. The present Safety Guide supersedes the IAEA Safety Guide on Operational Limits and Conditions for Nuclear Power Plants which was issued in 1979 as Safety Series No. 50-SG-O3. For a nuclear power plant to be operated in a safe manner, the provisions made in the final design and subsequent modifications shall be reflected in limitations on plant operating parameters and in the requirements on plant equipment and personnel. Under the responsibility of the operating organization, these shall be developed during the design safety evaluation as a set of operational limits and conditions (OLCs). A major contribution to compliance with the OLCs is made by the development and utilization of operating procedures (OPs) that are consistent with and fully implement the OLCs. The requirements for the OLCs and OPs are established in Section 5 of the IAEA Safety Requirements publication Safety of Nuclear Power Plants: Operation, which this Safety Guide supplements. The purpose of this Safety Guide is to provide guidance on the development, content and implementation of OLCs and OPs. The Safety Guide is directed at both regulators and owners/operators. This Safety Guide covers the concept of OLCs, their content as applicable to land based stationary power plants with thermal neutron reactors, and the responsibilities of the operating organization regarding their establishment, modification, compliance and documentation. The OPs to support the implementation of the OLCs and to ensure their observance are also within the scope of this Safety Guide. The particular aspects of the procedures for maintenance, surveillance, in-service inspection and other safety related activities in connection with the safe operation of nuclear power plants are outside the scope of this Safety Guide but can be found in other IAEA Safety Guides. Section 2 indicates the

  3. Operational limits and conditions and operating procedures for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    This Safety Guide was prepared as part of the Agency's programme for establishing safety standards relating to nuclear power plants. The present Safety Guide supersedes the IAEA Safety Guide on Operational Limits and Conditions for Nuclear Power Plants which was issued in 1979 as Safety Series No. 50-SG-O3. For a nuclear power plant to be operated in a safe manner, the provisions made in the final design and subsequent modifications shall be reflected in limitations on plant operating parameters and in the requirements on plant equipment and personnel. Under the responsibility of the operating organization, these shall be developed during the design safety evaluation as a set of operational limits and conditions (OLCs). A major contribution to compliance with the OLCs is made by the development and utilization of operating procedures (OPs) that are consistent with and fully implement the OLCs. The requirements for the OLCs and OPs are established in Section 5 of the IAEA Safety Requirements publication Safety of Nuclear Power Plants: Operation, which this Safety Guide supplements. The purpose of this Safety Guide is to provide guidance on the development, content and implementation of OLCs and OPs. The Safety Guide is directed at both regulators and owners/operators. This Safety Guide covers the concept of OLCs, their content as applicable to land based stationary power plants with thermal neutron reactors, and the responsibilities of the operating organization regarding their establishment, modification, compliance and documentation. The OPs to support the implementation of the OLCs and to ensure their observance are also within the scope of this Safety Guide. The particular aspects of the procedures for maintenance, surveillance, in-service inspection and other safety related activities in connection with the safe operation of nuclear power plants are outside the scope of this Safety Guide but can be found in other IAEA Safety Guides. Section 2 indicates the

  4. Packaging Review Guide for Reviewing Safety Analysis Reports for Packagings

    Energy Technology Data Exchange (ETDEWEB)

    DiSabatino, A; Biswas, D; DeMicco, M; Fisher, L E; Hafner, R; Haslam, J; Mok, G; Patel, C; Russell, E

    2007-04-12

    This Packaging Review Guide (PRG) provides guidance for Department of Energy (DOE) review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE Order 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his or her review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. This PRG is generally organized at the section level in a format similar to that recommended in Regulatory Guide 7.9 (RG 7.9). One notable exception is the addition of Section 9 (Quality Assurance), which is not included as a separate chapter in RG 7.9. Within each section, this PRG addresses the technical and regulatory bases for the review, the manner in which the review is accomplished, and findings that are generally applicable for a package that meets the approval standards. This Packaging Review Guide (PRG) provides guidance for DOE review and approval of packagings to transport fissile and Type B quantities of radioactive material. It fulfills, in part, the requirements of DOE O 460.1B for the Headquarters Certifying Official to establish standards and to provide guidance for the preparation of Safety Analysis Reports for Packagings (SARPs). This PRG is intended for use by the Headquarters Certifying Official and his review staff, DOE Secretarial offices, operations/field offices, and applicants for DOE packaging approval. The primary objectives of this PRG are to: (1) Summarize the regulatory requirements for package approval; (2) Describe the technical review procedures by which DOE determines that these requirements have been satisfied; (3) Establish and maintain the quality and uniformity of reviews; (4) Define the base from which to evaluate proposed changes in scope

  5. Probabilistic seismic safety assessment of a CANDU 6 nuclear power plant including ambient vibration tests: Case study

    Energy Technology Data Exchange (ETDEWEB)

    Nour, Ali [Hydro Québec, Montréal, Québec H2L4P5 (Canada); École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada); Cherfaoui, Abdelhalim; Gocevski, Vladimir [Hydro Québec, Montréal, Québec H2L4P5 (Canada); Léger, Pierre [École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada)

    2016-08-01

    Highlights: • In this case study, the seismic PSA methodology adopted for a CANDU 6 is presented. • Ambient vibrations testing to calibrate a 3D FEM and to reduce uncertainties is performed. • Procedure for the development of FRS for the RB considering wave incoherency effect is proposed. • Seismic fragility analysis for the RB is presented. - Abstract: Following the 2011 Fukushima Daiichi nuclear accident in Japan there is a worldwide interest in reducing uncertainties in seismic safety assessment of existing nuclear power plant (NPP). Within the scope of a Canadian refurbishment project of a CANDU 6 (NPP) put in service in 1983, structures and equipment must sustain a new seismic demand characterised by the uniform hazard spectrum (UHS) obtained from a site specific study defined for a return period of 1/10,000 years. This UHS exhibits larger spectral ordinates in the high-frequency range than those used in design. To reduce modeling uncertainties as part of a seismic probabilistic safety assessment (PSA), Hydro-Québec developed a procedure using ambient vibrations testing to calibrate a detailed 3D finite element model (FEM) of the containment and reactor building (RB). This calibrated FE model is then used for generating floor response spectra (FRS) based on ground motion time histories compatible with the UHS. Seismic fragility analyses of the reactor building (RB) and structural components are also performed in the context of a case study. Because the RB is founded on a large circular raft, it is possible to consider the effect of the seismic wave incoherency to filter out the high-frequency content, mainly above 10 Hz, using the incoherency transfer function (ITF) method. This allows reducing significantly the non-necessary conservatism in resulting FRS, an important issue for an existing NPP. The proposed case study, and related methodology using ambient vibration testing, is particularly useful to engineers involved in seismic re-evaluation of

  6. Radiological protection for medical exposure to ionizing radiation. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    radiotherapy owing to an ageing population. In addition, further growth in medical radiology can be expected in developing States, where at present facilities and services are often lacking. The risks associated with these expected increases in medical exposures should be outweighed by the benefits. For the purposes of radiation protection, ionizing radiation exposures are divided into three types: Medical exposure, which is mainly the exposure of patients as part of their diagnosis or treatment (see below); Occupational exposure, which is the exposure of workers incurred in the course of their work, with some specific exclusions; and Public exposure, which comprises all other exposures of members of the public that are susceptible to human control. Medical exposure is defined in the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS, the Standards) as: 'Exposure incurred by patients as part of their own medical or dental diagnosis or treatment; by persons, other than those occupationally exposed, knowingly while voluntarily helping in the support and comfort of patients; and by volunteers in a programme of biomedical research involving their exposure.' This Safety Guide covers all of the medical exposures defined above, with emphasis on the radiological protection of patients, but does not cover exposures of workers or the public derived from the application of medical radiation sources. Guidance relating to these exposures can be found in the Safety Guide on Occupational Radiation Protection. In addition to the IAEA, several intergovernmental and international organizations, among them the European Commission, the International Commission on Radiological Protection (ICRP), the Pan American Health Organization (PAHO) and the World Health Organization (WHO), have already published numerous recommendations, guides and codes of practice relevant to this subject area. National authorities should therefore

  7. Radiological protection for medical exposure to ionizing radiation. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    radiotherapy owing to an ageing population. In addition, further growth in medical radiology can be expected in developing States, where at present facilities and services are often lacking. The risks associated with these expected increases in medical exposures should be outweighed by the benefits. For the purposes of radiation protection, ionizing radiation exposures are divided into three types: Medical exposure, which is mainly the exposure of patients as part of their diagnosis or treatment (see below). Occupational exposure, which is the exposure of workers incurred in the course of their work, with some specific exclusions. And Public exposure, which comprises all other exposures of members of the public that are susceptible to human control. Medical exposure is defined in the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS, the Standards) as: 'Exposure incurred by patients as part of their own medical or dental diagnosis or treatment. By persons, other than those occupationally exposed, knowingly while voluntarily helping in the support and comfort of patients. And by volunteers in a programme of biomedical research involving their exposure.' This Safety Guide covers all of the medical exposures defined above, with emphasis on the radiological protection of patients, but does not cover exposures of workers or the public derived from the application of medical radiation sources. Guidance relating to these exposures can be found in the Safety Guide on Occupational Radiation Protection. In addition to the IAEA, several intergovernmental and international organizations, among them the European Commission, the International Commission on Radiological Protection (ICRP), the Pan American Health Organization (PAHO) and the World Health Organization (WHO), have already published numerous recommendations, guides and codes of practice relevant to this subject area. National authorities should therefore

  8. Meteorological events in site evaluation for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide provides recommendations and guidance on conducting hazard assessments of extreme and rare meteorological phenomena. It is of interest to safety assessors and regulators involved in the licensing process as well as to designers of nuclear power plants. This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It supplements the IAEA Safety Requirements publication on Site Evaluation for Nuclear Facilities which is to supersede the Code on the Safety of Nuclear Power Plants: Siting, Safety Series No. 50-C-S (Rev. 1), IAEA, Vienna (1988). The present Safety Guide supersedes two earlier Safety Guides: Safety Series No. 50-SG-S11A (1981) on Extreme Meteorological Events in Nuclear Power Plant Siting, Excluding Tropical Cyclones and Safety Series No. 50-SG-S11B (1984) on Design Basis Tropical Cyclone for Nuclear Power Plants. The purpose of this Safety Guide is to provide recommendations and guidance on conducting hazard assessments of extreme and rare meteorological phenomena. This Safety Guide provides interpretation of the Safety Requirements publication on Site Evaluation for Nuclear Facilities and guidance on how to fulfil these requirements. It is aimed at safety assessors or regulators involved in the licensing process as well as designers of nuclear power plants, and provides them with guidance on the methods and procedures for analyses that support the assessment of the hazards associated with extreme and rare meteorological events. This Safety Guide discusses the extreme values of meteorological variables and rare meteorological phenomena, as well as their rates of occurrence, according to the following definitions: (a) Extreme values of meteorological variables such as air temperature and wind speed characterize the meteorological or climatological environment. And (b) Rare meteorological phenomena

  9. Seismic fragility testing of naturally-aged, safety-related, class 1E battery cells

    International Nuclear Information System (INIS)

    Bonzon, L.L.; Hente, D.B.; Kukreti, B.M.; Schendel, J.S.; Black, D.A.; Paulsen, G.D.; Tulk, J.D.; Janis, W.J.; Aucoin, B.D.

    1984-01-01

    The concern over seismic susceptibility of naturally-aged lead-acid batteries used for safety-related emergency power in nuclear power stations was brought about by battery problems that periodically had been reported in Licensee Event Reports (LERs). The Turkey Point Station had reported cracked and buckled plates in several cells in October 1974 (LER 75-5). The Fitzpatrick Station had reported cracked battery cell cases in October 1977 (LER 77-55) and again in September 1979 (LER 79-59). The Browns Ferry Station had reported a cracked cell leaking a small quantity of electrolyte in July 1981 (LER 81-42). The Indian Point Station had reported cracked and leaking cells in both February (LER 82-7) and April 1982 (LER 82-16); both of these LERs indicated the cracked cells were due to expansion (i.e., growth) of the positive plates

  10. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

  11. German seismic regulations

    International Nuclear Information System (INIS)

    Danisch, Ruediger

    2002-01-01

    Rules and regulations for seismic design in Germany cover the following: seismic design of conventional buildings; and seismic design of nuclear facilities. Safety criteria for NPPs, accident guidelines, and guidelines for PWRs as well as safety standards are cited. Safety standards concerned with NPPs seismic design include basic principles, soil analysis, design of building structures, design of mechanical and electrical components, seismic instrumentation, and measures to be undertaken after the earthquake

  12. Nuclear regulatory guides for LWR (PWR) fuel in Japan and some related safety research

    International Nuclear Information System (INIS)

    Ichikawa, M.

    1994-01-01

    The general aspects of licensing procedure for NPPs in Japan and regulatory guides are described. The expert committee reports closely related to PWR fuel are reviewed. Some major results of reactor safety research experiments at NSPR (Nuclear Safety Research Reactor of JAERI) used for establishment of related guide, are discussed. It is pointed out that the reactor safety research in Japan supports the regularity activities by establishing and revising guides and preparing the necessary regulatory data as well as improving nuclear safety. 10 figs., 4 refs

  13. Organization and staffing of the regulatory body for nuclear facilities. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this safety guide is to provide recommendations for national authorities on the appropriate management system, organization and staffing for the regulatory body responsible for the regulation of nuclear facilities in order to achieve compliance with the applicable safety requirements. This safety guide covers the organization and staffing in relation to nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And radioactive waste management facilities such as treatment, storage and disposal facilities. This safety guide also covers issues related to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation

  14. Nuclear regulatory guides for LWR (PWR) fuel in Japan and some related safety research

    Energy Technology Data Exchange (ETDEWEB)

    Ichikawa, M [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    1994-12-31

    The general aspects of licensing procedure for NPPs in Japan and regulatory guides are described. The expert committee reports closely related to PWR fuel are reviewed. Some major results of reactor safety research experiments at NSPR (Nuclear Safety Research Reactor of JAERI) used for establishment of related guide, are discussed. It is pointed out that the reactor safety research in Japan supports the regularity activities by establishing and revising guides and preparing the necessary regulatory data as well as improving nuclear safety. 10 figs., 4 refs.

  15. Introduction of conditional mean spectrum and conditional spectrum in the practice of seismic safety evaluation in China

    Science.gov (United States)

    Ji, Kun; Bouaanani, Najib; Wen, Ruizhi; Ren, Yefei

    2018-05-01

    This paper aims at implementing and introducing the use of conditional mean spectrum (CMS) and conditional spectrum (CS) as the main input parameters in the practice of seismic safety evaluation (SSE) in China, instead of the currently used uniform hazard spectrum (UHS). For this purpose, a procedure for M-R-epsilon seismic hazard deaggregation in China was first developed. For illustration purposes, two different typical sites in China, with one to two dominant seismic zones, were considered as examples to carry out seismic hazard deaggregation and illustrate the construction of CMS/CS. Two types of correlation coefficients were used to generate CMS and the results were compared over a vibration period range of interest. Ground motion records were selected from the NSMONS (2007-2015) and PEER NGA-West2 databases to correspond to the target CMS and CS. Hazard consistency of the spectral accelerations of the selected ground motion records was evaluated and validated by computing the annual exceedance probability rate of the response spectra and comparing the results to the hazard curve corresponding to each site of concern at different periods. The tools developed in this work and their illustrative application to specific case studies in China are a first step towards the adoption of CMS and CS into the practice of seismic safety evaluation in this country.

  16. Seismic evaluation of existing nuclear power plants

    International Nuclear Information System (INIS)

    2003-01-01

    The IAEA nuclear safety standards publications address the site evaluation and the design of new nuclear power plants (NPPs), including seismic hazard assessment and safe seismic design, at the level of the Safety Requirements as well as at the level of dedicated Safety Guides. It rapidly became apparent that the existing nuclear safety standards documents were not adequate for handling specific issues in the seismic evaluation of existing NPPs, and that a dedicated document was necessary. This is the purpose of this Safety Report, which is written in the spirit of the nuclear safety standards and can be regarded as guidance for the interpretation of their intent. Worldwide experience shows that an assessment of the seismic capacity of an existing operating facility can be prompted for the following: (a) Evidence of a greater seismic hazard at the site than expected before, owing to new or additional data and/or to new methods; (b) Regulatory requirements, such as periodic safety reviews, to ensure that the plant has adequate margins for seismic loads; (c) Lack of anti-seismic design or poor anti-seismic design; (d) New technical finding such as vulnerability of some structures (masonry walls) or equipment (relays), other feedback and new experience from real earthquakes. Post-construction evaluation programmes evaluate the current capability of the plant to withstand the seismic concern and identify any necessary upgrades or changes in operating procedures. Seismic qualification is distinguished from seismic evaluation primarily in that seismic qualification is intended to be performed at the design stage of a plant, whereas seismic evaluation is intended to be applied after a plant has been constructed. Although some guidelines do exist for the evaluation of existing NPPs, these are not established at the level of a regulatory guide or its equivalent. Nevertheless, a number of existing NPPs throughout the world have been and are being subjected to review of their

  17. Protection of the patient in medical exposure - the related IAEA safety guide

    International Nuclear Information System (INIS)

    Turai, I.

    1999-01-01

    The Radiation Safety Section of the Agency has recently completed the draft Safety Guide on Radiation Protection in Medical Exposures' for submission to the Publication Committee of the IAEA. The author as served as one of the scientific secretaries responsible for the preparation and review of this document in the last two years. The drafts of this IAEA Safety Guide have undergone a detailed review process by specialists of 14 Member States and the co-sponsoring organizations, the Pan American Health Organization and the World Health Organization (WHO). The last draft is the primary source of this paper. The Safety Guide will be part of the Safety Standards Series. It is addressed to Regulatory Authorities and other National Institutions to provide them with guidance at the national level on the practical implementation of Appendix II (Medical Exposure) of the International Basic Safety Standards for the Protection against Ionizing Radiation and for the Safety of Radiation Sources

  18. SSI sensitivity studies and model improvements for the US NRC Seismic Safety Margins Research Program. Rev. 1

    International Nuclear Information System (INIS)

    Johnson, J.J.; Maslenikov, O.R.; Benda, B.J.

    1984-10-01

    The Seismic Safety Margins Research Program (SSMRP) is a US NRC-funded program conducted by Lawrence Livermore National Laboratory. Its goal is to develop a complete fully coupled analysis procedure for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. In Phase II of the SSMRP, the methodology was applied to the Zion nuclear power plant. Three topics in the SSI analysis of Zion were investigated and reported here - flexible foundation modeling, structure-to-structure interaction, and basemat uplift. The results of these investigations were incorporated in the SSMRP seismic risk analysis. 14 references, 51 figures, 13 tables

  19. Regulatory analysis for resolution of Unresolved Safety Issue A-46, seismic qualification of equipment in operating plants

    International Nuclear Information System (INIS)

    Chang, T.Y.; Anderson, N.R.

    1987-02-01

    The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform required safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants must be reassessed to determine whether requalification is necessary. The objective of technical studies performed under Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring these plants to meet the criteria that are applied to new plants. This report presents the regulatory analysis for Unresolved Safety Issue (USI) A-46. It includes: Statement of the Problem; the Objective of USI A-46; a Summary of A-46 Tasks; a Proposed Implementation Procedure; a Value-Impact Analysis; Application of the Backfit Rule; 10 CFR 50.109; Implementation; and Operating Plants To Be Reviewed to USI A-46 Requirements

  20. An introduction to a new IAEA safety guide: 'ageing management for nuclear power plants'

    International Nuclear Information System (INIS)

    Pachner, J.; Inagaki, T.; Kang, K.S.

    2008-01-01

    This paper reports on a new IAEA Safety Guide entitled 'Ageing Management for Nuclear Power Plants' which is currently in an advanced draft form, awaiting approval of publication. The new Safety Guide will be an umbrella document for a comprehensive set of guidance documents on ageing management which have been issued by the IAEA. The Safety Guide first presents basic concepts of ageing management as a common basis for the recommendations on: proactive management of ageing throughout the life cycle of a nuclear power plant (NPP); systematic approach to managing ageing in the operation of NPPs; managing obsolescence; and review of ageing management for long term operation (life extension). The Safety Guide is intended to assist operators in establishing, implementing and improving systematic ageing management programs in NPPs and may be used by regulators in preparing regulatory standards and guides, and in verifying that ageing in nuclear power plants is being effectively managed. (author)

  1. Lean Six-Sigma in Aviation Safety: An implementation guide for measuring aviation system’s safety performance

    OpenAIRE

    Panagopoulos, I.; Atkin, C.J.; Sikora, I.

    2016-01-01

    The paper introduces a conceptual framework that could improve the safety performance measurement process and ultimately the aviation system safety performance. The framework provides an implementation guide on how organisations could design and develop a proactive, measurement tool for assessing and measuring the Acceptable Level of Safety Performance (ALoSP) at sigma (σ) level, a statistical measurement unit. In fact, the methodology adapts and combines quality management tools, a leading i...

  2. Seismic safety margins research program. Phase I final report - Major structure response (Project IV)

    International Nuclear Information System (INIS)

    Benda, B.J.; Johnson, J.J.; Lo, T.Y.

    1981-08-01

    The primary task of the Major Structure Response Project within the Seismic Safety Margins Research Program (SSMRP) was to develop detailed finite element models of the Zion Nuclear Power Plant's containment building and auxiliary-fuel-turbine (AFT) complex. The resulting models served as input to the seismic methodology analysis chain. The containment shell was modeled as a series of beam elements with the shear and bending characteristics of a circular cylindrical shell. Masses and rotary inertias were lumped at nodal points; thirteen modes were included in the analysis. The internal structure was modeled with three-dimensional finite elements, with masses again lumped at selected nodes; sixty modes were included in the analysis. The model of the AFT complex employed thin plate and shell elements to represent the concrete shear walls and floor diaphragms, and beam and truss elements to model the braced frames. Because of the size and complexity of the model, and the potentially large number of degrees of freedom, masses were lumped at a limited number of node points. These points were selected so as to minimize the effect of the discrete mass distribution on structural response. One hundred and thirteen modes were extracted. A second objective of Project IV was to investigate the effects of uncertainty and variability on structural response. To this end, four side studies were conducted. Three of them, briefly summarized in this volume, addressed themselves respectively to an investigation of sources of random variability in the dynamic response of nuclear power plant structures; formulation of a methodology for modeling and evaluating the effects of structural uncertainty on predicted modal characteristics of major nuclear power plant structures and substructures; and a preliminary evaluation of nonlinear responses in shear-wall structures. A fourth side study, reported in detail in this volume, quantified variations in dynamic characteristics and seismic

  3. IAEA safety guides in the light of recent developments in earthquake engineering

    International Nuclear Information System (INIS)

    Gurpinar, A.

    1988-11-01

    The IAEA safety guides 50-SG-S1 and 50-SG-S2 emphasize on the determination of the design basis earthquake ground motion and earthquake resistant design considerations for nuclear power plants, respectively. Since the elaboration of these safety guides years have elapsed and a review of some of these concepts is necessary, taking into account the information collected and the technical developments. In this article, topics within the scope of these safety guides are discussed. In particular, the results of some recent research which may have a bearing on the nuclear industry are highlighted. Conclusions and recommendations are presented. 6 fig., 19 refs. (F.M.)

  4. Consecutive collection of new finding and knowledge on science and technology to be reflected to seismic safety assessment for nuclear facilities

    International Nuclear Information System (INIS)

    Tsutsumi, Hideaki; Iijima, Toru

    2013-05-01

    JNES had been collecting and analyzing new finding and knowledge on science and technology to be reflected to seismic safety assessment for nuclear facilities, which was updated so as to develop a system to organize and disseminate such information in response to Nuclear Regulation Authority (NRA)'s policy on new safety regulations requesting enhanced protective measures against extreme natural hazards. The tasks were as follows; (1) collection of new finding and knowledge from seismic safety research of JNES, (2) constructing database of seismic safety research from documents published by committees and including the Great East Japan Earthquake and (3) dissemination of information related to seismic research. As for JFY 2012 activities, collecting and analyzing new finding and knowledge were on three areas such as active fault, seismic source/ground motion and tsunami. 4 theme related with the Great East Japan Earthquake, 7 items not related with the Great East Japan Earthquake and one item on external event were collected and analyzed whether incorporating in seismic safety research important for regulation to increase seismic safety of nuclear facilities, with no such theme confirmed. (T. Tanaka)

  5. [Efficacy and safety of ultrasound-guided or neurostimulator-guided bilateral axillary brachial plexus block].

    Science.gov (United States)

    Xu, C S; Zhao, X L; Zhou, H B; Qu, Z J; Yang, Q G; Wang, H J; Wang, G

    2017-10-17

    Objective: To explore the efficacy and safety of bilateral axillary brachial plexus block under the guidance of ultrasound or neurostimulator. Methods: From February 2012 to April 2014, 120 patients undergoing bilateral hand/forearm surgery in Beijing Jishuitan Hospital were enrolled and anaesthetized with bilateral axillary brachial plexus block. All patients were divided into two groups randomly using random number table: the ultrasound-guided group (group U, n =60) and the neurostimulator-guidedgroup (group N, n =60). The block was performed with 0.5% ropivacaine. Patients' age, sex and operation duration were recorded. Moreover, success rate, performance time, onset of sensor and motor block, performance pain, patient satisfaction degree and the incidence of related complications were also documented. Venous samples were collected at selected time points and the total and the plasma concentrations of ropivacaine were analyzed with HPLC. Results: The performance time, the onset of sensor block and the onset of motor block of group U were (8.2±1.5), (14.2± 2.2)and (24.0±3.5)min respectively, which were markedly shorter than those in group N( (14.6±3.9), (19.9±3.8), (28.8±4.2)min, respectively), and the differences were statistically significant( t =11.74, 10.09, 6.73, respectively, all P 0.05). No analgesic was superadded and no other anesthesia methods were applied. No complications were detected perioperatively. Conclusions: The bilateral axillary brachial plexus block under the guidance of ultrasound or neurostimulator are both effective and safe for bilateral hand/forearm surgery. However, the ultrasound-guided block may be more clinically beneficial because of its shorter performance time, rapid onset and higher patient satisfaction degree.

  6. Guide to the safety design examination about light water reactor facilities for power generation

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This guide was compiled to evaluate the validity of the design policy when the safety design is examined at the time of the application for approval of the installation of nuclear reactors. About 7 years has elapsed since the existing guide was established, and the more appropriate guide to evaluate the safety should be made on the basis of the knowledge and experience accumulated thereafter. The range of application of this guide is limited to the above described evaluation, and it is not intended as the general standard for the design of nuclear reactors. First, the definition of the words used in this guide is given. Then, the guide to the safety examination is described about the general matters of reactor facilities, nuclear reactors and the measuring and controlling system, reactor-stopping system, reactivity-controlling system and safety protection system, reactor-cooling system, reactor containment vessels, fuel handling and waste treatment system. Several matters which require attention in the application of this guide or the clarification of the significance and interpretation of the guide itself were found, therefore the explanation about them was added at the end of this guide. (Kako, I.)

  7. Safety critical systems handbook a straightforward guide to functional safety : IEC 61508 (2010 edition) and related standards

    CERN Document Server

    Smith, David J

    2010-01-01

    Electrical, electronic and programmable electronic systems increasingly carry out safety functions to guard workers and the public against injury or death and the environment against pollution. The international functional safety standard IEC 61508 was revised in 2010, and this is the first comprehensive guide available to the revised standard. As functional safety is applicable to many industries, this book will have a wide readership beyond the chemical and process sector, including oil and gas, power generation, nuclear, aircraft, and automotive industries, plus project, instrumentation, design, and control engineers. * The only comprehensive guide to IEC 61508, updated to cover the 2010 amendments, that will ensure engineers are compliant with the latest process safety systems design and operation standards* Helps readers understand the process required to apply safety critical systems standards* Real-world approach helps users to interpret the standard, with case studies and best practice design examples...

  8. Application of the concepts of exclusion, exemption and clearance. Safety guide

    International Nuclear Information System (INIS)

    2007-01-01

    The objective of this Safety Guide is to provide guidance to national authorities, including regulatory bodies, and operating organizations on the application of the concepts of exclusion, exemption and clearance as established in the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS). The Safety Guide includes specific values of activity concentration for both radionuclides of natural origin and those of artificial origin that may be used for bulk amounts of material for the purpose of applying exclusion or exemption. It also elaborates on the possible application of these values to clearance

  9. Current USAEC seismic requirements for nuclear power plants

    International Nuclear Information System (INIS)

    Mehta, D.S.

    1975-01-01

    The principal seismic and geologic considerations which guide the USAEC in its evaluation of the suitability of proposed sites for nuclear power plants and plant design bases are set forth as design criteria in the AEC regulatory guides. The basic requirements of seismic design and analysis for seismic Category I structures, components, and systems important to public safety have been established in the USAEC regulatory guides and Code of Federal Regulations. It is pointed out that the current state-of-art techniques, best available technology, and additional studies in the field of earthquake engineering can be utilized to resolve seismic concerns. The seismic design requirements for nuclear plants to withstand postulated earthquakes can be standardized and this will be a significant milestone in the continuation of the Nuclear Standardization Program. (author)

  10. Protection against internal fires and explosions in the design of nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Experience of the past two decades in the operation of nuclear power plants and modern analysis techniques confirm that fire may be a real threat to nuclear safety and should receive adequate attention from the beginning of the design process throughout the life of the plant. Within the framework of the NUSS programme, a Safety Guide on fire protection had therefore been developed to enlarge on the general requirements given in the Code. Since its first publication in 1979, there has been considerable development in protection technology and analysis methods and after the Chernobyl accident it was decided to revise the existing Guide. This Safety Guide supplements the requirements established in Safety of Nuclear Power Plants: Design. It supersedes Safety Series No. 50-SG-D2 (Rev. 1), Fire Protection in Nuclear Power Plants: A Safety Guide, issued in 1992.The present Safety Guide is intended to advise designers, safety assessors and regulators on the concept of fire protection in the design of nuclear power plants and on recommended ways of implementing the concept in some detail in practice

  11. IAEA program for the preparation of safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    1975-01-01

    On the 13th of September, 1974, the IAEA Governors' Council has given its consent to the programme for the establishment of safety codes and guides (annex VII to IAEA document G.C. (XVIII/526)). The programme envisages the establishment of one code of practice for each of the issues governmental organization, siting, design, operation and quality assurance and also of about 50 safety guides between 1975 and 1980. These codes will contain the minimum requirements for the safety of the nuclear power stations, their systems and components. The guides will recommend methods to achieve the aims stated in the codes. It is the purpose of these IAEA activities to provide recommendations and guiding rules which may serve as standards for the assessment of the safety of nuclear power stations for all nations which may become participants in the peaceful use of nuclear energy within the next few years. (orig./AK) [de

  12. Safe adventures. An ethnographic study of safety and adventure guides in Arctic Norway

    OpenAIRE

    Johannessen, Mats Hoel

    2016-01-01

    With numerous entrepreneurs already established within the area, adventure tourism is a growing industry within Arctic Norway. The continuously expanding interest for the phenomenon has gained universities’ attention with recent education programs for guides being established. A cultural change involving a more professionalized approach to adventure tourism has also been noticed. At the forefront of ensuring tourists’ safety are the guides, who work in the area. In former research on safety i...

  13. ASN guide project. Safety policy and management in INBs (base nuclear installations)

    International Nuclear Information System (INIS)

    2010-01-01

    This guide presents the recommendations of the French Nuclear Safety Authority (ASN) in the field of safety policy and management (PMS) for base nuclear installations (INBs). It gives an overview and comments of some prescriptions of the so-called INB order and PMS decision. These regulatory texts define a framework for provisions any INB operator must implement to establish his safety policy, to define and implement a system which allows the safety to be maintained, the improvement of his INB safety to be permanently looked for. The following issues are addressed: operator's safety policy, identification of elements important for safety, of activities pertaining to safety, and of associated requirements, safety management organization and system, management of activities pertaining to safety, documentation and archiving

  14. Tornadoes: Nature's Most Violent Storms. A Preparedness Guide Including Safety Information for Schools.

    Science.gov (United States)

    American National Red Cross, Washington, DC.

    This preparedness guide explains and describes tornadoes, and includes safety information for schools. A tornado is defined as a violently rotating column of air extending from a thunderstorm to the ground. The guide explains the cause of tornadoes, provides diagrams of how they form, describes variations of tornadoes, and classifies tornadoes by…

  15. Field Test of the World Health Organization Multi-Professional Patient Safety Curriculum Guide

    Science.gov (United States)

    Farley, Donna; Zheng, Hao; Rousi, Eirini; Leotsakos, Agnès

    2015-01-01

    Introduction Although the importance of training in patient safety has been acknowledged for over a decade, it remains under-utilized and under-valued in most countries. WHO developed the Multi-professional Patient Safety Curriculum Guide to provide schools with the requirements and tools for incorporating patient safety in education. It was field tested with 12 participating schools across the six WHO regions, to assess its effectiveness for teaching patient safety to undergraduate and graduate students in a global variety of settings. Methods The evaluation used a combined prospective/retrospective design to generate formative information on the experiences of working with the Guide and summative information on the impacts of the Guide. Using stakeholder interviews and student surveys, data were gathered from each participating school at three times: the start of the field test (baseline), soon after each school started teaching, and soon after each school finished teaching. Results Stakeholders interviewed were strongly positive about the Guide, noting that it emphasized universally important patient safety topics, was culturally appropriate for their countries, and gave credibility and created a focus on patient safety at their schools. Student perceptions and attitudes regarding patient safety improved substantially during the field test, and their knowledge of the topics they were taught doubled, from 10.7% to 20.8% of correct answers on the student survey. Discussion This evaluation documented the effectiveness of the Curriculum Guide, for both ease of use by schools and its impacts on improving the patient safety knowledge of healthcare students. WHO should be well positioned to refine the contents of the Guide and move forward in encouraging broader use of the Guide globally for teaching patient safety. PMID:26406893

  16. Field Test of the World Health Organization Multi-Professional Patient Safety Curriculum Guide.

    Science.gov (United States)

    Farley, Donna; Zheng, Hao; Rousi, Eirini; Leotsakos, Agnès

    2015-01-01

    Although the importance of training in patient safety has been acknowledged for over a decade, it remains under-utilized and under-valued in most countries. WHO developed the Multi-professional Patient Safety Curriculum Guide to provide schools with the requirements and tools for incorporating patient safety in education. It was field tested with 12 participating schools across the six WHO regions, to assess its effectiveness for teaching patient safety to undergraduate and graduate students in a global variety of settings. The evaluation used a combined prospective/retrospective design to generate formative information on the experiences of working with the Guide and summative information on the impacts of the Guide. Using stakeholder interviews and student surveys, data were gathered from each participating school at three times: the start of the field test (baseline), soon after each school started teaching, and soon after each school finished teaching. Stakeholders interviewed were strongly positive about the Guide, noting that it emphasized universally important patient safety topics, was culturally appropriate for their countries, and gave credibility and created a focus on patient safety at their schools. Student perceptions and attitudes regarding patient safety improved substantially during the field test, and their knowledge of the topics they were taught doubled, from 10.7% to 20.8% of correct answers on the student survey. This evaluation documented the effectiveness of the Curriculum Guide, for both ease of use by schools and its impacts on improving the patient safety knowledge of healthcare students. WHO should be well positioned to refine the contents of the Guide and move forward in encouraging broader use of the Guide globally for teaching patient safety.

  17. Design of Instrumentation and Control Systems for Nuclear Power Plants. Specific Safety Guide

    International Nuclear Information System (INIS)

    2016-01-01

    This publication is a revision and combination of two Safety Guides, IAEA Safety Standards Series No. NS-G-1.1 and No. NS-G-1.3. The revision takes into account developments in instrumentation and control (I&C) systems since the publication of the earlier Safety Guides. The main changes relate to the continuing development of computer applications and the evolution of the methods necessary for their safe, secure and practical use. In addition, account is taken of developments in human factors engineering and the need for computer security. This Safety Guide references and takes into account other IAEA Safety Standards and Nuclear Security Series publications that provide guidance relating to I&C design

  18. Safety guide data on radiation shielding in a reprocessing facility

    International Nuclear Information System (INIS)

    Sekiguchi, Noboru; Naito, Yoshitaka

    1986-04-01

    In a reprocessing facility, various radiation sources are handled and have many geometrical conditions. To aim drawing up a safety guidebook on radiation shielding in order to evaluate shielding safety in a reprocessing facility with high reliability and reasonableness, JAERI trusted investigation on safety evaluation techniques of radiation shielding in a reprocessing facility to Nuclear Safety Research Association. This report is the collection of investigation results, and describes concept of shielding safety design principle, radiation sources in reprocessing facility and estimation of its strength, techniques of shielding calculations, and definite examples of shielding calculation in reprocessing facility. (author)

  19. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  20. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  1. Seismic qualification of equipment

    International Nuclear Information System (INIS)

    Heidebrecht, A.C.; Tso, W.K.

    1983-03-01

    This report describes the results of an investigation into the seismic qualification of equipment located in CANDU nuclear power plants. It is particularly concerned with the evaluation of current seismic qualification requirements, the development of a suitable methodology for the seismic qualification of safety systems, and the evaluation of seismic qualification analysis and testing procedures

  2. LASR-Guided Variability Subtraction: The Linear Algorithm for Significance Reduction of Stellar Seismic Activity

    Science.gov (United States)

    Horvath, Sarah; Myers, Sam; Ahlers, Johnathon; Barnes, Jason W.

    2017-10-01

    Stellar seismic activity produces variations in brightness that introduce oscillations into transit light curves, which can create challenges for traditional fitting models. These oscillations disrupt baseline stellar flux values and potentially mask transits. We develop a model that removes these oscillations from transit light curves by minimizing the significance of each oscillation in frequency space. By removing stellar variability, we prepare each light curve for traditional fitting techniques. We apply our model to $\\delta$-Scuti KOI-976 and demonstrate that our variability subtraction routine successfully allows for measuring bulk system characteristics using traditional light curve fitting. These results open a new window for characterizing bulk system parameters of planets orbiting seismically active stars.

  3. Radiation protection programmes for the transport of radioactive material. Safety guide

    International Nuclear Information System (INIS)

    2007-01-01

    This Safety Guide provides guidance on meeting the requirements for the establishment of radiation protection programmes (RPPs) for the transport of radioactive material, to optimize radiation protection in order to meet the requirements for radiation protection that underlie the Regulations for the Safe Transport of Radioactive Material. This Guide covers general aspects of meeting the requirements for radiation protection, but does not cover criticality safety or other possible hazardous properties of radioactive material. The annexes of this Guide include examples of RPPs, relevant excerpts from the Transport Regulations, examples of total dose per transport index handled, a checklist for road transport, specific segregation distances and emergency instructions for vehicle operators

  4. Seismic risk and safety of nuclear installations. A look at the Cadarache Centre

    International Nuclear Information System (INIS)

    Verrhiest-Leblanc, G.; Chevallier, A.

    2010-01-01

    After a brief recall of some important seismic events which occurred in the past in the south-eastern part of France, the authors indicate the nuclear installations present in this region. They outline the difference between requirements for a usual building and for basic nuclear installations. They indicate laws and regulations which are to be applied to these installations like to any hazardous industrial installation. They describe the seismic risk as it has been determined for the Cadarache area, and evoke the para-seismic design of new nuclear installations which are to be built in Cadarache and actions for a para-seismic reinforcement of existing constructions. Finally, they evoke organisational aspects (emergency plans) and the approach for a better information and transparency about the seismic risk

  5. Seismic assessment of safety-related structures: laboratory testing of the pressure relief duct frame at pickering NPP

    International Nuclear Information System (INIS)

    Ghobarah, A.; Biddah, A.; Pilette, C.

    1995-01-01

    The pressure relief duct (PRD) is a Special safety System in the CANDU-PHW multi-unit nuclear power plants (NPP). It is designed to contain and direct the outflow from the reactor building to the pressure suppression and containing systems in the vacuum building. The PRD is a large elevated reinforced concrete box structure of internal width of 6.1 m, height of 7.6 m, and wall thickness of 0.6 m. The PRD is 662 m long and is supported every 22 m by concrete frames of height of 21 m. Typical frame members are 1.8 m in depth and width. A representative elevation of the frame is presented. The section of the PRD under investigation was designed and constructed before the current seismic design codes were in effect. An assessment of the PRD structure when subjected to various levels of ground motion has shown that the frame has a limited seismic withstand capacity. Its seismic performance is dependent on the ductility of the beams and on the ability of the beam-column joint to transfer bending moments and shear forces. The objectives of this study are to provide the data to validate the frame analysis results through laboratory testing of a scaled specimen of the beam-column joint, and to compare the observed response with the response of a beam-column joint when the shear reinforcement is detailed according to current seismic design codes. (author). 3 refs., 10 figs

  6. Communication and Consultation with Interested Parties by the Regulatory Body. General Safety Guide

    International Nuclear Information System (INIS)

    2017-01-01

    This Safety Guide provides recommendations on meeting the safety requirements concerning communication and consultation with the public and other interested parties by the regulatory body about the possible radiation risks associated with facilities and activities, and about processes and decisions of the regulatory body. The Safety Guide can be used by authorized parties in circumstances where there are regulatory requirements placed on them for communication and consultation. It may also be used by other organizations or individuals considering their responsibilities for communication and consultation with interested parties.

  7. Forklift safety a practical guide to preventing powered industrial truck incidents and injuries

    CERN Document Server

    Swartz, George

    1999-01-01

    Written for the more than 1.5 million powered industrial truck operators and supervisors in general industry, as well as those in the construction and marine industries, this Second Edition provides an updated guide to training operators in safety and complying with OSHA's 1999 forklift standard. This edition of Forklift Safety includes a new chapter devoted to the new OSHA 1910.178 standard and new information regarding dock safety, narrow aisle trucks, off-dock incidents, tip-over safety, pallet safety, and carbon monoxide.

  8. Teaching Safety and Health in the Workplace. An Instructor's Guide.

    Science.gov (United States)

    Occupational Safety and Health Administration, Washington, DC.

    The primary concern of the Occupational Safety and Health Act (OSHA) is to provide a safe and healthful workplace for every working man and woman in the nation. One way to help reduce the number of injuries and illnesses in the workplace is by training workers to be more aware of the job safety and health hazards and to teach them the methods of…

  9. Construction safety in DOE. Part 1, Students guide

    Energy Technology Data Exchange (ETDEWEB)

    Handwerk, E C

    1993-08-01

    This report is the first part of a compilation of safety standards for construction activities on DOE facilities. This report covers the following areas: general safety and health provisions; occupational health and environmental control/haz mat; personal protective equipment; fire protection and prevention; signs, signals, and barricades; materials handling, storage, use, and disposal; hand and power tools; welding and cutting; electrical; and scaffolding.

  10. Importance of modeling beam-column joints for seismic safety of reinforced concrete structures

    International Nuclear Information System (INIS)

    Sharma, Akanshu; Reddy, G.R.; Vaze, K.K.; Eligehausen, R.; Hofmann, J.

    2011-01-01

    Almost all structures, except the containment building, in a NPP can be classified as reinforced concrete (RC) framed structures. In case of such structures subjected to seismic loads, beam-column joints are recognized as the critical and vulnerable zone. During an earthquake, the global behavior of the structure is highly governed by the behavior of the joints. If the joints behave in a ductile manner, the global behavior generally will be ductile, whereas if the joints behave in a brittle fashion then the structure will display a brittle behavior. The joints of old and non-seismically detailed structures are more vulnerable and behave poorly under the earthquakes compared to the joints of new and seismically detailed structures. Modeling of these joint regions is very important for correct assessment of the seismic performance of the structures. In this paper, it is shown with the help of a recently developed joint model that not modeling the inelastic behavior of the joints can lead to significantly misleading and unsafe results in terms of the performance assessment of the structures under seismic loads. Comparison of analytical and experimental results is shown for two structures, tested under lateral monotonic seismic pushover loads. It is displayed that the model can predict the inelastic seismic response of structures considering joint distortion with high accuracy by little extra effort in modeling. (author)

  11. Approach of seismic upgrading in Kashiwazaki-Kariwa Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sato, Hitoshi

    2009-01-01

    Because guide for reviewing seismic design of nuclear power reactor facilities was reworked in 2006, we formulated new Design Base Seismic Motion Ss, and we are doing evaluation of seismic safety (back-check). In Japan, depending on aseismatic importance, equipments are classified into S-class, B-class and C-class. For S-class equipments, we evaluate it on the basis of new Ss, and do seismic upgrading. For B-class and C-class equipments, we do seismic upgrading voluntarily on the basis of the experiences of the Niigataken Chuetsu-Oki (NCO) Earthquake. (author)

  12. Research and development studies on the seismic behaviour of the PEC fast reactor (safety analysis detailed report no. 8)

    Energy Technology Data Exchange (ETDEWEB)

    Martelli, A.; Forni, M.; Masoni, P.; Maresca, G.; Castoldi, A.; Muzzi, F. [ENEA, Rome (Italy); Ansaldo Spa, Genoa [Italy; ISMES Spa, Bergamo [Italy

    1988-01-15

    This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA (Italian Commission for Alternative Energy Sources) for the seismic verification of the Italian PEC fast reactor test facility. More precisely, the paper focuses on the wide-ranging research and development programme that has been performed (and recently completed) on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the feed-backs on the design, necessary safisfy the seismic safety requirements, are recalled. The general validity of the analyses in the framework of the research and development activities for nuclear reactor is also pointed out.

  13. Task force activity to take the effect of elastic-plastic behaviour into account on the seismic safety evaluation of nuclear piping systems

    International Nuclear Information System (INIS)

    Nakamura, Izumi; Shiratori, Masaki; Morishita, Masaki; Otani, Akihito; Shibutani, Tadahito

    2015-01-01

    According to investigations of several nuclear power plants (NPPs) hit by actual seismic events and a number of experimental researches on the failure behavior of piping systems under seismic loads, it is recognized that piping systems used in NPPs include a large seismic safety margin until boundary failure. Since the stress assessment based on the elastic analysis does not reflect actual seismic capability of piping systems including plastic region, it is necessary to develop a rational procedures to estimate the elastic-plastic behavior of piping systems under a large seismic load. With the aim of establishing a procedure that takes into account the elastic-plastic behavior effect in the seismic safety estimation of nuclear piping systems, a task force activity has been planned. Through the activity, the authors intend to establish guidelines to estimate the elastic-plastic behavior of piping systems rationally and conservatively, and to provide new rational seismic safety criteria taking the effect of elastic-plastic behavior into account. As the first step of making out the analysis guideline, benchmark analyses are conducted for a pipe element test and a piping system test. In this paper, the outline of the research activity and the preliminary results of benchmark analyses are described. (author)

  14. Recommended safety guides for industrial laboratories and shops

    Science.gov (United States)

    Allison, W. W.

    1971-01-01

    Booklet provides references to 29 publications providing information on hazard control and approved safety practices. Areas include pressurized gas and vacuum systems. Guidelines are presented for safeguarding facilities where machinery, equipment, electrical devices, or hazardous chemicals are used.

  15. Evaluation of safety assessment methodologies in Rocky Flats Risk Assessment Guide (1985) and Building 707 Final Safety Analysis Report (1987)

    International Nuclear Information System (INIS)

    Walsh, B.; Fisher, C.; Zigler, G.; Clark, R.A.

    1990-01-01

    FSARs. Rockwell International, as operating contractor at the Rocky Flats plant, conducted a safety analysis program during the 1980s. That effort resulted in Final Safety Analysis Reports (FSARs) for several buildings, one of them being the Building 707 Final Safety Analysis Report, June 87 (707FSAR) and a Plant Safety Analysis Report. Rocky Flats Risk Assessment Guide, March 1985 (RFRAG85) documents the methodologies that were used for those FSARs. Resources available for preparation of those Rocky Flats FSARs were very limited. After addressing the more pressing safety issues, some of which are described below, the present contractor (EG ampersand G) intends to conduct a program of upgrading the FSARs. This report presents the results of a review of the methodologies described in RFRAG85 and 707FSAR and contains suggestions that might be incorporated into the methodology for the FSAR upgrade effort

  16. Efficacy And Safety Of IVUS-Guided Percutaneous Coronary Interventions

    Directory of Open Access Journals (Sweden)

    Popovic Marija

    2015-06-01

    Full Text Available The inclusion of IVUS-guided PCI has yet to become a routine approach in invasive cardiology due to the relatively high cost of the procedure, equivocal positive results in important studies and the steep learning curve. As an additional diagnostic tool, IVUS seems to be irreplaceable in stent apposition research, edge dissections and the determination of plaque composition.

  17. Building competence in radiation protection and the safe use of radiation sources. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    An essential element of a national infrastructure for radiation protection and safety is the maintenance of an adequate number of competent personnel. This Safety Guide makes recommendations concerning the building of competence in protection and safety, which relate to the training and assessment of qualification of new personnel and retraining of existing personnel in order to develop and maintain appropriate levels of competence. This Safety Guide addresses training in protection and safety aspects in relation to all practices and intervention situations in nuclear and radiation related technologies. This document covers the following aspects: the categories of persons to be trained. The requirements for education, training and experience for each category. The processes of qualification and authorization of persons. A national strategy for building competence

  18. Radiation protection aspects in the design of nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    . The IAEA takes seriously the enduring challenge for users and regulators everywhere: that of ensuring a high level of safety in the use of nuclear materials and radiation sources around the world. Their continuing utilization for the benefit of humankind must be managed in a safe manner, and the IAEA safety standards are designed to facilitate the achievement of that goal. This Safety Guide has been prepared as a part of the IAEA programme on safety standards for nuclear power plants. It includes recommendations on how to satisfy the requirements established in the Safety Requirements publication on the Safety of Nuclear Power Plants: Design. It addresses the provisions that should be made in the design of nuclear power plants in order to protect site personnel, the public and the environment against radiological hazards for operational states, decommissioning and accident conditions. The recommendations on radiation protection provided in this Safety Guide are consistent with the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS), which were jointly sponsored by the Food and Agriculture Organization of the United Nations (FAO), the IAEA, the International Labour Organisation (ILO), the OECD Nuclear Energy Agency (OECD/NEA), the Pan American Health Organization (PAHO) and the World Health Organization (WHO). This Safety Guide supersedes Safety Series No. 50-SG-D9, Design Aspects of Radiation Protection for Nuclear Power Plants, published in 1985. Effective radiation protection is a combination of good design, high quality construction and proper operation. Procedures that address the radiation protection aspects of operation are covered in the Safety Guide on Radiation Protection and Radioactive Waste Management in the operation of Nuclear Power Plants

  19. Radiation protection aspects of design for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    . The IAEA takes seriously the enduring challenge for users and regulators everywhere: that of ensuring a high level of safety in the use of nuclear materials and radiation sources around the world. Their continuing utilization for the benefit of humankind must be managed in a safe manner, and the IAEA safety standards are designed to facilitate the achievement of that goal. This Safety Guide has been prepared as a part of the IAEA programme on safety standards for nuclear power plants. It includes recommendations on how to satisfy the requirements established in the Safety Requirements publication on the Safety of Nuclear Power Plants: Design. It addresses the provisions that should be made in the design of nuclear power plants in order to protect site personnel, the public and the environment against radiological hazards for operational states, decommissioning and accident conditions. The recommendations on radiation protection provided in this Safety Guide are consistent with the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS), which were jointly sponsored by the Food and Agriculture Organization of the United Nations (FAO), the IAEA, the International Labour Organisation (ILO), the OECD Nuclear Energy Agency (OECD/NEA), the Pan American Health Organization (PAHO) and the World Health Organization (WHO). This Safety Guide supersedes Safety Series No. 50-SG-D9, Design Aspects of Radiation Protection for Nuclear Power Plants, published in 1985. Effective radiation protection is a combination of good design, high quality construction and proper operation. Procedures that address the radiation protection aspects of operation are covered in the Safety Guide on Radiation Protection and Radioactive Waste Management in the operation of Nuclear Power Plants

  20. Seismic qualification method of equipment for nuclear power plant

    International Nuclear Information System (INIS)

    Kim, J.S.; Choi, T.H.; Sulaimana, R.A.

    1995-01-01

    Safety related equipment installed in Korean Nuclear Power Plants are required to perform a safety function during and after a seismic event. To accomplish this safety function, they must be seismically qualified in accordance with the intent and requirements of the USNRC Reg. Guide 1.100 Rev. 02 and IEEE Std. 344-1987. This paper defines and summarizes acceptable criteria and procedures, based on the Korean experience, for seismic qualification of purchased equipment to be installed in a nuclear power plant. As such the paper is intended to be a concise reference by equipment designers, architectural engineering company and plant owners in uniform implementation of commitments to nuclear regulatory agencies such as the USNRC or Korea Institute of Nuclear Safety (KINS) relating to adequacy of seismic Category 1 equipment. Thus, the paper provides the methodologies which can be used for qualifying equipment for safely related service in Nuclear Power Plants in a cost effective manner

  1. Application of the management system for facilities and activities. Safety guide

    International Nuclear Information System (INIS)

    2006-01-01

    This Safety Guide supports the Safety Requirements publication on The Management System for Facilities and Activities. It provides generic guidance to aid in establishing, implementing, assessing and continually improving a management system that complies with the requirements established. In addition to this Safety Guide, there are a number of Safety Guides for specific technical areas. Together these provide all the guidance necessary for implementing these requirements. This publication supersedes Safety Series No. 50-SG-Q1-Q7 (1996). The guidance provided here may be used by organizations in the following ways: - To assist in the development of the management systems of organizations directly responsible for operating facilities and activities and providing services for: Nuclear facilities; Activities using sources of ionizing radiation; Radioactive waste management; The transport of radioactive material; Radiation protection activities; Any other practices or circumstances in which people may be exposed to radiation from naturally occurring or artificial sources; The regulation of such facilities and activities; - To assist in the development of the management systems of the relevant regulatory bodies; - By the operator, to specify to a supplier, via contractual documentation, any guidance of this Safety Guide that should be included in the supplier's management system for the supply and delivery of products

  2. Building competence in radiation protection and the safe use of radiation sources. Safety guide (Spanish ed.)

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide makes recommendations concerning the building of competence in protection and safety within a national radiation protection infrastructure and provides guidance for setting up the structure for a national strategy. It relates to the training and assessment of qualification of new personnel and the retraining of existing personnel in order to develop and maintain appropriate levels of competence. It provides the necessary guidance to meet the requirements laid down in Safety Series No. 115, International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. Contents: 1. Introduction; 2. Responsibilities for building competence in protection and safety; 3. Education, training and work experience; 4. A national strategy for building competence in protection and safety.

  3. Predisposal Management of Low and Intermediate Level Radioactive Waste. Safety Guide

    International Nuclear Information System (INIS)

    2009-01-01

    The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established for the predisposal management of low and intermediate level waste. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. General safety considerations; 5. Safety features for the predisposal management of LILW; 6. Record keeping and reporting; 7. Safety assessment; 8. Quality assurance; Annex I: Nature and sources of LILW from nuclear facilities; Annex II: Development of specifications for waste packages; Annex III: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Postulated initiating events for consideration in a safety assessment (internal phenomena).

  4. Building competence in radiation protection and the safe use of radiation sources. Safety guide (Arabic ed.)

    International Nuclear Information System (INIS)

    2006-01-01

    This Safety Guide makes recommendations concerning the building of competence in protection and safety within a national radiation protection infrastructure and provides guidance for setting up the structure for a national strategy. It relates to the training and assessment of qualification of new personnel and the retraining of existing personnel in order to develop and maintain appropriate levels of competence. It provides the necessary guidance to meet the requirements laid down in Safety Series No. 115, International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. Contents: 1. Introduction; 2. Responsibilities for building competence in protection and safety; 3. Education, training and work experience; 4. A national strategy for building competence in protection and safety.

  5. Management of Radioactive Waste from the Mining and Milling of Ores. Safety Guide (Spanish ed.)

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide provides recommendations and guidance on the safe management of radioactive waste resulting from the mining and milling of ores, with the purpose of protecting workers, the public and the environment from the consequences of these activities. It supplements Safety Standards Series No. WS-R-1, Near Surface Disposal of Radioactive Waste. Contents: 1. Introduction; 2. Administrative, legal and regulatory framework; 3. Protection of human health and the environment; 4. Strategy for waste management; 5. Safety considerations in different phases of operations; 6. Safety assessment; 7. Quality assurance; 8. Monitoring and surveillance; 9. Institutional control for the post-closure phase.

  6. Nuclear installations safety in France. Compilation of regulatory guides

    International Nuclear Information System (INIS)

    1988-01-01

    General plan: 1. General organization of public officials. Procedures 1.1. Texts defining the general organization and the procedures 1.2. Interventing organisms; 2. Texts presenting a technical aspect other than basic safety rules and associated organization texts; 2.1. Dispositions relating to safety of nuclear installations 2.2. Dispositions relating to pressure vessels 2.3. Dispositions relating to quality 2.4. Dispositions relating to radioactive wastes release 2.5. Dispositions relating to activities depending of classified installations; 3. Basic Safety Rules (BSR) 3.1. BSR relating to PWR 3.2. BSR relating to nuclear installations other than PWR 3.3. Other BSR [fr

  7. Radiation protection and safety guide no. GRPB-G-4: inspection

    International Nuclear Information System (INIS)

    Schandorf, C.; Darko, O.; Yeboah, J.; Osei, E.K.; Asiamah, S.D.

    1995-01-01

    The use of ionizing radiation and radiation sources in Ghana is on the increase due to national developmental efforts in Health Care, Food and Agriculture, Industry, Science and Technology. This regulatory Guide has been developed to assist both the Regulatory Body (Radiation Protection Board) and operating organizations to perform systematic inspections commensurate with the level of hazard associated with the application of radiation sources and radioactive materials. The present Guide applies to the Radiation Protection and Safety inspection and/or audit conducted by the Radiation Protection Board or Radiation Safety Officer. The present Guide is applicable in Ghana and to foreign suppliers of radiation sources. The present Guide applies to notifying person, licensee, or registrant and unauthorized practice

  8. Proceedings of the OECD/NEA workshop on seismic risk - Summary and conclusions - Committee on the Safety of Nuclear Installations PWG3 and PWG5

    International Nuclear Information System (INIS)

    2001-01-01

    The objectives of the Workshop were: - To provide a forum to review the recent advances in methodology and application of seismic probabilistic safety assessment and seismic margin analysis of nuclear installations, - To discuss the effective uses of the seismic PSA/margin analysis with consideration of merits and limitations of probabilistic methods, - To review the state of the art methodology to provide guidance for conducting seismic PSA, and - To discuss methodological issues and identify areas in which further research is needed for enhancing the usefulness of seismic PSA. The emphasis of the Workshop was placed on the exchange of ideas on effective ways of using seismic PSA rather than the numerical PSA results for specific plants such as core damage frequencies or seismic hazard. From the presentations and discussions in this workshop, it can be concluded that the seismic PSA/Margin methods have been and are being used world-wide, providing useful information for safety improvement or decision making, and great amount of experience has been accumulated, although the status of programs in member countries vary widely. The objectives of such studies include the following: - To examine whether there are cost effective ways to improve safety from ALARP point of view - To assist in decision making in backfitting by identifying cost effective improvements - To demonstrate the seismic margin of existing or future plants - To examine the vulnerabilities in protection against severe accident - To improve design of future reactors by identifying relatively weak points - To assist in selection of new sites for NPPs. Although numerical results from seismic PSA have not been directly used in seismic design as an alternate or supplement to current deterministic analysis methods, some countries have already adopted the use of probabilistic seismic hazard analysis for determining design basis earthquakes (SSE in USA) and some activities are ongoing to develop methods for

  9. Co-operative development of nuclear safety regulations, guides and standards based on NUSS

    International Nuclear Information System (INIS)

    Pachner, J.; Boyd, F.C.; Yaremy, E.M.

    1985-01-01

    A major need of developing Member States building nuclear power plants (NPPs) of foreign origin is to acquire a capability to regulate such nuclear plants independently. Among other things, this requires the development of national nuclear safety regulations, guides and standards to govern the development and use of nuclear technology. Recognizing the importance and complexity of this task, it seems appropriate that the NPP-exporting Member States share their experience and assist the NPP-importing Member States in the development of their national regulations and guides. In 1983, the Atomic Energy Control Board and Atomic Energy of Canada Ltd. conducted a study of a possible joint programme involving Canada, an NPP-importing Member State and the IAEA for the development of the national nuclear safety regulations and guides based on NUSS documents. During the study, a work plan with manpower estimates for the development of design regulations, safety guides and a guide for regulatory evaluation of design was prepared as an investigatory exercise. The work plan suggests that a successful NUSS implementation in developing Member States will require availability of significant resources at the start of the programme. The study showed that such a joint programme could provide an effective mechanism for transfer of nuclear safety know-how to the developing Member States through NUSS implementation. (author)

  10. A development of an evaluation flow chart for seismic stability of rock slopes based on relations between safety factor and sliding failure

    International Nuclear Information System (INIS)

    Kawai, Tadashi; Ishimaru, Makoto

    2010-01-01

    Recently, it is necessary to assess quantitatively seismic safety of critical facilities against the earthquake- induced rock slope failure from the viewpoint of seismic PSA. Under these circumstances, it is needed to evaluate the seismic stability of surrounding slopes against extremely strong ground motions. In order to evaluate the seismic stability of surrounding slopes, the most conventional method is to compare safety factors on an expected sliding surface, which is calculated from the stability analysis based on the limit equilibrium concept, to a critical value which judges stability or instability. The method is very effective to examine whether or not the sliding surface is safe. However, it does not mean that the sliding surface falls whenever the safety factor becomes smaller than the critical value during an earthquake. Therefore the authors develop a new evaluation flow chart for the seismic stability of rock slopes based on relations between safety factor and sliding failure. Furthermore, the developed flow chart was validated by comparing two kinds of safety factors calculated from a centrifuge test result concerned with a rock slope. (author)

  11. Predisposal management of low and intermediate level radioactive waste. Safety guide

    International Nuclear Information System (INIS)

    2003-01-01

    Radioactive waste is generated in the generation of electricity in nuclear power reactors and in the use of radioactive material in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized. The principles and requirements that govern the safety of the management of radioactive waste are presented in 'The Principles of Radioactive Waste Management', 'Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety' and 'Predisposal Management of Radioactive Waste, Including Decommissioning'. The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established in Refs for the predisposal management of LLW. This Safety Guide deals with the safety issues associated with the predisposal management of LLW from nuclear fuel cycle facilities, large research and development installations and radioisotope production facilities. This includes all steps and activities in the management of waste, from its initial generation to its final acceptance at a waste disposal facility or the removal of regulatory control. The predisposal management of radioactive waste includes decommissioning. The term 'decommissioning' encompasses both the process of decommissioning a facility and the management of the waste that results (prior to its disposal). Recommendations on the process of decommissioning are provided in Refs. Recommendations on the management of the waste resulting from decommissioning are included in this Safety Guide. Although the mining and milling of uranium and thorium ores is part of the nuclear fuel cycle, the management of the operational waste (e.g. waste rock, tailings and effluent treatment waste) from these activities is not within the scope of this Safety Guide. The LLW that is

  12. Walkdown procedure: Seismic adequacy review of safety class 3 ampersand 4 commodities in 2736-Z ampersand ZB buildings at PFP facility

    International Nuclear Information System (INIS)

    Ocoma, E.C.

    1995-01-01

    Seismic evaluation of existing safety class (SC) 3 and non-SC 4 commodities at the Plutonium Finishing Plant (PFP) is integrated into an area walkdown program. Field walkdowns of potential PFP seismic deficiencies associated with structural failure and falling will be performed using the DOE SQUG/EPRI methodology. Potential proximity interactions are also addressed. Objective of the walkdown is to qualify as much of the equipment as practical and to identify candidates for further evaluation

  13. The NUSS safety guides in design and the use of computers

    International Nuclear Information System (INIS)

    Fischer, J.

    1986-01-01

    After a brief summary of the NUSS programme, the two design guides are discussed which deal with instrumentation and control circuitry. The potential use of computers is covered differently in these guides because of the historical development and more importantly because of the difference in importance to safety of the I and C systems which are dealt with in these papers. The Agency would consider modifications to the existing guides only when sufficient consensus about the use of computers would warrant a revision of the documents. (author)

  14. Estimation of seismic velocity in the subducting crust of the Pacific slab beneath Hokkaido, northern Japan by using guided waves

    Science.gov (United States)

    Shiina, T.; Nakajima, J.; Toyokuni, G.; Kita, S.; Matsuzawa, T.

    2014-12-01

    A subducting crust contains a large amount of water as a form of hydrous minerals (e.g., Hacker et al., 2003), and the crust plays important roles for water transportation and seismogenesis in subduction zones at intermediate depths (e.g., Kirby et al., 1996; Iwamori, 2007). Therefore, the investigation of seismic structure in the crust is important to understand ongoing physical processes with subduction of oceanic lithosphere. A guided wave which propagates in the subducting crust is recorded in seismograms at Hokkaido, northern Japan (Shiina et al., 2014). Here, we estimated P- and S-wave velocity in the crust with guided waves, and obtained P-wave velocity of 6.6-7.3 km/s and S-wave velocity of 3.6-4.2 km/s at depths of 50-90 km. Moreover, Vp/Vs ratio in the crust is calculated to be 1.80-1.85 in that depth range. The obtained P-wave velocity about 6.6km/s at depths of 50-70 km is consistent with those estimated in Tohoku, northeast Japan (Shiina et al., 2013), and this the P-wave velocity is lower than those expected from models of subducting crustal compositions, such as metamorphosed MORB model (Hacker et al., 2003). In contrast, at greater depths (>80 km), the P-wave velocity marks higher velocity than the case of NE Japan and the velocity is roughly comparable to those of the MORB model. The obtained S-wave velocity distribution also shows characteristics similar to P waves. This regional variation may be caused by a small variation in thermal regime of the Pacific slab beneath the two regions as a result of the normal subduction in Tohoku and oblique subduction in Hokkaido. In addition, the effect of seismic anisotropy in the subducting crust would not be ruled out because rays used in the analysis in Hokkaido propagate mostly in the trench-parallel direction, while those in Tohoku are sufficiently criss-crossed.

  15. Comments on the seismic safety of nuclear power plants in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Tarics, A G [29 Winward Road, Belvedere, CA 94920 (United States); Kelly, J M [Earthquake Engineering Research Center, University of California, Berkeley, CA (United States); Csorba, E M [Technical University Vienna, Vienna (Austria)

    2001-03-01

    After the break-up of the Soviet Union, ten countries in Eastern Europe inherited Soviet-designed nuclear power plants which were constructed without adequate provisions to resist earthquake-generated lateral forces. An earthquake at their locations could seriously damage these plants and could result in Chernobyl-like consequences on the environment. There is an ongoing program to reinforce these plants using conventional piecemeal methods. A newly developed seismic protection strategy called 'base isolation' or 'seismic isolation', widely used in the United States to retrofit existing buildings, is recommended as an economical, technically superior, and more effective solution - where applicable - to make these nuclear power plants capable of resisting seismic forces. (author)

  16. Comments on the seismic safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Tarics, A.G.; Kelly, J.M.; Csorba, E.M.

    2001-01-01

    After the break-up of the Soviet Union, ten countries in Eastern Europe inherited Soviet-designed nuclear power plants which were constructed without adequate provisions to resist earthquake-generated lateral forces. An earthquake at their locations could seriously damage these plants and could result in Chernobyl-like consequences on the environment. There is an ongoing program to reinforce these plants using conventional piecemeal methods. A newly developed seismic protection strategy called 'base isolation' or 'seismic isolation', widely used in the United States to retrofit existing buildings, is recommended as an economical, technically superior, and more effective solution - where applicable - to make these nuclear power plants capable of resisting seismic forces. (author)

  17. Construction safety in DOE. Part 2, Students guide

    Energy Technology Data Exchange (ETDEWEB)

    Handwerk, E.C.

    1993-08-01

    This report is the second part of a compilation of safety standards for construction activities on DOE facilities. This report covers the following areas: floor and wall openings; cranes, derricks, hoists, elevators, and conveyors; motor vehicles, mechanized equipment, and marine operations; excavations; concrete and masonry construction; steel erection; underground construction, caisson, cofferdams, and compressed air; demolition; blasting and the use of explosives; power transmission and distribution; rollover protective structures, overhead protection; and ladders.

  18. The operating organization and the recruitment, training and qualification of personnel for research reactors. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide provides recommendations on meeting the requirements on the operating organization and on personnel for research reactors. It covers the typical operating organization for research reactor facilities; the recruitment process and qualification in terms of education, training and experience; programmes for initial and continuing training; the authorization process for those individuals having an immediate bearing on safety; and the processes for their requalification and reauthorization

  19. A procedure for seismic risk reduction in Campania Region

    Science.gov (United States)

    Zuccaro, G.; Palmieri, M.; Maggiò, F.; Cicalese, S.; Grassi, V.; Rauci, M.

    2008-07-01

    The Campania Region has set and performed a peculiar procedure in the field of seismic risk reduction. Great attention has been paid to public strategic buildings such as town halls, civil protection buildings and schools. The Ordinance 3274 promulgate in the 2004 by the Italian central authority obliged the owners of strategic buildings to perform seismic analyses within 2008 in order to check the safety of the structures and the adequacy to the use. In the procedure the Campania region, instead of the local authorities, ensure the complete drafting of seismic checks through financial resources of the Italian Government. A regional scientific technical committee has been constituted, composed of scientific experts, academics in seismic engineering. The committee has drawn up guidelines for the processing of seismic analyses. At the same time, the Region has issued a public competition to select technical seismic engineering experts to appoint seismic analysis in accordance with guidelines. The scientific committee has the option of requiring additional documents and studies in order to approve the safety checks elaborated. The Committee is supported by a technical and administrative secretariat composed of a group of expert in seismic engineering. At the moment several seismic safety checks have been completed. The results will be presented in this paper. Moreover, the policy to mitigate the seismic risk, set by Campania region, was to spend the most of the financial resources available on structural strengthening of public strategic buildings rather than in safety checks. A first set of buildings of which the response under seismic action was already known by data and studies of vulnerability previously realised, were selected for immediate retrofitting designs. Secondly, an other set of buildings were identified for structural strengthening. These were selected by using the criteria specified in the Guide Line prepared by the Scientific Committee and based on

  20. A procedure for seismic risk reduction in Campania Region

    International Nuclear Information System (INIS)

    Zuccaro, G.; Palmieri, M.; Cicalese, S.; Grassi, V.; Rauci, M.; Maggio, F.

    2008-01-01

    The Campania Region has set and performed a peculiar procedure in the field of seismic risk reduction. Great attention has been paid to public strategic buildings such as town halls, civil protection buildings and schools. The Ordinance 3274 promulgate in the 2004 by the Italian central authority obliged the owners of strategic buildings to perform seismic analyses within 2008 in order to check the safety of the structures and the adequacy to the use. In the procedure the Campania region, instead of the local authorities, ensure the complete drafting of seismic checks through financial resources of the Italian Government. A regional scientific technical committee has been constituted, composed of scientific experts, academics in seismic engineering. The committee has drawn up guidelines for the processing of seismic analyses. At the same time, the Region has issued a public competition to select technical seismic engineering experts to appoint seismic analysis in accordance with guidelines. The scientific committee has the option of requiring additional documents and studies in order to approve the safety checks elaborated. The Committee is supported by a technical and administrative secretariat composed of a group of expert in seismic engineering. At the moment several seismic safety checks have been completed. The results will be presented in this paper. Moreover, the policy to mitigate the seismic risk, set by Campania region, was to spend the most of the financial resources available on structural strengthening of public strategic buildings rather than in safety checks. A first set of buildings of which the response under seismic action was already known by data and studies of vulnerability previously realised, were selected for immediate retrofitting designs. Secondly, an other set of buildings were identified for structural strengthening. These were selected by using the criteria specified in the Guide Line prepared by the Scientific Committee and based on

  1. Review and assessment of nuclear facilities by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on reviewing and assessing the various safety related submissions made by the operator of a nuclear facility at different stages (siting, design, construction, commissioning, operation and decommissioning or closure) in the facility's lifetime to determine whether the facility complies with the applicable safety objectives and requirements. This Safety Guide covers the review and assessment of submissions in relation to the safety of nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Objectives, management, planning and organizational matters relating to the review and assessment process are presented in Section 2. Section 3 deals with the bases for decision making and conduct of the review and assessment process. Section 4 covers aspects relating to the assessment of this process. The Appendix provides a generic list of topics to be covered in the review and assessment process

  2. Review and assessment of nuclear facilities by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on reviewing and assessing the various safety related submissions made by the operator of a nuclear facility at different stages (siting, design, construction, commissioning, operation and decommissioning or closure) in the facility's lifetime to determine whether the facility complies with the applicable safety objectives and requirements. This Safety Guide covers the review and assessment of submissions in relation to the safety of nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Objectives, management, planning and organizational matters relating to the review and assessment process are presented in Section 2. Section 3 deals with the bases for decision making and conduct of the review and assessment process. Section 4 covers aspects relating to the assessment of this process. The Appendix provides a generic list of topics to be covered in the review and assessment process

  3. Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    The IAEA's Statute authorizes the Agency to establish safety standards to protect health and minimize danger to life and property - standards which the IAEA must use in its own operations, and which a State can apply by means of its regulatory provisions for nuclear and radiation safety. A comprehensive body of safety standards under regular review, together with the IAEA's assistance in their application, has become a key element in a global safety regime. In the mid-1990s, a major overhaul of the IAEA's safety standards programme was initiated, with a revised oversight committee structure and a systematic approach to updating the entire corpus of standards. The new standards that have resulted are of a high calibre and reflect best practices in Member States. With the assistance of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its safety standards. Safety standards are only effective, however, if they are properly applied in practice. The IAEA's safety services - which range in scope from engineering safety, operational safety, and radiation, transport and waste safety to regulatory matters and safety culture in organizations - assist Member States in applying the standards and appraise their effectiveness. These safety services enable valuable insights to be shared and I continue to urge all Member States to make use of them. Regulating nuclear and radiation safety is a national responsibility, and many Member States have decided to adopt the IAEA's safety standards for use in their national regulations. For the contracting parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions. The standards are also applied by designers, manufacturers and operators around the world to enhance nuclear and radiation safety in power generation, medicine, industry, agriculture, research and education

  4. Comparison and Analysis of IEEE 344 and IEC 60980 standards for harmonization of seismic qualification of safety-related equipment

    International Nuclear Information System (INIS)

    Lee, Young Ok; Kim, Jong Seog; Seo, Jeong Ho; Kim, Myung Jun

    2011-01-01

    The seismic qualification of safety related equipment in nuclear power plants should demonstrate an equipment's ability to perform its safety function during/or after the time it is subjected to the forces resulting from one SSE. In addition, the equipment must withstand the effects of a number of OBEs, preceding the SSE. IEEE 344 and IEC 60980 present the criteria for establishing procedures demonstrating that the Class 1E equipment can meet its performance requirement during seismic events. Currently, IEEE 344 is used for regulation of nuclear power plant in the United State whereas IEC 60980 is mainly used in Europe. In particular, NPPs of France and China apply with RCC-E and GB that are domestic standards, respectively. Equipment supplier and Utility have difficulties because of different applicable standards. Equipment supplier to export S/R components/equipment to other standard area performs additional seismic qualification. For example, equipment are qualifies according to IEC 60980, RCC-E, GB although they have been qualified in accordance with IEEE 344. Also, utility to attempt power up-rate, life extension of NPP constructed under rules of RCC-E such as Ulchin NPP 1 and 2 has similar difficulties. RCC-E endorses IEC 60980 and GB is almost same as IEC 60980 except minor difference of earthquake environment definition. Therefore this paper surveys the similarities and differences between IEEE 344 and IEC 60980. In addition, this paper considers how the two sets of standards may be used in a complementary fashion to be possible using one or the other standard area

  5. Safety Software Guide Perspectives for the Design of New Nuclear Facilities (U)

    International Nuclear Information System (INIS)

    VINCENT, Andrew

    2005-01-01

    In June of this year, the Department of Energy (DOE) issued directives DOE O 414.1C and DOE G 414.1-4 to improve quality assurance programs, processes, and procedures among its safety contractors. Specifically, guidance entitled, ''Safety Software Guide for use with 10 CFR 830 Subpart A, Quality Assurance Requirements, and DOE O 414.1C, Quality Assurance, DOE G 414.1-4'', provides information and acceptable methods to comply with safety software quality assurance (SQA) requirements. The guidance provides a roadmap for meeting DOE O 414.1C, ''Quality Assurance'', and the quality assurance program (QAP) requirements of Title 10 Code of Federal Regulations (CFR) 830, Subpart A, Quality Assurance, for DOE nuclear facilities and software application activities. [1, 2] The order and guide are part of a comprehensive implementation plan that addresses issues and concerns documented in Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1. [3] Safety SQA requirements for DOE as well as National Nuclear Security Administration contractors are necessary to implement effective quality assurance (QA) processes and achieve safe nuclear facility operations. DOE G 414.1-4 was developed to provide guidance on establishing and implementing effective QA processes tied specifically to nuclear facility safety software applications. The Guide includes software application practices covered by appropriate national and international consensus standards and various processes currently in use at DOE facilities. While the safety software guidance is considered to be of sufficient rigor and depth to ensure acceptable reliability of safety software at all DOE nuclear facilities, new nuclear facilities are well suited to take advantage of the guide to ensure compliant programs and processes are implemented. Attributes such as the facility life-cycle stage and the hazardous nature of each facility operations are considered, along with the category and level of importance of the

  6. Hazard screening application guide. Safety Analysis Report Update Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1992-06-01

    The basic purpose of hazard screening is to group precesses, facilities, and proposed modifications according to the magnitude of their hazards so as to determine the need for and extent of follow on safety analysis. A hazard is defined as a material, energy source, or operation that has the potential to cause injury or illness in human beings. The purpose of this document is to give guidance and provide standard methods for performing hazard screening. Hazard screening is applied to new and existing facilities and processes as well as to proposed modifications to existing facilities and processes. The hazard screening process evaluates an identified hazards in terms of the effects on people, both on-site and off-site. The process uses bounding analyses with no credit given for mitigation of an accident with the exception of certain containers meeting DOT specifications. The process is restricted to human safety issues only. Environmental effects are addressed by the environmental program. Interfaces with environmental organizations will be established in order to share information.

  7. Assessment of occupational exposure due to external sources of radiation. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

  8. Assessment of occupational exposure due to external sources of radiation. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

  9. Assessment of occupational exposure due to external sources of radiation. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

  10. IAEA activities to prepare safety codes and guides for thermal neutron nuclear power plants

    International Nuclear Information System (INIS)

    Iansiti, E.

    1977-01-01

    In accordance with the programme presented to, and endorsed by, the eighteenth General Conference in September 1974, the IAEA is now developing a complete set of safety codes and guides that will represent recommendations for the safety of thermal neutron power plants. The safety codes outline the minimum requirements for achieving this safety, and the safety guides set forth the criteria, procedures and methods to implement the safety codes. The whole programme is directed towards the five areas of Governmental Organization, Siting, Design, Operation, and Quality Assurance. One Scientific Secretary from the Agency Secretariat is responsible for each of these areas and a Co-ordinator takes care of common problems. For the development of each of these documents a working group of a few world experts is first convened which prepare a preliminary draft. This draft is then reviewed by a larger, international Technical Review Committee (one for each of the five areas) and a subsequent review by the Senior Advisory Group - with representatives from 20 states - ensures that the document is well coordinated within the programme. At this stage, it is sent to Member States for comments. The Technical Review Committee concerned is reconvened to integrate these comments into the document, and, after a final review by the Senior Advisory Group, the document is ready for transmission to the Director General of the Agency for endorsement and publication. A preliminary to this procedure is the collation by the Secretariat of large amounts of information submitted by Member States so that the first draft is really based on a very complete knowledge of what is done in each area all over the world. This collation frequently reveals differences in approach which are not random but due, rather, to the local conditions and the types of reactors. These differences must be harmonized in the documents produced without detracting from the effectiveness of the code or guide. The whole

  11. Recruitment, qualification and training of personnel for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The objective of this Safety Guide is to outline the various factors that should to be considered in order to ensure that the operating organization has a sufficient number of qualified personnel for safe operation of a nuclear power plant. In particular, the objective of this publication is to provide general recommendations on the recruitment and selection of plant personnel and on the training and qualification practices that have been adopted in the nuclear industry since the predecessor Safety Guide was published in 1991. In addition, this Safety Guide seeks to establish a framework for ensuring that all managers and staff employed at a nuclear power plant demonstrate their commitment to the management of safety to high professional standards. This Safety Guide deals specifically with those aspects of qualification and training that are important to the safe operation of nuclear power plants. It provides recommendations on the recruitment, selection, qualification, training and authorization of plant personnel. That is, of all personnel in all safety related functions and at all levels of the plant. Some parts or all of this Safety Guide may also be used, with due adaptation, as a guide to the recruitment, selection, training and qualification of staff for other nuclear installations (such as research reactors or nuclear fuel cycle facilities). Section 2 gives guidance on the recruitment and selection of suitable personnel for a nuclear power plant. Section 3 gives guidance on the establishment of personnel qualification, explains the relationship between qualification and competence, and identifies how competence may be developed through education, experience and training. Section 4 deals with general aspects of the training policy for nuclear power plant personnel: the systematic approach, training settings and methods, initial and continuing training, and the keeping of training records. Section 5 provides guidance on the main aspects of training programmes

  12. Recruitment, qualification and training of personnel for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    The objective of this Safety Guide is to outline the various factors that should to be considered in order to ensure that the operating organization has a sufficient number of qualified personnel for safe operation of a nuclear power plant. In particular, the objective of this publication is to provide general recommendations on the recruitment and selection of plant personnel and on the training and qualification practices that have been adopted in the nuclear industry since the predecessor Safety Guide was published in 1991. In addition, this Safety Guide seeks to establish a framework for ensuring that all managers and staff employed at a nuclear power plant demonstrate their commitment to the management of safety to high professional standards. This Safety Guide deals specifically with those aspects of qualification and training that are important to the safe operation of nuclear power plants. It provides recommendations on the recruitment, selection, qualification, training and authorization of plant personnel; that is, of all personnel in all safety related functions and at all levels of the plant. Some parts or all of this Safety Guide may also be used, with due adaptation, as a guide to the recruitment, selection, training and qualification of staff for other nuclear installations (such as research reactors or nuclear fuel cycle facilities). Section 2 gives guidance on the recruitment and selection of suitable personnel for a nuclear power plant. Section 3 gives guidance on the establishment of personnel qualification, explains the relationship between qualification and competence, and identifies how competence may be developed through education, experience and training. Section 4 deals with general aspects of the training policy for nuclear power plant personnel: the systematic approach, training settings and methods, initial and continuing training, and the keeping of training records. Section 5 provides guidance on the main aspects of training programmes

  13. Havsnaes Wind Farm. A guide on Health and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Wettin, Martina [Goovinn, Goeteborg (Sweden); Jiven, Anna [Nordisk Vindkraft, Goeteborg (Sweden)

    2011-07-01

    This guide has been developed by Nordisk Vindkraft in co-operation with Energimyndigheten (the Swedish Energy Agency). The main purpose is to raise the attention to work environment matters when developing, constructing and operating large-scale wind farms in the northern parts of Sweden. The Swedish wind industry is relatively young. Work Environment Management and legal awareness has been identified by the Swedish Government as one crucial factor for ensuring sound and rapid development of the industry. Havsnaes is to date the largest onshore wind farm in Sweden. Whilst developing this project, Nordisk Vindkraft has gained extensive experience and developed a profound understanding of the prerequisites for the Swedish large-scale wind industry. The wind farm's size, complexity and geographical location near Stroemsund in Jaemtland qualify the project as an excellent pilot study. Preventing accidents and ill-health largely depends on your knowledge, competence and the resources you are able and willing to spend. The key for success is a pro-active approach. During the Havsnaes project, Nordisk Vindkraft has spent considerable amount of time and effort on the management of Work Environment. We hope our gained experience will be beneficial for our industry as a whole.

  14. Planning and Preparing for Emergency Response to Transport Accidents Involving Radioactive Material. Safety Guide

    International Nuclear Information System (INIS)

    2009-01-01

    This Safety Guide provides guidance on various aspects of emergency planning and preparedness for dealing effectively and safely with transport accidents involving radioactive material, including the assignment of responsibilities. It reflects the requirements specified in Safety Standards Series No. TS-R-1, Regulations for the Safe Transport of Radioactive Material, and those of Safety Series No. 115, International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. Contents: 1. Introduction; 2. Framework for planning and preparing for response to accidents in the transport of radioactive material; 3. Responsibilities for planning and preparing for response to accidents in the transport of radioactive material; 4. Planning for response to accidents in the transport of radioactive material; 5. Preparing for response to accidents in the transport of radioactive material; Appendix I: Features of the transport regulations influencing emergency response to transport accidents; Appendix II: Preliminary emergency response reference matrix; Appendix III: Guide to suitable instrumentation; Appendix IV: Overview of emergency management for a transport accident involving radioactive material; Appendix V: Examples of response to transport accidents; Appendix VI: Example equipment kit for a radiation protection team; Annex I: Example of guidance on emergency response to carriers; Annex II: Emergency response guide.

  15. Compliance assurance for the safe transport of radioactive material. Safety guide

    International Nuclear Information System (INIS)

    2009-01-01

    The objectives of this Safety Guide are to assist competent authorities in the development and maintenance of compliance assurance programmes in connection with the transport of radioactive material, and to assist applicants, licensees and organizations in their interactions with competent authorities. In order to increase cooperation between competent authorities and to promote the uniform application of international regulations and recommendations, it is desirable to adopt a common approach to regulatory activities. This Safety Guide is intended to assist in accomplishing such a uniform application by recommending most of the actions for which competent authorities need to provide in their programmes for ensuring compliance with the Transport Regulations. This Safety Guide addresses radiation safety aspects of the transport of radioactive material; that is, the subjects that are covered by the Transport Regulations. Radioactive material may have other dangerous properties, however, such as explosiveness, flammability, pyrophoricity, chemical toxicity and corrosiveness; these properties are required to be taken into account in the regulatory control of the design and transport of packages. Physical protection and systems for accounting for and control of nuclear material are also discussed in this Safety Guide. These subjects are not within the scope of the Transport Regulations, but information on them is included here because they must be taken into account in the overall regulatory control of transport, especially when the regulatory framework is being established. Section 1 informs about the background, the objective, the scope and the structure of this publication. Section 2 provides recommendations on the responsibilities and functions of the competent authority. Section 3 provides information on the various national and international regulations and guides for the transport of radioactive material. Section 4 provides recommendations on carrying out

  16. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  17. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  18. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  19. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  20. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  1. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.

  2. Health and Safety Guide for Home Performance Contractors

    Energy Technology Data Exchange (ETDEWEB)

    Stratton, Chris [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Walker, Iain S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2012-02-15

    This report is intended to provide home performance contractor trainers with a resource to keep both their workers and home residents safe and healthy. This document is an attempt to describe what we currently believe is safe, what we believe is unsafe, and what we’re unsure about. It is intended to identify health and safety issues and provide historical context and current understanding of both risks and mitigation strategies. In addition, it provides links to more in-depth resources for each issue. When we tighten the thermal envelope of a house to improve comfort and reduce energy use, we have to be sure that we are not compromising the indoor air quality of the home. This means identifying and mitigating or eliminating pollution sources before and after you make changes to the home. These sources can include materials and finishes in the home, exhaust gasses from combustion appliances, soil gasses such as radon, and moisture from a bathroom, kitchen, or unvented clothes dryer. Our first responsibility is to do no harm — this applies both to our clients and to our employees. Currently, there are many new products that are widely used but whose health effects are not well understood. Our in ability to have perfect information means the directive to do no harm can be difficult to obey. Each home is a little bit different, and in the face of a situation you’ve never encountered, it’s important to have a solid grasp of the fundamental concepts of building science when the hard and fast rules don’t apply . The home performance industry is gaining momentum, and has the potential to expand greatly as energy costs continue to rise. It is imperative that we remain vigilant about protecting the health and safety of our workers and our customers. It only takes a few news stories about a family that got sick after their home was tightened by a home performance contractor to scare off potential customers and taint the reputation of the entire industry. Good

  3. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  4. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  5. Regulatory inspection of nuclear facilities and enforcement by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on the inspection of nuclear facilities, regulatory enforcement and related matters. The objective is to provide the regulatory body with a high level of confidence that operators have the processes in place to ensure compliance and that they do comply with legal requirements, including meeting the safety objectives and requirements of the regulatory body. However, in the event of non-compliance, the regulatory body should take appropriate enforcement action. This Safety Guide covers regulatory inspection and enforcement in relation to nuclear facilities such as: enrichment and fuel manufacturing plants; nuclear power plants; other reactors such as research reactors and critical assemblies; spent fuel reprocessing plants; and facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Section 2 sets out the objectives of regulatory inspection and enforcement. Section 3 covers the management of regulatory inspections. Section 4 covers the performance of regulatory inspections, including internal guidance, planning and preparation, methods of inspection and reports of inspections. Section 5 deals with regulatory enforcement actions. Section 6 covers the assessment of regulatory inspections and enforcement activities. The Appendix provides further details on inspection areas for nuclear facilities

  6. Safety measurement and monitoring in healthcare: a framework to guide clinical teams and healthcare organisations in maintaining safety

    Science.gov (United States)

    Vincent, Charles; Burnett, Susan; Carthey, Jane

    2014-01-01

    Patients, clinicians and managers all want to be reassured that their healthcare organisation is safe. But there is no consensus about what we mean when we ask whether a healthcare organisation is safe or how this is achieved. In the UK, the measurement of harm, so important in the evolution of patient safety, has been neglected in favour of incident reporting. The use of softer intelligence for monitoring and anticipation of problems receives little mention in official policy. The Francis Inquiry report into patient treatment at the Mid Staffordshire NHS Foundation Trust set out 29 recommendations on measurement, more than on any other topic, and set the measurement of safety an absolute priority for healthcare organisations. The Berwick review found that most healthcare organisations at present have very little capacity to analyse, monitor or learn from safety and quality information. This paper summarises the findings of a more extensive report and proposes a framework which can guide clinical teams and healthcare organisations in the measurement and monitoring of safety and in reviewing progress against safety objectives. The framework has been used so far to promote self-reflection at both board and clinical team level, to stimulate an organisational check or analysis in the gaps of information and to promote discussion of ‘what could we do differently’. PMID:24764136

  7. Seismic safety margins research program. Phase I final report - Plant/site selection and data collection (Project I)

    International Nuclear Information System (INIS)

    Chuang, T.Y.

    1981-07-01

    Project I of Phase I of the Seismic Safety Margins Research Program (SSMRP) comprised two parts: the selection of a representative nuclear power plant/site for study in Phase I and the collection of data needed by the other SSMRP projects. Unit 1 of the Zion Nuclear Power Plant in Zion, Illinois, was selected for the SSMRP Phase I studies. Unit 1 of the Zion plant has been validated as a good choice for the Phase I study plant. Although no single nuclear power plant can represent all such plants equally well, selection criteria were developed to maximize the generic implications of Phase I of the SSMRP. On the basis of the selection criteria, the Zion plant and its site were found to be reasonably representative of operating and future plants with regard to its nuclear steam supply system; the type of containment structure (prestressed concrete); its electrical capacity (1100 MWe); its location (the Midwest); the peak seismic acceleration used for design (0.17g); and the properties of the underlying soil (the low-strain shear-wave velocity is 1650 ft/s in a 50- to 100-ft-thick layer of soil overlying sedimentary bedrock). (author)

  8. Maintenance, surveillance and in-service inspection in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    Effective maintenance, surveillance and in-service inspection (MS and I) are essential for the safe operation of a nuclear power plant. The objective of this Safety Guide is to provide recommendations and guidance for MS and I activities to ensure that SSCs important to safety are available to perform their functions in accordance with the assumptions and intent of the design. This Safety Guide covers the organizational and procedural aspects of MS and I. However, it does not give detailed technical advice in relation to particular items of plant equipment, nor does it cover inspections made for and/or by the regulatory body. This Safety Guide provides recommendations and guidance for preventive and remedial measures, including testing, surveillance and in-service inspection, that are necessary to ensure that all plant structures, systems and components (SSCs) important to safety are capable of performing as intended. This Safety Guide covers measures for fulfilling the organizational and administrative requirements for: establishing and implementing schedules for preventive and predictive maintenance, repairing defective plant items, selecting and training personnel, providing related facilities and equipment, procuring stores and spare parts, and generating, collecting and retaining maintenance records for establishing and implementing an adequate feedback system for information on maintenance. MS and I should be subject to quality assurance in relation to all aspects important to safety. Quality assurance has been dealt with in detail in other IAEA safety standards and is covered here only in specific instances, for emphasis. In Section 2, a concept of MS and I is presented and the interrelationship between maintenance, surveillance and inspection is discussed. Section 3 concerns the functions and responsibilities of different organizations involved in MS and I activities. Section 4 provides recommendations and guidance on such organizational aspects as

  9. Maintenance, surveillance and in-service inspection in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Effective maintenance, surveillance and in-service inspection (MS and I) are essential for the safe operation of a nuclear power plant. The objective of this Safety Guide is to provide recommendations and guidance for MS and I activities to ensure that SSCs important to safety are available to perform their functions in accordance with the assumptions and intent of the design. This Safety Guide covers the organizational and procedural aspects of MS and I. However, it does not give detailed technical advice in relation to particular items of plant equipment, nor does it cover inspections made for and/or by the regulatory body. This Safety Guide provides recommendations and guidance for preventive and remedial measures, including testing, surveillance and in-service inspection, that are necessary to ensure that all plant structures, systems and components (SSCs) important to safety are capable of performing as intended. This Safety Guide covers measures for fulfilling the organizational and administrative requirements for: establishing and implementing schedules for preventive and predictive maintenance, repairing defective plant items, selecting and training personnel, providing related facilities and equipment, procuring stores and spare parts, and generating, collecting and retaining maintenance records for establishing and implementing an adequate feedback system for information on maintenance. MS and I should be subject to quality assurance in relation to all aspects important to safety. Quality assurance has been dealt with in detail in other IAEA safety standards and is covered here only in specific instances, for emphasis. In Section 2, a concept of MS and I is presented and the interrelationship between maintenance, surveillance and inspection is discussed. Section 3 concerns the functions and responsibilities of different organizations involved in MS and I activities. Section 4 provides recommendations and guidance on such organizational aspects as

  10. Material presented to advisory committee on reactor safeguards, subcommittee on extreme external phenomena, January 29-30, 1981, Los Angeles, California. Seismic safety margins research program

    International Nuclear Information System (INIS)

    Smith, P.D.; Bernreuter, D.L.; Bohn, M.P.; Chuang, T.Y.; Cummings, G.E.; Dong, R.G.; Johnson, J.J.; Wells, J.E.

    1981-01-01

    The January 29-30, 1981, meeting of the Advisory Committee on Reactor Safeguards (ACRS), Subcommittee on Extreme External Phenomena, mark the close of Phase I efforts on the Seismic Safety Margins Research Program (SSMRP). Presentations at the meeting focused on results produced. These included computer codes, response computations, failure and release probabilities, data bases, and fragilities and parameter characteristics

  11. Seismic safety margins research program. Phase I. Project VII: systems analysis specifications of computational approach

    International Nuclear Information System (INIS)

    Collins, J.D.; Hudson, J.M.; Chrostowski, J.D.

    1979-02-01

    A computational methodology is presented for the prediction of core melt probabilities in a nuclear power plant due to earthquake events. The proposed model has four modules: seismic hazard, structural dynamic (including soil-structure interaction), component failure and core melt sequence. The proposed modules would operate in series and would not have to be operated at the same time. The basic statistical approach uses a Monte Carlo simulation to treat random and systematic error but alternate statistical approaches are permitted by the program design

  12. Subsystem response determination for the US NRC Seismic Safety Margins Research Program

    International Nuclear Information System (INIS)

    Johnson, J.J.

    1979-01-01

    The initial portion of the task described deals with a definition of the state-of-the-art of seismic qualification methods for subsystems. Too facilitate treatment of this broad class of subsystems, three classifications have been identified: multiply supported subsystems (e.g., piping systems); mechanical components (e.g., valves, pumps, control rod drives, hydraulic systems, etc.); and electrical components (e.g., electrical control panels). Descriptions of the available analysis and/or testing techniques for the above classifications are sought. The results of this assessment will be applied to the development of structural subsystem transfer functions

  13. The SAFER guides: empowering organizations to improve the safety and effectiveness of electronic health records.

    Science.gov (United States)

    Sittig, Dean F; Ash, Joan S; Singh, Hardeep

    2014-05-01

    Electronic health records (EHRs) have potential to improve quality and safety of healthcare. However, EHR users have experienced safety concerns from EHR design and usability features that are not optimally adapted for the complex work flow of real-world practice. Few strategies exist to address unintended consequences from implementation of EHRs and other health information technologies. We propose that organizations equipped with EHRs should consider the strategy of "proactive risk assessment" of their EHR-enabled healthcare system to identify and address EHR-related safety concerns. In this paper, we describe the conceptual underpinning of an EHR-related self-assessment strategy to provide institutions a foundation upon which they could build their safety efforts. With support from the Office of the National Coordinator for Health Information Technology (ONC), we used a rigorous, iterative process to develop a set of 9 self-assessment tools to optimize the safety and safe use of EHRs. These tools, referred to as the Safety Assurance Factors for EHR Resilience (SAFER) guides, could be used to self-assess safety and effectiveness of EHR implementations, identify specific areas of vulnerability, and create solutions and culture change to mitigate risks. A variety of audiences could conduct these assessments, including frontline clinicians or care teams in different practices, or clinical, quality, or administrative leaders within larger institutions. The guides use a multifaceted systems-based approach to assess risk and empower organizations to work with internal or external stakeholders (eg, EHR developers) on optimizing EHR functionality and using EHRs to drive improvements in the quality and safety of healthcare.

  14. Inspection and enforcement by the regulatory body for nuclear power plants. A safety guide. A publication within the NUSS programme

    International Nuclear Information System (INIS)

    1996-01-01

    The purpose of this Safety Guide is to provide guidance on fulfilling the requirements for inspection and enforcement by the regulatory body, as set out in the Code on the Safety of Nuclear Power Plants; Governmental Organization. This Safety Guide deals with the responsibilities of the regulatory body, the organization of inspection programmes, the inspection resources of the regulatory body, methods of inspection, requirements on the applicant/licensee in regard to regulatory inspection, inspection reports, and regulatory action and enforcement. It is recognized that many of the provisions of this Safety Guide may be applicable to the regulations of other nuclear facilities and related activities including research reactors, fuel processing and manufacturing plants, irradiated fuel processing plants and radioactive waste management facilities. This Safety Guide does not deal specifically with the functions of a regulatory body responsible for such matters; however, the guidance presented here may be applied as appropriate to these activities. 11 refs, 1 fig

  15. Seismic design of equipment and piping systems for nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Minematsu, Akiyoshi

    1997-01-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on 'Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981' (referred to as 'Examination Guide' hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in 'Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association'. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  16. Seismic design of equipment and piping systems for nuclear power plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

    1997-03-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  17. Conduct of Operations at Nuclear Power Plants. Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide identifies the main responsibilities and practices of nuclear power plant (NPP) operations departments in relation to their responsibility for the safe functioning of the plant. The guide presents the factors to be considered in structuring the operations department of an NPP; setting high standards of performance; making safety related decisions in an effective manner; conducting control room and field activities in a thorough and professional manner; and maintaining an NPP within established operational limits and conditions. Contents: 1. Introduction; 2. Management and organization of plant operations; 3. Shift complement and functions; 4. Shift routines and operating practices; 5. Control of equipment and plant status; 6. Operations equipment and operator aids; 7. Work control and authorization.

  18. Predisposal Management of Radioactive Waste from Nuclear Fuel Cycle Facilities. Specific Safety Guide

    International Nuclear Information System (INIS)

    2016-01-01

    This Safety Guide provides guidance on the predisposal management of all types of radioactive waste (including spent nuclear fuel declared as waste and high level waste) generated at nuclear fuel cycle facilities. These waste management facilities may be located within larger facilities or may be separate, dedicated waste management facilities (including centralized waste management facilities). The Safety Guide covers all stages in the lifetime of these facilities, including their siting, design, construction, commissioning, operation, and shutdown and decommissioning. It covers all steps carried out in the management of radioactive waste following its generation up to (but not including) disposal, including its processing (pretreatment, treatment and conditioning). Radioactive waste generated both during normal operation and in accident conditions is considered

  19. External Events Excluding Earthquakes in the Design of Nuclear Power Plants. Safety Guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide provides recommendations and guidance on design for the protection of nuclear power plants from the effects of external events (excluding earthquakes), i.e. events that originate either off the site or within the boundaries of the site but from sources that are not directly involved in the operational states of the nuclear power plant units. In addition, it provides recommendations on engineering related matters in order to comply with the safety objectives and requirements established in the IAEA Safety Requirements publication, Safety of Nuclear Power Plants: Design. It is also applicable to the design and safety assessment of items important to the safety of land based stationary nuclear power plants with water cooled reactors. Contents: 1. Introduction; 2. Application of safety criteria to the design; 3. Design basis for external events; 4. Aircraft crash; 5. External fire; 6. Explosions; 7. Asphyxiant and toxic gases; 8. Corrosive and radioactive gases and liquids; 9. Electromagnetic interference; 10. Floods; 11. Extreme winds; 12. Extreme meteorological conditions; 13. Biological phenomena; 14. Volcanism; 15. Collisions of floating bodies with water intakes and UHS components; Annex I: Aircraft crashes; Annex II: Detonation and deflagration; Annex III: Toxicity limits.

  20. Preliminary standard review guide for Environmental Restoration/Decontamination and Decommissioning safety analyses

    International Nuclear Information System (INIS)

    Ellingson, D.R.

    1993-06-01

    The review guide is based on the shared experiences, approaches, and philosophies of the Environmental Restoration/Decontamination and Decommissioning (ER/D ampersand D) subgroup members. It is presented in the form of a review guide to maximize the benefit to both the safety analyses practitioner and reviewer. The guide focuses on those challenges that tend to be unique to ER/D ampersand D cleanup activities. Some of these experiences, approaches, and philosophies may find application or be beneficial to a broader spectrum of activities such as terminal cleanout or even new operations. Challenges unique to ER/D ampersand D activities include (1) consent agreements requiring activity startup on designated dates; (2) the increased uncertainty of specific hazards; and (3) the highly variable activities covered under the broad category of ER/D ampersand D. These unique challenges are in addition to the challenges encountered in all activities; e.g., new and changing requirements and multiple interpretations. The experiences in approaches, methods, and solutions to the challenges are documented from the practitioner and reviewer's perspective, thereby providing the viewpoints on why a direction was taken and the concerns expressed. Site cleanup consent agreements with predetermined dates for restoration activity startup add the dimension of imposed punitive actions for failure to meet the date. Approval of the safety analysis is a prerequisite to startup. Actions that increase expediency are (1) assuring activity safety; (2) documenting that assurance; and (3) acquiring the necessary approvals. These actions increase the timeliness of startup and decrease the potential for punitive action. Improvement in expediency has been achieved by using safety analysis techniques to provide input to the line management decision process rather than as a review of line management decisions. Expediency is also improved by sharing the safety input and resultant decisions with

  1. Working material. IAEA seismic safety of nuclear power plants. International workshop on lessons learned from strong earthquake

    International Nuclear Information System (INIS)

    2008-08-01

    The International Workshop on Lessons Learned from Strong Earthquake was held at Kashiwazaki civic plaza, Kashiwazaki, Niigata-prefecture, Japan, for three days in June 2008. Kashiwazaki-Kariwa NPP (KK-NPP) is located in the city of Kashiwazaki and the village of Kariwa, and owned and operated by Tokyo Electric Power Company Ltd. (TEPCO). After it experienced the Niigata-ken Chuetsu-oki earthquake in July 2007, IAEA dispatched experts' missions twice and held technical discussions with TEPCO. Through such activities, the IAEA secretariat and experts obtained up-dated information of plant integrity, geological and seismological evaluation and developments of the consultation in the regulatory framework of Japan. Some of the information has been shared with the member states through the reports on findings and lessons learned from the missions to Japan. The international workshop was held to discuss and share the information of lessons learned from strong earthquakes in member states' nuclear installations. It provided the opportunity for participants from abroad to share the information of the recent earthquake and experience in Japan and to visit KK-NPP. And for experts in Japan, the workshop provided the opportunity to share the international approach on seismic-safety-related measures and experiences. The workshop was organised by the IAEA as a part of an extra budgetary project, in cooperation with OECD/NEA, hosted by Japanese organisations including Nuclear and Industrial Safety Agency (NISA), Nuclear Safety Commission (NSC), and Japan Nuclear Energy Safety Organization (JNES). The number of the workshop participants was 70 experts from outside Japan, 27 countries and 2 international organisations, 154 Japanese experts and 81 audience and media personnel, totalling to 305 participants. The three-day workshop was open to the media including the site visit, and covered by NHK (the nation's public broadcasting corporation) and nation-wide and local television

  2. Seismic qualification of safety-related instrumentation cabinets for nuclear generating stations

    International Nuclear Information System (INIS)

    Sauve, R.G.; Bell, R.P.; Cuttler, J.M.

    1979-06-01

    The problem of seismically qualifying different electrical instruments mounted in cabinets of a standard design is discussed and the following economical approach is described in detail. An analytical model of the cabinet structure is developed and validated by comparison with the results of shake table tests on a prototype cabinet. Modal analysis is then used to calculate the input spectra for shake table tests to qualify the individual instruments that are mounted at the required elevations in the cabinet. The worst input spectrum, appropriate to qualify each instrument, is then specified to the suppliers. This approach avoids the need to test each cabinet configuration in an assembled state in order to qualify it. (auth)

  3. A Framework for Seismic Design of Items in Safety-Critical Facilities for Implementing a Risk-Informed Defense-in-Depth-Based Concept

    Directory of Open Access Journals (Sweden)

    Tatsuya Itoi

    2017-05-01

    Full Text Available Recently, especially after the 2011 off the Pacific coast of Tohoku earthquake and the Fukushima Daiichi nuclear power plant accident, the need for treating residual risks and cliff-edge effects in safety-critical facilities has been widely recognized as an extremely important issue. In this article, the sophistication of seismic designs in safety-critical facilities is discussed from the viewpoint of mitigating the consequences of accidents, such as the avoidance of cliff-edge effects. For this purpose, the implementation of a risk-informed defense-in-depth-based framework is proposed in this study. A basic framework that utilizes diversity in the dynamic characteristics of items and also provides additional seismic margin to items important for safety when needed is proposed to prevent common cause failure and to avoid cliff-edge effects as far as practicable. The proposed method is demonstrated to be effective using an example calculation.

  4. Yield and Safety Profile of Ultrasound Guided Fine Needle Aspiration Cytology (FNAC) of Lymph Nodes

    International Nuclear Information System (INIS)

    Sattar, A.; Wahab, S.; Javed, A.; Shamim, S. H.

    2016-01-01

    Objective: To determine the re-biopsy rate, positive yield and safety profile of ultrasound guided fine needle aspiration cytology (FNAC) in cervical lymph nodes in terms of its complications and repeat procedures. Study Design: An analytical study. Place and Duration of Study: Department of Vascular and Interventional Radiology, Dow University Hospital, Dow University of Health Sciences, Karachi, from June to December 2013. Methodology: Eighty neck swellings, which were found to be lymph nodes on ultrasound, underwent ultrasound guided FNAC, from outpatients. Lymph nodes which were included in the study were those that were not easily palpable, located near major blood vessels, where patient refused of direct palpation and wanted image guided FNAC, those directly sent by physician for image guided FNAC and where blind biopsy remained inconclusive. Patients who refused on explanation or did not give consent were excluded. Complications and repeat biopsy were noted. Result: This study consisted of 80 cases, of which 51 cases (63.75 percentage) were female and 29 cases (36.25 percentage) were male. Repeat biopsy was required in 1 case (1.6 percentage). There were no procedure-related complications. A total of 44 cases (55 percentage) revealed evidence suggesting or confirming the existence of tuberculosis. Rest of the others showed other benign lesions, reactive lymphadenopathy and malignancy. Conclusion: Ultrasound guided FNAC is a safe procedure with low re-biopsy rate that aids diagnosis. The predominant cause of cervical lymphadenopathy in this study was tuberculous lymphadenitis. (author)

  5. A SIL quantification approach based on an operating situation model for safety evaluation in complex guided transportation systems

    International Nuclear Information System (INIS)

    Beugin, J.; Renaux, D.; Cauffriez, L.

    2007-01-01

    Safety analysis in guided transportation systems is essential to avoid rare but potentially catastrophic accidents. This article presents a quantitative probabilistic model that integrates Safety Integrity Levels (SIL) for evaluating the safety of such systems. The standardized SIL indicator allows the safety requirements of each safety subsystem, function and/or piece of equipment to be specified, making SILs pivotal parameters in safety evaluation. However, different interpretations of SIL exist, and faced with the complexity of guided transportation systems, the current SIL allocation methods are inadequate for the task of safety assessment. To remedy these problems, the model developed in this paper seeks to verify, during the design phase of guided transportation system, whether or not the safety specifications established by the transport authorities allow the overall safety target to be attained (i.e., if the SIL allocated to the different safety functions are sufficient to ensure the required level of safety). To meet this objective, the model is based both on the operating situation concept and on Monte Carlo simulation. The former allows safety systems to be formalized and their dynamics to be analyzed in order to show the evolution of the system in time and space, and the latter make it possible to perform probabilistic calculations based on the scenario structure obtained

  6. Environmental and Source Monitoring for Purposes of Radiation Protection. Safety Guide (Spanish ed.)

    International Nuclear Information System (INIS)

    2010-01-01

    The purpose of this Safety Guide is to provide international guidance, coherent with contemporary radiation protection principles and IAEA safety requirements, on the strategy of monitoring in relation to: (a) control of radionuclide discharges under practice conditions, and (b) intervention, such as in cases of nuclear or radiological emergencies or past contamination of areas with long lived radionuclides. Three categories of monitoring are discussed: monitoring at the source of the discharge (source monitoring), monitoring in the environment (environmental monitoring) and monitoring of individual exposure in emergencies (individual monitoring). The Safety Guide also provides general guidance on assessment of the doses to critical groups of the population due to the presence of radioactive materials or radiation fields in the environment both from routine operation of nuclear and other related facilities (practice) and from nuclear or radiological emergencies and past contamination of areas with long lived radionuclides (intervention). The dose assessments are based on the results of source monitoring, environmental monitoring, individual monitoring or their combinations. This Safety Guide is primarily intended for use by national regulatory bodies and other agencies involved in national systems of radiation monitoring, as well as by operators of nuclear installations and other facilities where natural or human made radionuclides are treated and monitored. Contents: 1. Introduction; 2. Meeting regulatory requirements for monitoring in practices and interventions; 3. Responsibilities for monitoring; 4. Generic aspects of monitoring programmes; 5. Programmes for monitoring in practices and interventions; 6. Technical conditions for monitoring procedures; 7. Considerations in dose assessment; 8. Interpretation of monitoring results; 9. Quality assurance; 10. Recording of results; 11. Education and training; Glossary.

  7. Efforts toward enhancing seismic safety at Kashiwazaki Kariwa nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Yamashita, Kazuhiko

    2010-09-15

    Kashiwazaki-Kariwa Nuclear Power Station, 8212MW, was struck by M6.8 quakes in July 2007. TEPCO has steadily been conducting restoration and post-earthquake equipment integrity assessment, aiming to make it a disaster-resistant power station. 2 units among 7 resumed commercial operation by June 2010. This earthquake has provided a great deal of knowledge and information useful for nuclear safety improvement. It has also served as a valuable reference for the IAEA in developing earthquake-related guidelines. TEPCO would like to share the knowledge and information thereby contributing to improving the safety of nuclear power generation. We will introduce some of our activities.

  8. Seismic analysis and testing of nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    The following subjects are discussed in this guide: General Recommendations for seismic classification, loading combinations and allowable limits; seismic analysis methods; implications for seismic design; seismic testing and qualification; seismic instrumentation; modelling techniques; material property characterization; seismic response of soil deposits and earth structures; liquefaction and ground failure; slope stability; sloshing effects in water pools; qualification testing by means of the transport vehicle

  9. An Incremental, Measurable Approach to Increased Seismic Safety in Latin America and the Caribbean

    Science.gov (United States)

    Dickey, J. S.

    2001-05-01

    Plans for a multiyear effort to assess and mitigate seismic risks in municipalities throughout Latin America and the Caribbean are being developed by a committee of scientists, engineers and public servants from throughout the region. Prompted by AGU and GeoHazards International, with start-up funding from the AGU Council through the AGU Committee on International Participation, the effort will involve scientists, engineers, architects, urban planners, civil defense authorities, municipal authorities, public health authorities, and commerical interests. With technical guidance provided by the project, teams of volunteers will assess risks in their own municipalities and will identify and adopt measures to reduce those risks. Planned by Latin Americans for the benefit of Latin America, the process, which is intended to run for a ten year period, will be iterative and incremental. Progress will be measurable and will be reported at triennial conferences. As an international organization, well-represented in the region and unencumbered by political or commercial relationships, AGU is able to provide effective administrative support for this challenging endeavor.

  10. Development of seismic safety reevaluation procedure considering the ageing of NPP facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Kue [Jeonju Univ., Cheonju (Korea, Republic of); Kim, J. M. [Cheonnam National Univ., Gwangju (Korea, Republic of); Kim, Y. S.; Cheong, S. H.; Kim, I. S.; Lee, M. G.; Kim, D. O. [Andong National Univ., Andong (Korea, Republic of); Lee, G. H. [Mokpo National Maritime Univ., Mokpo (Korea, Republic of)

    2003-03-15

    There are three of Nuclear Power Plants subject to the USI A-46 in Korea, including Kori No 1 and No 2 and Wolsung No 1. For the sake of resolution of the issue the possibility of adopting the GIP developed by the SQUG in USA is very high. In relation to the issue, this study addresses some technical improvements of the GIP including sloshing analysis based on multiple modes, seismic retrofit of cabinet for reduction of ICRS and modification of IRS depending on damping ratio. Dominant degradation factor and its affects NPP concrete elements are reviewed : chloride induced corrosion, carbonation of concrete elements, freezing and thawing of concrete elements, chemical and biological process, crack affect on concrete degradation. Various technical reports and papers about age-related degradation are reviewed for identification of degradation properties of NPP structures and components and degradation trend in NPP structures and components. This report summarizes numerical model for concrete degradation and development procedure of numerical models for concrete degradation. This report proposes the research necessity for performance evaluation of degraded concrete structure and selection of element for further study.

  11. Recommendations for resolution of public comments on USI [Unresolved Safety Issues] A-40, ''Seismic Design Criteria''

    International Nuclear Information System (INIS)

    Philippacopoulos, A.J.

    1989-06-01

    In June 1988 the Nuclear Regulatory Commission (NRC) issued for public comment the proposed Revision 2 of the Standard Review Plan (SRP) Sections 2.5.2, 3.7.1, 3.7.2. and 3.7.3. Comments were received from six organizations. Brookhaven National Laboratory (BNL) was requested by NRC to provide expert consultation in the seismic and soil-structure interaction areas for the review and resolution of these comments. For this purpose, a panel of consultants was established to assist BNL with the review and evaluation of the public comments. This review was carried out during the period of October 1988 through January 1989. Many of the suggestions given in the public comments were found to be significant and a number of modifications to appropriate SRP sections are recommended. Other public comments were found to have no impact on the proposed Revision 2 of the SRP. Major changes are recommended to the SRP sections dealing with (a) Power Spectral Density (PSD) and ground motion requirements and (b) soil-structure interaction requirements. This report contains specific recommendations to NRC for resolution of the public comments made on the proposed Revision 2 of the SRP

  12. Regional relationships among earthquake magnitude scales. Seismic safety margins research program

    International Nuclear Information System (INIS)

    Chung, D.H.; Bernreuter, D.L.

    1980-09-01

    The seismic body-wave magnitude m b of an earthquake is strongly affected by regional variations in the Q structure, composition, and physical state within the earth. Therefore, because of differences in attenuation of P-waves between the western and eastern United States, a problem arises when comparing m b 's for the two regions. A regional m b magnitude bias exists which, depending on where the earthquake occurs and where the P-waves are recorded, can lead to magnitude errors as large as one-third unit. There is also a significant difference between m b and M L values for earthquakes in the western United States. An empirical link between the m b of an eastern U.S. earthquake and the M L of an equivalent western earthquake is given y M L = 0.57 + 0.92(m b ) East . This result is important when comparing ground motion between the two regions and for choosing a set of real western U.S. earthquake records to represent eastern earthquakes. (author)

  13. Seismic Studies

    Energy Technology Data Exchange (ETDEWEB)

    R. Quittmeyer

    2006-09-25

    This technical work plan (TWP) describes the efforts to develop and confirm seismic ground motion inputs used for preclosure design and probabilistic safety 'analyses and to assess the postclosure performance of a repository at Yucca Mountain, Nevada. As part of the effort to develop seismic inputs, the TWP covers testing and analyses that provide the technical basis for inputs to the seismic ground-motion site-response model. The TWP also addresses preparation of a seismic methodology report for submission to the U.S. Nuclear Regulatory Commission (NRC). The activities discussed in this TWP are planned for fiscal years (FY) 2006 through 2008. Some of the work enhances the technical basis for previously developed seismic inputs and reduces uncertainties and conservatism used in previous analyses and modeling. These activities support the defense of a license application. Other activities provide new results that will support development of the preclosure, safety case; these results directly support and will be included in the license application. Table 1 indicates which activities support the license application and which support licensing defense. The activities are listed in Section 1.2; the methods and approaches used to implement them are discussed in more detail in Section 2.2. Technical and performance objectives of this work scope are: (1) For annual ground motion exceedance probabilities appropriate for preclosure design analyses, provide site-specific seismic design acceleration response spectra for a range of damping values; strain-compatible soil properties; peak motions, strains, and curvatures as a function of depth; and time histories (acceleration, velocity, and displacement). Provide seismic design inputs for the waste emplacement level and for surface sites. Results should be consistent with the probabilistic seismic hazard analysis (PSHA) for Yucca Mountain and reflect, as appropriate, available knowledge on the limits to extreme ground

  14. Seismic Studies

    International Nuclear Information System (INIS)

    R. Quittmeyer

    2006-01-01

    This technical work plan (TWP) describes the efforts to develop and confirm seismic ground motion inputs used for preclosure design and probabilistic safety 'analyses and to assess the postclosure performance of a repository at Yucca Mountain, Nevada. As part of the effort to develop seismic inputs, the TWP covers testing and analyses that provide the technical basis for inputs to the seismic ground-motion site-response model. The TWP also addresses preparation of a seismic methodology report for submission to the U.S. Nuclear Regulatory Commission (NRC). The activities discussed in this TWP are planned for fiscal years (FY) 2006 through 2008. Some of the work enhances the technical basis for previously developed seismic inputs and reduces uncertainties and conservatism used in previous analyses and modeling. These activities support the defense of a license application. Other activities provide new results that will support development of the preclosure, safety case; these results directly support and will be included in the license application. Table 1 indicates which activities support the license application and which support licensing defense. The activities are listed in Section 1.2; the methods and approaches used to implement them are discussed in more detail in Section 2.2. Technical and performance objectives of this work scope are: (1) For annual ground motion exceedance probabilities appropriate for preclosure design analyses, provide site-specific seismic design acceleration response spectra for a range of damping values; strain-compatible soil properties; peak motions, strains, and curvatures as a function of depth; and time histories (acceleration, velocity, and displacement). Provide seismic design inputs for the waste emplacement level and for surface sites. Results should be consistent with the probabilistic seismic hazard analysis (PSHA) for Yucca Mountain and reflect, as appropriate, available knowledge on the limits to extreme ground motion at

  15. Lessons Learned from Process Safety Management: A Practical Guide to Defence in Depth

    Energy Technology Data Exchange (ETDEWEB)

    Langerman, N., E-mail: neal@chemical-safety.com [Advanced Chemical Safety, Inc., San Diego (United States)

    2014-10-15

    Full text: Beginning with the experiences of Alfred Nobel, the chemical enterprise has learned from failures and implemented layers of protection to prevent unwanted incidents. Nobel developed dynamite as a more stable alternative to nitroglycerin, a process we would today call “inherently safer technology”. In recent years, the USA has issued regulations requiring formal “risk management plans” to identify and mitigate production risks. The USA set up the “Chemical Safety and Hazard Investigation Board” as an independent investigator of serious chemical enterprise incidents with a mission to issue recommendations aimed at preventing repeated incidents based on lessons learned. Following a particularly violent explosion in Texas in 1989, the US Occupational Safety and Health Administration issued the “Process Safety Management” (PSM) rule. PSM is a singular guide to defence in depth for preventing large-scale production incidents. The formalism is equally applicable to the chemical enterprise and the nuclear installation enterprise. This presentation will discuss the key elements of PSM and offer suggestions on using PSM as a guide to developing multiple layers of protection. The methods of PSM are applicable to Nuclear Generating Stations, research reactors, fuel reprocessing plants and fissile material storage and handling. Examples from both the chemical and nuclear enterprises will be used to illustrate key points. (author)

  16. Safety and Efficacy of Ultrasound-Guided Fiducial Marker Implantation for CyberKnife Radiation Therapy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Hyun; Hong, Seong; Sook; Kim, Jung Hoon; Park, Hyun Jeong; Chang, Yun Woo; Chang, A Ram [Soonchunhyang University Seoul Hospital, Seoul (Korea, Republic of); Kwon, Seok Beom [Hallym University College of Medicine, Chuncheon (Korea, Republic of)

    2012-06-15

    To evaluate the safety and technical success rate of an ultrasound-guided fiducial marker implantation in preparation for CyberKnife radiation therapy. We retrospectively reviewed 270 percutaneous ultrasound-guided fiducial marker implantations in 77 patients, which were performed from June 2008 through March 2011. Of 270 implantations, 104 were implanted in metastatic lymph nodes, 96 were in the liver, 39 were in the pancreas, and 31 were in the prostate. During and after the implantation, major and minor procedure-related complications were documented. We defined technical success as the implantation enabling adequate treatment planning and CT simulation. The major and minor complication rates were 1% and 21%, respectively. One patient who had an implantation in the liver suffered severe abdominal pain, biloma, and pleural effusion, which were considered as major complication. Abdominal pain was the most common complication in 11 patients (14%). Among nine patients who had markers inserted in the prostate, one had transient hematuria for less than 24 hours, and the other experienced transient voiding difficulty. Of the 270 implantations, 261 were successful (97%). The reasons for unsuccessful implantations included migration of fiducial markers (five implantations, 2%) and failure to discriminate the fiducial markers (three implantations, 1%). Among the unsuccessful implantation cases, six patients required additional procedures (8%). The symptomatic complications following ultrasound-guided percutaneous implantation of fiducial markers are relatively low. However, careful consideration of the relatively higher rate of migration and discrimination failure is needed when performing ultrasound-guided percutaneous implantations of fiducial markers.

  17. Seismic safety reexaminations to NPPs in Taiwan. Lessons learned from 20061226 Taiwan Hengchun and 20070716 Japan Niigata-Chuetsu oki earthquakes

    International Nuclear Information System (INIS)

    Chow Ting; Wu Yuanchieh; Gau Yunchau

    2008-01-01

    On December 26 2006, a strong earthquake with a local magnitude M L of 7.0 hit the most southern part of Taiwan, Hengchun village, where the Maanshan Nuclear Power Station is located. This is a historic high earthquake ever been experienced to Taiwan's existing nuclear power units, and it raised high public concerns about the seismic safety of the nuclear power plants operation. More recently on July 16 2007, in Japan, where the earthquake focal mechanisms are very similar to those in Taiwan, all 7 nuclear power units in Kashiwazaki-Kariwa site were struck by a more devastating earthquake and as the result, the design earthquakes for all the nuclear units have been exceeded. Therefore, the assurance of good seismic design and the appropriateness of associated post-earthquake actions to the nuclear power units in Taiwan become very urgent topics. Based on the experiences learned from the above mentioned two earthquakes, this paper will focus on the seismic safety reexamination of Taiwan's existing nuclear power plants of the following aspects: (1) current US orientated seismic designs/regulations from earthquake probabilistic risk point of view, (2) earthquake shut-down criterion, especially the CAV parameter and its threshold value, and (3) current post earthquake actions. (author)

  18. Resolution no. 15/2012 Safety Guide for the practice of nuclear meters

    International Nuclear Information System (INIS)

    2012-01-01

    1. This guide is Intended to complement the requirements for practice Nuclear meters out in September: • Joint Resolution CITMA-MINSAP Regulation Basic Standards Radiation safety of November 30, 2001, hereinafter NBS. • CITMA Resolution 121/2000, Regulations for the Safe Transport Radioactive Materials; hereinafter transport regulations. • Resolution 35/2003 of CITMA Regulation for the safe management of Radioactive waste of March 7, 2003, hereinafter Regulation waste. • Joint Resolution CITMA-MINSAP Regulations for the Selection, Training Authorization and Associated Personnel performing Employment Practices of Ionizing Radiation of December 19, 2003, hereinafter Staff Rules. 2. The requirements of this guide are applicable to entities and performing practice-related activities Nuclear Meters throughout the national territory.

  19. Seismic instrumentation

    International Nuclear Information System (INIS)

    1984-06-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The aim of this RFS is to define the type, location and operating conditions for seismic instrumentation needed to determine promptly the seismic response of nuclear power plants features important to safety to permit comparison of such response with that used as the design basis

  20. Ageing Management for Nuclear Power Plants. Safety Guide (Russian Edition); Upravlenie stareniem atomnykh ehlektrostantsij. Rukovodstvo po bezopasnosti

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-01-15

    The median age of nuclear power plants connected to the grid worldwide is increasing. Ageing management has become an important issue in ensuring the availability of required safety functions throughout the service life of a plant. This Safety Guide provides recommendations on meeting the requirements for safe long term operation and identifies key elements of effective ageing management for nuclear power plants.

  1. Determination of Seismic Safety Zones during the Surface Mining Operation Development in the Case of the “Buvač” Open Pit

    Directory of Open Access Journals (Sweden)

    Vladimir Malbasic

    2018-02-01

    Full Text Available Determination of the blasting safety area is a very important step in the process of drilling and blasting works, and the preparation of solid rock materials for loading. Through monitoring and analysis of the negative seismic effects to the objects and infrastructures around and at the mine area, we were able to adapt the drilling and blasting parameters and organization of drilling and blasting operation according to the mining progress so that the affected infrastructures could be protected. This paper analyses the safety distances and model safety zones of drilling and blasting for the period 2013–2018 at the open pit at “Buvač”, Omarska. This mathematical calculation procedure can be used during the whole life of the mine. By monitoring of the blasting seismic influence in first years of the mine's work, as well as by using recorded vibration velocities, mathematical dependence of the important parameters can be defined. Additionally, the level and laws of distribution and intensity of the seismic activity can be defined. On one hand, those are known quantities of the explosive and the distances between blasting location and endangered objects. On the other hand, those are coefficients of the manner of blasting and the environment where blasting is done, K, as well as the coefficient of the weakening of seismic waves as they spread, n. With the usage of the allowed vibration velocities, based on certain safety criteria and mathematical formulas of laws of spreading and intensity of seismic influence for a concrete case, it is possible to calculate explosive quantities and distances, with numerically-defined values of parameter K and n. Minimum distances are calculated based on defined or projected explosive quantities. Additionally, we calculate the maximum allowed explosive quantities based on known distances which can be used based on projected drilling-blasting parameters. For the purpose of the planning of drilling and blasting

  2. External human induced events in site evaluation for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The purpose of the present Safety Guide is to provide recommendations and guidance for the examination of the region considered for site evaluation for a plant in order to identity hazardous phenomena associated with human induced events initiated by sources external to the plant. In some cases it also presents preliminary guidance for deriving values of relevant parameters for the design basis. This Safety Guide is also applicable for periodic site evaluation and site evaluation following a major human induced event, and for the design and operation of the site's environmental monitoring system. Site evaluation includes site characterization. Consideration of external events that could lead to a degradation of the safety features of the plant and cause a release of radioactive material from the plant and/or affect the dispersion of such material in the environment. And consideration of population issues and access issues significant to safety (such as the feasibility of evacuation, the population distribution and the location of resources). The process of site evaluation continues throughout the lifetime of the facility, from siting to design, construction, operation and decommissioning. The external human induced events considered in this Safety Guide are all of accidental origin. Considerations relating to the physical protection of the plant against wilful actions by third parties are outside its scope. However, the methods described herein may also have some application for the purposes of such physical protection. The present Safety Guide may also be used for events that may originate within the boundaries of the site, but from sources which are not directly involved in the operational states of the nuclear power plant units, such as fuel depots or areas for the storage of hazardous materials for the construction of other facilities at the same site. Special consideration should be given to the hazardous material handled during the construction, operation and

  3. Seismic design and qualification for nuclear power plants

    International Nuclear Information System (INIS)

    1992-01-01

    This safety guide, which supplements the IAEA Code on the Safety of Nuclear Power Plants (NPP); Design (IAEA Safety Series No.50-C-D (Rev.1)), forms part of the Agency's programme, referred to as the NUSS programme, for establishing Codes and Guides relating to land based stationary thermal neutron power plants. The present Guide was originally issued in 1979 as Safety Guide 50-SG-S2 within the series of NUSS guides for the siting of NPP, extending seismic considerations from Safety Guide 50-SG-S1 into the design and verification field. During the revision phase in 1988-1990, this emphasis on design aspects was confirmed and consequently the Guides have been reclassified as a design Guide with the corresponding identification number 50-SG-D15. The general character of the Guide has not been changed an it still relates strongly to 50-SG-S1, which gives guidance on how to determine design basis ground motion for a NPP at a given site

  4. LISA. A code for safety assessment in nuclear waste disposals program description and user guide

    International Nuclear Information System (INIS)

    Saltelli, A.; Bertozzi, G.; Stanners, D.A.

    1984-01-01

    The code LISA (Long term Isolation Safety Assessment), developed at the Joint Research Centre, Ispra is a useful tool in the analysis of the hazard due to the disposal of nuclear waste in geological formations. The risk linked to preestablished release scenarios is assessed by the code in terms of dose rate to a maximum exposed individual. The various submodels in the code simulate the system of barriers -both natural and man made- which are interposed between the contaminants and man. After a description of the code features a guide for the user is supplied and then a test case is presented

  5. Planning and preparing for emergency response to transport accidents involving radioactive material. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The objective of this Safety Guide is to provide guidance to the public authorities and others (including consignors, carriers and emergency response authorities) who are responsible for developing and establishing emergency arrangements for dealing effectively and safely with transport accidents involving radioactive material. It may assist those concerned with establishing the capability to respond to such transport emergencies. It provides guidance for those Member States whose involvement with radioactive material is just beginning. It also provides guidance for those Member States that have already developed their radioactive material industries and the attendant emergency plans but that may need to review and improve these plans

  6. Cooperative development of nuclear safety regulations, guides and standards based on NUSS

    International Nuclear Information System (INIS)

    Pachner, J.; Boyd, F.C.; Yaremy, E.M.

    1984-10-01

    In 1983, the Atomic Energy Control Board and Atomic Energy of Canada Limited conducted a study of a possible joint program involving Canada, a nuclear power plant importing Member State and the IAEA for the development of the national nuclear safety regulations and guides based on NUSS documents. During the study, a work plan with manpower estimates for the development of design was prepared as an investigatory exercise. The work plan suggests that a successful NUSS implementation in developing Member States will require availability of significant resources at the start of the program. The study showed that such a joint program could provide an effective mechanism for transfer of nuclear safety know-how to the developing Member States through NUSS implementation

  7. Role of field testing and shaking table test on full scale structure for NPP seismic-safety, and its relation to computational mechanics

    International Nuclear Information System (INIS)

    Shibata, Heki

    1988-01-01

    Field testing on the dynamic behavior of actual structures is significant for the seismic safety of nuclear power plants. For their mechanical components and piping systems, the full scale testings are also important as well as the in-situ test of buildings. In general, it is often observed that they don't behave as that of analytical model for the design. This article tries to discuss how such discrepancy is occurring, and how to overcome it. (author)

  8. Determination of Seismic Safety Zones during the Surface Mining Operation Development in the Case of the “Buvač” Open Pit

    OpenAIRE

    Vladimir Malbasic; Lazar Stojanovic

    2018-01-01

    Determination of the blasting safety area is a very important step in the process of drilling and blasting works, and the preparation of solid rock materials for loading. Through monitoring and analysis of the negative seismic effects to the objects and infrastructures around and at the mine area, we were able to adapt the drilling and blasting parameters and organization of drilling and blasting operation according to the mining progress so that the affected infrastructures could be protecte...

  9. Role of field testing and shaking table test on full scale structure for NPP seismic-safety, and its relation to computational mechanics

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Heki [Institute of Industrial Science, University of Tokyo (Japan)

    1988-07-01

    Field testing on the dynamic behavior of actual structures is significant for the seismic safety of nuclear power plants. For their mechanical components and piping systems, the full scale testings are also important as well as the in-situ test of buildings. In general, it is often observed that they don't behave as that of analytical model for the design. This article tries to discuss how such discrepancy is occurring, and how to overcome it. (author)

  10. Guiding principles for the implementation of non-animal safety assessment approaches for cosmetics: skin sensitisation.

    Science.gov (United States)

    Goebel, Carsten; Aeby, Pierre; Ade, Nadège; Alépée, Nathalie; Aptula, Aynur; Araki, Daisuke; Dufour, Eric; Gilmour, Nicola; Hibatallah, Jalila; Keller, Detlef; Kern, Petra; Kirst, Annette; Marrec-Fairley, Monique; Maxwell, Gavin; Rowland, Joanna; Safford, Bob; Schellauf, Florian; Schepky, Andreas; Seaman, Chris; Teichert, Thomas; Tessier, Nicolas; Teissier, Silvia; Weltzien, Hans Ulrich; Winkler, Petra; Scheel, Julia

    2012-06-01

    Characterisation of skin sensitisation potential is a key endpoint for the safety assessment of cosmetic ingredients especially when significant dermal exposure to an ingredient is expected. At present the mouse local lymph node assay (LLNA) remains the 'gold standard' test method for this purpose however non-animal test methods are under development that aim to replace the need for new animal test data. COLIPA (the European Cosmetics Association) funds an extensive programme of skin sensitisation research, method development and method evaluation and helped coordinate the early evaluation of the three test methods currently undergoing pre-validation. In May 2010, a COLIPA scientific meeting was held to analyse to what extent skin sensitisation safety assessments for cosmetic ingredients can be made in the absence of animal data. In order to propose guiding principles for the application and further development of non-animal safety assessment strategies it was evaluated how and when non-animal test methods, predictions based on physico-chemical properties (including in silico tools), threshold concepts and weight-of-evidence based hazard characterisation could be used to enable safety decisions. Generation and assessment of potency information from alternative tools which at present is predominantly derived from the LLNA is considered the future key research area. Copyright © 2012 Elsevier Inc. All rights reserved.

  11. Twenty-third water reactor safety information meeting. Volume 3, structural and seismic engineering, primary systems integrity, equipment operability and aging, ECCS strainer blockage research and regulatory issues

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 3, presents topics in Structural & Seismic Engineering, Primary Systems Integrity, Equipment Operability and Aging, and ECCS Strainer Blockage Research & Regulatory Issues. Individual papers have been cataloged separately.

  12. Twenty-third water reactor safety information meeting. Volume 3, structural and seismic engineering, primary systems integrity, equipment operability and aging, ECCS strainer blockage research and regulatory issues

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 3, presents topics in Structural ampersand Seismic Engineering, Primary Systems Integrity, Equipment Operability and Aging, and ECCS Strainer Blockage Research ampersand Regulatory Issues. Individual papers have been cataloged separately

  13. Research items regarding seismic residual risk evaluation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    After learning the Fukushima Dai-ichi NPP severe accidents in 2011, the government investigation committee proposed the effective use of probabilistic safety assessment (PSA), and now it is required to establish new safety rules reflecting the results of probabilistic risk assessment (PRA) and proposed severe accident measures. Since the Seismic Design Guide has been revised on September 19, 2006, JNES has been discussing seismic PRA (Levels 1-3) methods to review licensees' residual risk assessment while preparing seismic PRA models. Meanwhile, new safety standards for light water reactors are to be issued and enforced on July 2013, which require the residual risk of tsunami, in addition to earthquakes, should be lowered as much as possible. The Fukushima accidents raised the problems related to risk assessment, e.g. approaches based on multi-hazard (earthquake and tsunami), multi-unit, multi-site, and equipment's common cause failure. This fiscal year, while performing seismic and/or tsunami PRA to work on these problems, JNES picked up the equipment whose failure greatly contribute to core damage, surveyed accident management measures on those equipment as well as effectiveness to reduce core damage probability. (author)

  14. The management system for the disposal of radioactive waste. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    The objective of this Safety Guide is to provide recommendations on developing and implementing management systems for all phases of facilities for the disposal of radioactive waste and related activities. It covers the management systems for managing the different stages of waste disposal facilities, such as siting, design and construction, operation (i.e. the activities, which can extend over several decades, involving receipt of the waste product in its final packaging (if it is to be disposed of in packaged form), waste emplacement in the waste disposal facility, backfilling and sealing, and any subsequent period prior to closure), closure and the period of institutional control (i.e. either active control - monitoring, surveillance and remediation; or passive control - restricted land use). The management systems apply to various types of disposal facility for different categories of radioactive waste, such as: near surface (for low level waste), geological (for low, intermediate and/or high level waste), boreholes (for sealed sources), surface impoundment (for mining and milling waste) and landfill (for very low level waste). It also covers management systems for related processes and activities, such as extended monitoring and surveillance during the period of active institutional control in the post-closure phase, safety and performance assessments and development of the safety case for the waste disposal facility and regulatory authorization (e.g. licensing). This Safety Guide is intended to be used by organizations that are directly involved in, or that regulate, the facilities and activities described in paras 1.15 and 1.16, and by the suppliers of nuclear safety related products that are required to meet some or all of the requirements established in IAEA Safety Standards Series No. GS-R-3 'The Management System for Facilities and Activities'. It will also be useful to legislators and to members of the public and other parties interested in the nuclear

  15. Australian Radiation Protection and Nuclear Safety Act 1998. Guide to the Australian radiation protection and nuclear safety licensing framework. 1. ed.

    International Nuclear Information System (INIS)

    1999-03-01

    The purpose of this guide is to provide information to Commonwealth entities who may require a license under the Australian Radiation Protection and Nuclear Safety (ARPANS) Act 1998 to enable them to posses, have control of, use, operate or dispose of radiation sources. The guide describes to which agencies and what activities require licensing. It also addresses general administrative and legal matters such as appeal procedures, ongoing licensing requirements, monitoring and compliance. Applicants are advised to consult the Australian Radiation Protection and Nuclear Safety Act 1998 and accompanying Regulations when submitting applications

  16. Australian Radiation Protection and Nuclear Safety Act 1998. Guide to the Australian radiation protection and nuclear safety licensing framework; 1. ed

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

    The purpose of this guide is to provide information to Commonwealth entities who may require a license under the Australian Radiation Protection and Nuclear Safety (ARPANS) Act 1998 to enable them to posses, have control of, use, operate or dispose of radiation sources. The guide describes to which agencies and what activities require licensing. It also addresses general administrative and legal matters such as appeal procedures, ongoing licensing requirements, monitoring and compliance. Applicants are advised to consult the Australian Radiation Protection and Nuclear Safety Act 1998 and accompanying Regulations when submitting applications

  17. Ultrasonography-guided Transthoracic Cutting Biopsy of Pulmonary Lesion: Diagnostic Benefits and Safety

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Mei Ah; Park, Mi Hyun [Dankook University Hospital, Cheonan (Korea, Republic of); Shin, Byung Seok [Chungnam National University Hospital, Daejeon (Korea, Republic of); Ohm, Joon Young [Bucheon St. Mary' s Hospital, Bucheon (Korea, Republic of)

    2012-06-15

    To assess the safety and usefulness of ultrasonography-guided transthoracic cutting biopsy for lung lesions. Eighty-eight patients (66 men, 22 women, mean age 59 years) with lung lesions underwent an ultrasonography(USG)-guided transthoracic cutting biopsy. The final diagnosis was based on the findings of surgery and clinical and radiological follow-ups. The histopathologic results and diagnostic accuracy of cutting biopsy were determined. Also, the complication rate was statistically evaluated according to the mass size, number of biopsies, and the presence or absence of pleural effusion. Biopsy specimens were successfully obtained in all patients. 79 of 88 lesions (89.8%) were established by histopathology. The final diagnosis was malignant in 58 and benign in 28. The remaining 2 patients were lost to follow-up. Diagnostic sensitivity for malignant lesions was 89.6% (52/58) and that for benign lesions was 96.4% (27/28). Procedure-related complications occurred in 9 patients (10.2%) including pneumothorax (n = 2) and hemoptysis (n = 7). And there was no significant difference according to mass size, number of biopsies, or presence of pleural effusion. USG-guided transthoracic cutting biopsy is a useful and safe method for technically-feasible lung lesions

  18. Ultrasonography-guided Transthoracic Cutting Biopsy of Pulmonary Lesion: Diagnostic Benefits and Safety

    International Nuclear Information System (INIS)

    Yang, Mei Ah; Park, Mi Hyun; Shin, Byung Seok; Ohm, Joon Young

    2012-01-01

    To assess the safety and usefulness of ultrasonography-guided transthoracic cutting biopsy for lung lesions. Eighty-eight patients (66 men, 22 women, mean age 59 years) with lung lesions underwent an ultrasonography(USG)-guided transthoracic cutting biopsy. The final diagnosis was based on the findings of surgery and clinical and radiological follow-ups. The histopathologic results and diagnostic accuracy of cutting biopsy were determined. Also, the complication rate was statistically evaluated according to the mass size, number of biopsies, and the presence or absence of pleural effusion. Biopsy specimens were successfully obtained in all patients. 79 of 88 lesions (89.8%) were established by histopathology. The final diagnosis was malignant in 58 and benign in 28. The remaining 2 patients were lost to follow-up. Diagnostic sensitivity for malignant lesions was 89.6% (52/58) and that for benign lesions was 96.4% (27/28). Procedure-related complications occurred in 9 patients (10.2%) including pneumothorax (n = 2) and hemoptysis (n = 7). And there was no significant difference according to mass size, number of biopsies, or presence of pleural effusion. USG-guided transthoracic cutting biopsy is a useful and safe method for technically-feasible lung lesions

  19. Seismic testing

    International Nuclear Information System (INIS)

    Sollogoub, Pierre

    2001-01-01

    This lecture deals with: qualification methods for seismic testing; objectives of seismic testing; seismic testing standards including examples; main content of standard; testing means; and some important elements of seismic testing

  20. Structural Safety Analysis Based on Seismic Service Conditions for Butterfly Valves in a Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Sang-Uk Han

    2014-01-01

    Full Text Available The structural integrity of valves that are used to control cooling waters in the primary coolant loop that prevents boiling within the reactor in a nuclear power plant must be capable of withstanding earthquakes or other dangerous situations. In this study, numerical analyses using a finite element method, that is, static and dynamic analyses according to the rigid or flexible characteristics of the dynamic properties of a 200A butterfly valve, were performed according to the KEPIC MFA. An experimental vibration test was also carried out in order to verify the results from the modal analysis, in which a validated finite element model was obtained via a model-updating method that considers changes in the in situ experimental data. By using a validated finite element model, the equivalent static load under SSE conditions stipulated by the KEPIC MFA gave a stress of 135 MPa that occurred at the connections of the stem and body. A larger stress of 183 MPa was induced when we used a CQC method with a design response spectrum that uses 2% damping ratio. These values were lower than the allowable strength of the materials used for manufacturing the butterfly valve, and, therefore, its structural safety met the KEPIC MFA requirements.

  1. On thin ice: ground penetrating radar improves safety for seismic crews in frigid arctic darkness

    Energy Technology Data Exchange (ETDEWEB)

    Smith, M.

    2002-02-01

    The fact that workers are pushing the limits of the Canadian Arctic's ice is more than act of faith; it is the result of rapidly advancing technologies that are taking the guesswork, and therefore the risk, out of icetop exploration. The most important element to improve safety in recent years has been the increased use of ground penetrating radar (GPR) which allows the most detailed images yet of ice thickness. It is an absolutely invaluable tool for allowing vehicles to drive along the ice roads up the rivers and offshore, with significantly reduced risk for the people involved. GPR is an essential part of the equipment usually tied into global positioning system (GPS) and and geographic information system (GIS). The collected GPS and GPR data are loaded into the workstation and merged to produce a GIS map where the colored map of ice thickness is overlaid over satellite image or aerial photographs. Ground penetrating radar was first used in Austria in 1929 to measure glacial ice thickness. It fell into disuse during the 1950s but the technology advanced rapidly in subsequent years; it was used as part of Apollo 17's lunar sounder experiment in 1972. It is particularly useful in northern Arctic regions to determine near-surface thickness. With pipeline developments in the active planning stages, measuring the thickness of ice is more vital than ever; investors will not commit to multi-billion dollar projects before the resource base is fully delineated.

  2. Ultrasound-guided lumbar puncture in pediatric patients: technical success and safety.

    Science.gov (United States)

    Pierce, David B; Shivaram, Giri; Koo, Kevin S H; Shaw, Dennis W W; Meyer, Kirby F; Monroe, Eric J

    2018-06-01

    Disadvantages of fluoroscopically guided lumbar puncture include delivery of ionizing radiation and limited resolution of incompletely ossified posterior elements. Ultrasound (US) allows visualization of critical soft tissues and the cerebrospinal fluid (CSF) space without ionizing radiation. To determine the technical success and safety of US-guided lumbar puncture in pediatric patients. A retrospective review identified all patients referred to interventional radiology for lumbar puncture between June 2010 and June 2017. Patients who underwent lumbar puncture with fluoroscopic guidance alone were excluded. For the remaining procedures, technical success and procedural complications were assessed. Two hundred and one image-guided lumbar punctures in 161 patients were included. Eighty patients (43%) had previously failed landmark-based attempts. One hundred ninety-six (97.5%) patients underwent lumbar puncture. Five procedures (2.5%) were not attempted after US assessment, either due to a paucity of CSF or unsafe window for needle placement. Technical success was achieved in 187 (95.4%) of lumbar punctures attempted with US guidance. One hundred seventy-seven (90.3%) were technically successful with US alone (age range: 2 days-15 years, weight range: 1.9-53.1 kg) and an additional 10 (5.1%) were successful with US-guided thecal access and subsequent fluoroscopic confirmation. Three (1.5%) cases were unsuccessful with US guidance but were subsequently successful with fluoroscopic guidance. Of the 80 previously failed landmark-based lumbar punctures, 77 (96.3%) were successful with US guidance alone. There were no reported complications. US guidance is safe and effective for lumbar punctures and has specific advantages over fluoroscopy in pediatric patients.

  3. Diagnostic accuracy and safety of CT-guided fine needle aspiration biopsy in cavitary pulmonary lesions

    Energy Technology Data Exchange (ETDEWEB)

    Zhuang, Yi-Ping, E-mail: yipingzhuang2010@sina.com [Department of Radiology, Jiangsu Cancer Institute and Hospital, No. 42 Baiziting Road, Nanjing 210009, Jiangsu (China); Wang, Hai-Yan, E-mail: mycherishgirl@sohu.com [Department of Radiology, Jiangsu Cancer Institute and Hospital, No. 42 Baiziting Road, Nanjing 210009, Jiangsu (China); Zhang, Jin, E-mail: yari_zj@hotmail.com [Department of Radiology, Jiangsu Cancer Institute and Hospital, No. 42 Baiziting Road, Nanjing 210009, Jiangsu (China); Feng, Yong, E-mail: fengyong119@sohu.com [Department of Radiology, Jiangsu Cancer Institute and Hospital, No. 42 Baiziting Road, Nanjing 210009, Jiangsu (China); Zhang, Lei, E-mail: motozl163@163.com [Department of Radiology, Jiangsu Cancer Institute and Hospital, No. 42 Baiziting Road, Nanjing 210009, Jiangsu (China)

    2013-01-15

    Objective: CT-guided transthoracic biopsy is a well-established method in the cytologic or histologic diagnosis of pulmonary lesions. The knowledge of its diagnostic performance and complications for cavitary pulmonary lesions is limited. The purpose of this study was to determine the diagnostic accuracy and safety of CT-guided fine needle aspiration biopsy (FNAB) in cavitary pulmonary lesions. Materials and methods: 102 consecutive patients with pulmonary cavitary lesions received CT-guided FNAB with use of an 18-gauge (n = 35) or 20-gauge (n = 67) Chiba for histology diagnosis. The sensitivity, specificity, and diagnostic accuracy of FNAB were calculated as compared with the final diagnosis. Complications associated with FNAB were observed. The diagnostic accuracy and complications were compared between patients with different lesion sizes and different cavity wall thickness. Results: The overall sensitivity, specificity, and accuracy of FNAB were 96.3%, 98.0%, and 96.1%, respectively. The sensitivity, specificity, and diagnosis accuracy in different lesion size (<2 cm vs ≥2 cm), or different cavity wall thickness (<5 mm vs ≥5 mm) were not different (P > 0.05; 0.235). More nondiagnostic sample was found in wall thickness <5 mm lesions (P = 0.017). Associated complications included pneumothorax in 9 (8.8%) patients and alveolar hemorrhage in 14 patients (13.7%) and hemoptysis in 1 patient (1%). No different rate of complications was found with regard to lesion size, wall thickness, length of the needle path and needle size (P > 0.05). Conclusion: CT-guided FNAB can be effectively ad safely used for patients with pulmonary cavitary lesions.

  4. Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material. Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide provides recommendations on achieving and demonstrating compliance with IAEA Safety Standards Series No. TS-R-1, Regulations for the Safe Transport of Radioactive Material, 2005 Edition, establishing safety requirements to be applied to the national and international transport of radioactive material. Transport is deemed to comprise all operations and conditions associated with and involved in the movement of radioactive material; these include the design, fabrication and maintenance of packaging, and the preparation, consigning, handling, carriage, storage in transit and receipt at the final destination of packages. This publication supersedes IAEA Safety Series No. TS-G-1.1, 2002 Edition

  5. Packaging review guide for reviewing safety analysis reports for packagings: Revision 1

    International Nuclear Information System (INIS)

    Fisher, L.E.; Chou, C.K.; Lloyd, W.R.; Mount, M.E.; Nelson, T.A.; Schwartz, M.W.; Witte, M.C.

    1988-10-01

    The Department of Energy (DOE) has established procedures for obtaining certification of packagings used by DOE and its contractors for the transport of radioactive materials. The principal purpose of this document is to assure the quality and uniformity of PCS reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The Packaging Review Guide (PRG) also sets forth solutions and approaches determined to be acceptable in the past in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a SARP does not have to follow the solutions or approaches presented. It is also a purpose of the PRG to make information about DOE certification policy and procedures widely available to DOE field offices, DOE contractors, federal agencies, and interested members of the public. 77 refs., 16 figs., 15 tabs

  6. The management system for the safe transport of radioactive material. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    The purpose of this Safety Guide is to provide information to organizations that are developing, implementing or assessing a management system for activities relating to the transport of radioactive material. Such activities include, but are not limited to, design, fabrication, inspection and testing, maintenance, transport and disposal of radioactive material packaging. This publication is intended to assist those establishing or improving a management system to integrate safety, health, environmental, security, quality and economic elements to ensure that safety is properly taken into account in all activities of the organization. Contents: 1. Introduction; 2. Management system; 3. Management responsibility; 4. Resource management; 5. Process implementation; 6. Measurement, assessment and improvement; Appendix: Graded approach for management systems for the safe transport of radioactive materials; Annex I: Two examples of management systems; Annex II: Examples of management system standards; Annex III: Example of a documented management system (or quality assurance programme) for an infrequent consignor; Annex IV: Example of a documented management system (or quality assurance programme) description for an infrequent carrier; Annex V: Example of a procedure for control of records; Annex VI: Example of a procedure for handling packages containing radioactive materials, including receipt and dispatch; Annex VII: Example of a packaging maintenance procedure in a complex organization; Annex VIII: Example of an internal audit procedure in a small organization; Annex IX: Example of a corrective and preventive action procedure

  7. ACED devices and SECAF supports for the control of structure, pipe network and equipment behaviour at seismic movements in order to enhance the safety margin

    International Nuclear Information System (INIS)

    Serban, Viorel; Prisecaru, I.; Cretu, D.; Moldoveanu, T.

    2002-01-01

    In order to enhance the safety margin of structure, pipe networks and equipment associated to the existing NPPs, the classic consolidation solutions are very expensive and many times, impossible to be implemented. Structures, pipe networks, systems and equipment have geometries imposed by the basic construction requirements, operating and safety requirements and their modifications is not always possible. In order to enhance the strength capacity of (new or old) structures, systems and equipment mechanical devices with controlled elasticity and damping (ACED) have been designed, constructed and experimented. These devices are capable to support very large static loads over which dynamic loads (shock, vibration and seismic movements) overlap (which are damped). To increase the strength capacity of (new or existing) pipe networks and equipment connecting with pipes, SECAF supports that allow displacements from thermal expansions with low reaction force have been designed, constructed and experimented. SECAF supports are capable elastically to take permanent loads over which shocks, vibrations and seismic movements (which are damp) overlap. ACED devices and SECAF supports can be used to rehabilitate the existing NPPs with law financial costs and an increase of their strength capacity up to 100% under seismic movements, shocks and vibrations. ACED devices and SECAF supports do not require maintenance, are not affected by presence of a radiation field and their estimated service-life is similar to the NPPs

  8. Seismicity and seismic monitoring in the Asse salt mine

    International Nuclear Information System (INIS)

    Flach, D.; Gommlich, G.; Hente, B.

    1987-01-01

    Seismicity analyses are made in order to assess the safety of candidate sites for ultimate disposal of hazardous wastes. The report in hand reviews the seismicity history of the Asse salt mine and presents recent results of a measuring campaign made in the area. The monitoring network installed at the site supplies data and information on the regional seismicity, on seismic amplitudes under ground and above ground, and on microseismic activities. (DG) [de

  9. Packaging review guide for reviewing safety analysis reports for packagings: Revision 0

    International Nuclear Information System (INIS)

    Fischer, L.E.; Chou, C.K.; Lloyd, W.R.; Mount, M.E.; Nelson, T.A.; Schwartz, M.W.; Witte, M.C.

    1987-09-01

    The Department of Energy (DOE) has established procedures for obtaining certification of packagings used by DOE and its contractors for the transport of radioactive materials. These certification review policies and procedures are established to ensure that DOE packaging designs and operations meet safety criteria at least equivalent to the standards prescribed by the Nuclear Regulatory Commission (NRC) certification process for packaging. The Packaging Review Guide (PRG) is not a DOE order, but has been prepared as guidance for the Packaging Certification Staff (PCS) under the Certifying Official, Office of Security Evaluations, or designated representatives. The principal purpose of the PRG is to assure the quality and uniformity of PCS reviews, and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The PRG also sets forth solutions and approaches determined to be acceptable in the past by the PCS in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a SARP does not have to follow the solutions or approaches presented in the PRG. However, applicants should recognize that the PCS has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches. Finally, it is also a purpose of the PRG to make information about DOE certification policy and procedures widely available to DOE field offices, DOE contractors, federal agencies, and interested members of the public. 7 refs., 15 figs., 14 tabs

  10. A Study on the Development of Prototype Seismic Isolation Device for NPP

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hongpyo; Cho, Myungsug; Kim, Sunyong; Lee, Yonghee; Kang Kyunghun [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-05-15

    Korean nuclear power plants have been and still are based on seismic resistance design including all of the natural disasters. However, in regions of high seismic hazard, seismic isolation technology is needed to guarantee the seismic safety on nuclear power plants. To achieve this purpose, the research and development of seismic isolation system for the construction in high seismicity area is on-going in Korea. In this study, prototype seismic isolation devices as mentioned above are developed and tested to identify the basic shear and compressive characteristics of them. In this study, assessment performance of basic characteristics on the prototype LRB and EQS seismic isolation for nuclear power plant structures is employed to compare with design values. Based on the test results of compression and shear characteristics, it is judged that they meet the measuring efficiency range conditions which are presented in ISO 22762 and AASHOT guide specification. Therefore, prototype seismic isolation devices like LRB and EQS developed in this study can be expected to be used as reference data when designing a seismic isolation system for nuclear power plant structures in the future.

  11. MASCOT and MOP programs for probabilistic safety assessment. Pt. E. MOP (Version 3A) user guide

    International Nuclear Information System (INIS)

    Agg, P.J.; Hopper, M.J.; Sinclair, J.E.; Sumner, P.J.

    1994-04-01

    MOP is a post-processor for the probabilistic safety assessment program MASCOT, which models the consequences of the disposal of radioactive waste. This document provides a general description of the capabilities of the MOP program, together with a comprehensive guide to the MOP user command language. MOP is able to calculate and present various statistical measures of the modelled radiological consequences, in both printed and graphical form. The results of intermediate analyses can be saved from one MOP job to the next, and this allows MOP to be used as many times as desired to process the results of the same MASCOT job. MOP can work with the quantities passed to it from the MASCOT job or with new quantities, defined and calculated according to individual requirements. This is usually done by transforming the MASCOT quantities using algebraic expressions. (Author)

  12. 41 CFR 102-80.45 - What are Federal agencies' responsibilities concerning seismic safety in Federal facilities?

    Science.gov (United States)

    2010-07-01

    ... seismic risks in those buildings. Risks and Risk Reduction Strategies ... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false What are Federal... Public Contracts and Property Management Federal Property Management Regulations System (Continued...

  13. Considerations on safety against seismic excitations in the project of reactor auxiliary building and control building in nuclear power plants

    International Nuclear Information System (INIS)

    Santos, S.H.C.; Castro Monteiro, I. de

    1986-01-01

    The seismic requests to be considered in the project of main buildings of a nuclear power plant are discussed. The models for global seismic analysis of nuclear power plant structures, as well as models for global strength distribution are presented. The models for analysing reactor auxiliary building and control building, which together with the reactor building and turbine building form the main energy generation complex in a nuclear power plant, are described. (M.C.K.) [pt

  14. Proceedings of third Indo-German workshop and theme meeting on seismic safety of structures, risk assessment and disaster mitigation

    International Nuclear Information System (INIS)

    Reddy, G.R.; Parulekar, Y.M.

    2007-01-01

    This Indo-German workshop focuses and emphasises the current research and development activities in both the countries. Themes of this meeting are Earthquake Hazard and Vulnerability Assessment, Risk Assessment Techniques, Seismic Risk to Mega Cities, Testing and Evaluation of Structures and Components, Base Isolation and other Control Techniques, Seismic Strengthening of Structures, Design Practices and Specifications, Remote Sensing and GIS Applications, Structural Materials and Composites, Containment and Other Special Structures. Papers relevant to INIS are indexed separately

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

    International Nuclear Information System (INIS)

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

    2009-03-01

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

  16. Treatment of benign cold thyroid nodule: efficacy and safety of US-guided percutaneous ethanol injection

    International Nuclear Information System (INIS)

    Kim, Jeong Kon; Lee, Ho Kyu; Lee, Myung Joon; Choi, Choong Gon; Suh, Dae Chul; Ahn, Il Min

    1998-01-01

    The purpose of this study was to evaluate the efficacy and safety of US-guided percutaneous ethanol injection for the treatent of benign cold thyroid nodules. Twenty-five patients with benign cold thyroid nodules (volume of each at least 2ml proven by PCNA to be adenomatous hyperplasia, and cold nodule by thyroid scan) underwent a total of one to three percutaneous ethanol injections (PEI) at intervals of one or two months. The mean amount of ethanol used was 6.2(range, 1.5-8)ml, depending on the volume of the nodule. Follow up ultrasonography was performed one to four months after the final session. The initial volume of nodules was 11.4±4.1(range, 2.5-41.4)ml, and in all cases this fell by 56.1±22.3%(range, 10.9-92.1%);in all cases, follow-up ultrasonography showed that echogeneity was lower and its pattern was heterogeneous. There were no important longstanding complications;the most common side effect was acute pain at the injection site(n=3D9), and in one case, transient vocal cord palsy occurred. Our results show that US-guided percutaneous injection of ethanol is an effective and a safe procedure for the treatment of benign cold thyroid nodules, and is thus an alternative to surgery or hormone therapy.=20

  17. Component fragility analysis methodology for seismic risk assessment projects. Proven PSA safety document processing and assessment procedures

    International Nuclear Information System (INIS)

    Kolar, Ladislav

    2013-03-01

    The seismic risk task assessment task should be structured as follows: (i) Define all reactor unit building structures, components and equipment involved in the creation of an initiating event (IE) induced by an seismic event or contributing to the reliability of reactor unit response to an IE; (ii) construct and estimate of the fragility curves for the building and component groups sub (i); (iii) determine the HCLPF for each group of buildings, components or equipment; (iv) determine the nuclear source's seismic resistance (SME) as the minimum HCLPF from the group of equipment in the risk-dominant scenarios; (v) define the risk-limiting group of components, equipment and building structures to the SME value; (vi) based on the fragility levels, identify component groups for which a more detailed fragility analysis is needed; and (vii) recommend groups of equipment or building structures that should be taken into account with respect to the seismic risk, i.e. such groups of equipment or building structures as exhibit a low seismic resistance (HCLPF) and, at the same time, are involved to a significant extent in the reactor unit's seismic risk (are present in the dominant risk scenarios). (P.A.)

  18. Evaluation of Seismic Safety for Existing Nuclear Installations. Safety Guide (Russian Edition); Оценка сейсмической безопасности существующих ядерных установок. Руководство по безопасности

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-01-15

    The IAEA's Statute authorizes the Agency to establish safety standards to protect health and minimize danger to life and property - standards which the IAEA must use in its own operations, and which a State can apply by means of its regulatory provisions for nuclear and radiation safety. A comprehensive body of safety standards under regular review, together with the IAEA's assistance in their application, has become a key element in a global safety regime. In the mid-1990s, a major overhaul of the IAEA's safety standards programme was initiated, with a revised oversight committee structure and a systematic approach to updating the entire corpus of standards. The new standards that have resulted are of a high calibre and reflect best practices in Member States. With the assistance of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its safety standards. Safety standards are only effective, however, if they are properly applied in practice. The IAEA's safety services - which range in scope from engineering safety, operational safety, and radiation, transport and waste safety to regulatory matters and safety culture in organizations - assist Member States in applying the standards and appraise their effectiveness. These safety services enable valuable insights to be shared and I continue to urge all Member States to make use of them. Regulating nuclear and radiation safety is a national responsibility, and many Member States have decided to adopt the IAEA's safety standards for use in their national regulations. For the contracting parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions. The standards are also applied by designers, manufacturers and operators around the world to enhance nuclear and radiation safety in power generation, medicine, industry, agriculture, research and education

  19. Percutaneous Image-guided Radiofrequency Ablation of Tumors in Inoperable Patients - Immediate Complications and Overall Safety.

    Science.gov (United States)

    Sahay, Anubha; Sahay, Nishant; Kapoor, Ashok; Kapoor, Jyoti; Chatterjee, Abhishek

    2016-01-01

    Percutaneous destruction of cancer cells using a radiofrequency energy source has become an accepted part of the modern armamentarium for managing malignancies. Radiofrequency ablation (RFA) is a relatively novel procedure for treating recurrent and metastatic tumors. It is used for debulking tumors and as adjuvant therapy for palliative care apart from its role as a pain management tool. Its use in the third world countries is limited by various factors such as cost and expertise. In the remotest parts of India, where economic development has been slow, abject poverty with poor health care facilities advanced malignancies present a challenge to health care providers. We undertook this study to assess the safety of the percutaneous RFA tumor ablation as a therapeutic or palliative measure in patients where surgery was not possible. We observed that RFA may be an effective, alternative therapeutic modality for some inoperable tumors where other therapeutic modalities cannot be considered. Palliative and therapeutic image-guided RFAs of tumors may be the only treatment option in patients who are inoperable for a variety of reasons. To assess the safety and complications of RFA in such a patient population is important before embarking upon any interventions given their physically, mentally, and socially compromised status in a country such as India. To assess the safety of percutaneous image-guided radiofrequency tumor ablation and to note the various immediate and early complications of the intervention. This was a prospective, observational study conducted in Tata Main Hospital, Jamshedpur, Jharkhand, India. After approval by the Hospital Approval Committee all patients who consented for percutaneous RFA of their tumor admitted in the hospital were included after taking fully informed consent from patient/close relative keeping the following criteria in view. Patients who were likely to derive a direct benefit in the survival or as a palliative measure for relief

  20. Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (2012 Ed.). Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-06-15

    This Safety Guide provides recommendations and guidance on achieving and demonstrating compliance with IAEA Safety Standards Series No. SSR-6, Regulations for the Safe Transport of Radioactive Material (2012 Edition), which establishes the requirements to be applied to the national and international transport of radioactive material. Transport is deemed to comprise all operations and conditions associated with and involved in the movement of radioactive material, including the design, fabrication and maintenance of packaging, and the preparation, consigning, handling, carriage, storage in transit and receipt at the final destination of packages. This publication supersedes IAEA Safety Standards Series No. TS-G-1.1 Rev. 1, which was issued in 2008.

  1. Seismic methodology in determining basis earthquake for nuclear installation

    International Nuclear Information System (INIS)

    Ameli Zamani, Sh.

    2008-01-01

    Design basis earthquake ground motions for nuclear installations should be determined to assure the design purpose of reactor safety: that reactors should be built and operated to pose no undue risk to public health and safety from earthquake and other hazards. Regarding the influence of seismic hazard to a site, large numbers of earthquake ground motions can be predicted considering possible variability among the source, path, and site parameters. However, seismic safety design using all predicted ground motions is practically impossible. In the determination of design basis earthquake ground motions it is therefore important to represent the influences of the large numbers of earthquake ground motions derived from the seismic ground motion prediction methods for the surrounding seismic sources. Viewing the relations between current design basis earthquake ground motion determination and modem earthquake ground motion estimation, a development of risk-informed design basis earthquake ground motion methodology is discussed for insight into the on going modernization of the Examination Guide for Seismic Design on NPP

  2. Radiation Protection and Radioactive Waste Management in the Operation of Nuclear Power Plants. Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    The purpose of this Safety Guide is to provide recommendations to the regulatory body, focused on the operational aspects of radiation protection and radioactive waste management in nuclear power plants, and on how to ensure the fulfilment of the requirements established in the relevant Safety Requirements publications. It will also be useful for senior managers in licensee or contractor organizations who are responsible for establishing and managing programmes for radiation protection and for the management of radioactive waste. This Safety Guide gives general recommendations for the development of radiation protection programmes at nuclear power plants. The issues are then elaborated by defining the main elements of a radiation protection programme. Particular attention is paid to area classification, workplace monitoring and supervision, application of the principle of optimization of protection (also termed the 'as low as reasonably achievable' (ALARA) principle), and facilities and equipment. This Safety Guide covers all the safety related aspects of a programme for the management of radioactive waste at a nuclear power plant. Emphasis is placed on the minimization of waste in terms of both activity and volume. The various steps in predisposal waste management are covered, namely processing (pretreatment, treatment and conditioning), storage and transport. Releases of effluents, the application of authorized limits and reference levels are discussed, together with the main elements of an environmental monitoring programme

  3. Management of waste from the use of radioactive material in medicine, industry, agriculture, research and education safety guide

    CERN Document Server

    2005-01-01

    This Safety Guide provides recommendations and guidance on the > fulfilment of the safety requirements established in Safety Standards > Series No. WS-R-2, Predisposal Management of Radioactive Waste, > Including Decommissioning. It covers the roles and responsibilities of > different bodies involved in the predisposal management of radioactive > waste and in the handling and processing of radioactive material. It > is intended for organizations generating and handling radioactive > waste or handling such waste on a centralized basis for and the > regulatory body responsible for regulating such activities.  > Contents: 1. Introduction; 2. Protection of human health and the > environment; 3. Roles and responsibilities; 4. General safety > considerations; 5. Predisposal management of radioactive waste; 6. > Acceptance of radioactive waste in disposal facilities; 7. Record > keeping and reporting; 8. Management systems; Appendix I: Fault > schedule for safety assessment and environmental impact assessment; > Ap...

  4. Seismic isolation - efficient procedure for seismic response assessement

    International Nuclear Information System (INIS)

    Zamfir, M. A.; Androne, M.

    2016-01-01

    The aim of this analysis is to reduce the dynamic response of a structure. The seismic isolation solution must take into consideration the specific site ground motion. In this paper will be presented results obtained by applying the seismic isolation method. Based on the obtained results, important conclusions can be outlined: the seismic isolation device has the ability to reduce seismic acceleration of the seismic isolated structure to values that no longer present a danger to people and environment; the seismic isolation solution is limiting devices deformations to safety values for ensuring structural integrity and stability of the entire system; the effective seismic energy dissipation and with no side effects both for the seismic isolated building and for the devices used, and the return to the initial position before earthquake occurence are obtained with acceptable permanent displacement. (authors)

  5. Approach to seismic hazard analysis for dam safety in the Sierra Nevada and Modoc Plateau of California

    International Nuclear Information System (INIS)

    Savage, W.U.; McLaren, M.K.; Edwards, W.D.; Page, W.D.

    1991-01-01

    Pacific Gas and Electric Company's hydroelectric generating system involves about 150 dams located in the Sierra Nevada and Modoc Plateau region of central and northern California. The utility's strategy for earthquake hazard assessment is described. The approach includes the following strategies: integrating regional tectonics, seismic geology, historical seismicity, microseismicity, and crustal structure to form a comprehensive regional understanding of the neotectonic setting; performing local studies to acquire data as needed to reduce uncertainties in geologic and seismic parameters of fault characteristics near specific dam sites; applying and extending recently developed geologic, seismologic, and earthquake engineering technologies to the current regional and site-specific information to evaluate fault characteristics, to estimate maximum earthquakes, and to characterize ground motion; and encouraging multiple independent reviews of earthquake hazard studies by conducting peer reviews, making field sites available to regulating agencies, and publishing results, methods and data in open literature. 46 refs., 8 tabs

  6. Efficacy and Safety of Procalcitonin-Guided Antibiotic Therapy in Lower Respiratory Tract Infections

    Directory of Open Access Journals (Sweden)

    Werner C. Albrich

    2013-01-01

    Full Text Available Background: In 14 randomized controlled studies to date, a procalcitonin (PCT-based algorithm has been proven to markedly reduce the use of antibiotics along with an unimpaired high safety and low complication rates in patients with lower respiratory tract infections (LRTIs. However, compliance with the algorithm and safety out of controlled study conditions has not yet been sufficiently investigated. Methods: We performed a prospective international multicenter observational post-study surveillance of consecutive adults with community-acquired LRTI in 14 centers (Switzerland (n = 10, France (n = 3 and the United States (n = 1. Results: Between September 2009 and November 2010, 1,759 patients were enrolled (median age 71; female sex 44.4%. 1,520 (86.4% patients had a final diagnosis of LRTI (community-acquired pneumonia (CAP, 53.7%; acute exacerbation of chronic obstructive pulmonary disease (AECOPD, 17.1%; and acute bronchitis, 14.4%. Compliance with the PCT-guided therapy (overall 68.2% was highest in patients with bronchitis (81.0% vs. AECOPD, 70.1%; CAP, 63.7%; p < 0.001, outpatients (86.1% vs. inpatients, 65.9%; p < 0.001 and algorithm-experienced centers (82.5% vs. algorithm-naive, 60.1%; p < 0.001 and showed significant geographical differences. The initial decision about the antibiotic therapy was based on PCT value in 72.4%. In another 8.6% of patients, antibiotics were administered despite low PCT values but according to predefined criteria. Thus, the algorithm was followed in 81.0% of patients. In a multivariable Cox hazard ratio model, longer antibiotic therapy duration was associated with algorithm-non-compliance, country, hospitalization, CAP vs. bronchitis, renal failure and algorithm-naïvety of the study center. In a multivariable logistic regression complications (death, empyema, ICU treatment, mechanical ventilation, relapse, and antibiotic-associated side effects were significantly associated with increasing CURB65-Score, CAP

  7. Seismic examination for assessment of safety of location of atomic energy objects (by the example of the WWR-K reactor, Ala-Tau village)

    International Nuclear Information System (INIS)

    Belyashova, N.N.

    2001-01-01

    In the Republic of Kazakhstan there are 3 research reactors (the fourth one is temporarily stopped). One of the reactors in 1998 (WWR-K, situated in the Ala Tau village, nearby Almaty city) was conserved because of a number of reasons. Including the reason of the earth crust geological structure insufficient study for the ensuring the seismic safety of the reactor site location. In 1994-1996 a number of geological-geophysical studies was carried out by Kazakhstan specialists confirming the the geological-geophysical conditions in the reactor site location in view of its safety. These condition are meeting to IAEA requirements and up-to-date standards acting in Kazakhstan

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  9. Use of a Graded Approach in the Application of the Safety Requirements for Research Reactors. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    The IAEA's Statute authorizes the Agency to 'establish or adopt? standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

  10. Radiation protection and safety guide no. GRPB-G-1: qualification and certification of radiation protection personnel

    International Nuclear Information System (INIS)

    Schandorf, C.; Darko, O.; Yeboah, J.; Osei, E.K.; Asiamah, S.D.

    1995-01-01

    A number of accidents with radiation sources are invariably due to human factors. The achievement and maintenance of proficiency in protection and safety in working with radiation devices is a necessary prerequisite. This guide specifies the national scheme and minimum requirements for qualification and certification of radiation protection personnel. The objective is to ensure adequate level of skilled personnel by continuous upgrading of knowledge and skill of personnel. The following sectors are covered by this guide: medicine, industry, research and training, nuclear facility operations, miscellaneous activities

  11. Inelastic seismic behavior of post-installed anchors for nuclear safety related structures: Generation of experimental database

    Energy Technology Data Exchange (ETDEWEB)

    Mahadik, Vinay, E-mail: vinay.mahadik@iwb.uni-stuttgart.de; Sharma, Akanshu; Hofmann, Jan

    2016-02-15

    Highlights: • Experiments for evaluating seismic behavior of anchors were performed. • Two undercut anchor products in use in nuclear facilities were considered. • Monotonic tension, shear and cycling tension tests at different crack widths. • Crack cycling tests at constant, in-phase and out-of phase tension loads. • Characteristics for the two anchors as a function of crack width were identified. - Abstract: Post installed (PI) anchors are often employed for connections between concrete structure and components or systems in nuclear power plants (NPP) and related facilities. Standardized practices for nuclear related structures demand stringent criteria, which an anchor has to satisfy in order to qualify for use in NPP related structures. In NPP and related facilities, the structure–component interaction in the event of an earthquake depends on the inelastic behavior of the concrete structure, the component system and also the anchorage system that connects them. For analysis, anchorages are usually assumed to be rigid. Under seismic actions, however, it is known that anchors may undergo significant plastic displacement and strength degradation. Analysis of structure–component interaction under seismic loads calls for numerical models simulating inelastic behavior of anchorage systems. A testing program covering different seismic loading scenarios in a reasonably conservative manner is required to establish a basis for generating numerical models of anchorage systems. Currently there is a general lack of modeling techniques to consider the inelastic behavior of anchorages in structure–component interaction under seismic loads. In this work, in view of establishing a basis for development of numerical models simulating the inelastic behavior of anchors, seismic tests on two different undercut anchors qualified for their use in NPP related structures were carried out. The test program was primarily based on the DIBt-KKW-Leitfaden (2010) guidelines

  12. Design and commissioning of the Seismicity Network of Darkhovein Nuclear Power Plant (IR360)

    International Nuclear Information System (INIS)

    Aram, M. R.

    2012-01-01

    The study of micro seismicity and monitoring the micro seismic for the purpose of surveying the existing faults treatments and recognition of blind faults and other active tectonic structures in various phases of constructing the important structures, specially nuclear power plants, is unavoidable. According to IAEA safety guides and US-NRC regulatory guides, suitable instrumentation must be provided so that the seismic response of nuclear power plant features importantly from the safety point of view. According to R.G. 1.165 seismic monitoring by a network of seismic stations in the site area should be established as soon as possible after the site selection. Also, it is necessary to shutdown the nuclear power plant if vibratory ground motion exceeds the operating basis earthquake. The current research demonstrates the field works and studies for locating the local seismograph network in Darkhovein nuclear power plant. After the official studies and the primary visit of the old seismograph stations it was found that the mentioned network doesn't cover completely the geological structures around the power plant. Therefore, new locations have been introduced through the field investigation and computational methods of optimization. In positioning the new stations, places with the least amount of noise and the best coverage for seismic sources were selected. The modeling with considering an imaginative station at the selected places shows that the thresholds of the complete records of earthquakes around Darkhovein site is under the magnitude 1 (about 0.8).

  13. The IAEA's activities in safeguarding nuclear materials and in developing internationally acceptable safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    Rometsch, Rudolf; Specter, Herschel

    1977-01-01

    Promoting the peaceful use of nuclear energy and aiming at the international sharing of its benefits are objectives that guide the activities of the Agency. But this promotional work is carried out on condition that security and safety are provided for. All Agency assistance involving nuclear facilities will be subjected to standards of safety or other standards, which are proposed by a State the Agency finds essentially equivalent. Safeguards are always applied on the basis of agreement. States party to NPT are obligated to negotiate and conclude with the Agency agreements which cover all their peaceful nuclear activities. Safeguards agreements concluded outside NPT are applied to specific supplies of facilities, equipment and material. To assist countries in laying down their nuclear safety regulations the Agency's program for the developing of codesof practice and safety guides for nuclear power plants draws up guidelines for governmental organizations, siting, design, operation and quality assurance. Codes are the fundamental documents laying down the objectives of each field of nuclear safety

  14. Guide on a national system for collecting, assessing and disseminating information on safety-related events in nuclear power plants

    International Nuclear Information System (INIS)

    1983-02-01

    There is a wide spectrum of safety significance in the events that can occur during nuclear power plant operations. It is important that lessons be learned from safety-related events (hereinafter referred to as unusual events) so as to improve the safety of nuclear power plants. Hence formal procedures should be established for this purpose. The purpose of this document is to provide guidance to Member States for establishing a system (hereinafter referred to as a national system) for collecting, storing, retrieving, assessing and disseminating information on unusual events in nuclear power plants. The guidance given is based on experience gained in the use of existing national and international systems. This guide covers a national system that is part of a programme to improve nuclear power plant safety using experience gained from operating plants both within and outside the country. Implementing the recommendations in this guide would render any national system compatible with other national systems and facilitate the participation in the IAEA System for Reporting Unusual Events with Safety Significance (hereinafter referred to as the IAEA Incident Reporting System, IAEA-IRS) for more widespread dissemination of lessons learned from nuclear power plant operation

  15. A Safety and Health Guide for Vocational Educators. Incorporating Requirements of the Occupational Safety and Health Act of 1970, Relevant Pennsylvania Requirements with Particular Emphasis for Those Concerned with Cooperative Education and Work Study Programs. Volume 15. Number 1.

    Science.gov (United States)

    Wahl, Ray

    Intended as a guide for vocational educators to incorporate the requirements of the Occupational Safety and Health Act (1970) and the requirements of various Pennsylvania safety and health regulations with their cooperative vocational programs, the first chapter of this document presents the legal implications of these safety and health…

  16. Seismic assessment of Technical Area V (TA-V).

    Energy Technology Data Exchange (ETDEWEB)

    Medrano, Carlos S.

    2014-03-01

    The Technical Area V (TA-V) Seismic Assessment Report was commissioned as part of Sandia National Laboratories (SNL) Self Assessment Requirement per DOE O 414.1, Quality Assurance, for seismic impact on existing facilities at Technical Area-V (TA-V). SNL TA-V facilities are located on an existing Uniform Building Code (UBC) Seismic Zone IIB Site within the physical boundary of the Kirtland Air Force Base (KAFB). The document delineates a summary of the existing facilities with their safety-significant structure, system and components, identifies DOE Guidance, conceptual framework, past assessments and the present Geological and Seismic conditions. Building upon the past information and the evolution of the new seismic design criteria, the document discusses the potential impact of the new standards and provides recommendations based upon the current International Building Code (IBC) per DOE O 420.1B, Facility Safety and DOE G 420.1-2, Guide for the Mitigation of Natural Phenomena Hazards for DOE Nuclear Facilities and Non-Nuclear Facilities.

  17. The health and safety implications of local medical support for land seismic crews in remote locations in southeast Asia

    International Nuclear Information System (INIS)

    Win, P.M.; Suter, P.C.

    1991-01-01

    The paper attempts to detail the benefits and drawbacks of hiring local doctors to support the medical services needed for land seismic acquisition crews in remote locations in South East Asia. The actual conditions prevailing among such seismic operations will be presented highlighting the problems and risks involved. The management of these problems will be outlined in terms of prevention and actual diagnosis and treatment of disease and injuries including emergency stabilization and evacuation of critically ill patients. The results and lessons learned will be evaluated and discussed including the economics of setting up a reasonably reliable medical facility. The paper will conclude that local knowledge, high levels of training and low costs make this type of medical support beneficial for such operations and may well be applicable for similar operations in other parts of the world

  18. Conduct of Operations at Nuclear Power Plants. Safety Guide (Spanish Edition); Realizacion de operaciones en centrales nucleares. Guia de seguridad

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This Safety Guide identifies the main responsibilities and practices of nuclear power plant (NPP) operations departments in relation to their responsibility for the safe functioning of the plant. The guide presents the factors to be considered in structuring the operations department of an NPP; setting high standards of performance; making safety related decisions in an effective manner; conducting control room and field activities in a thorough and professional manner; and maintaining an NPP within established operational limits and conditions. Contents: 1. Introduction; 2. Management and organization of plant operations; 3. Shift complement and functions; 4. Shift routines and operating practices; 5. Control of equipment and plant status; 6. Operations equipment and operator aids; 7. Work control and authorization.

  19. Safety and quality management for radiotherapy treatments - ASN guide nr 5 - Index 1 - Release of the 10/04/2009

    International Nuclear Information System (INIS)

    2009-01-01

    This guide aims at proposing a framework for the safety and quality management for radiotherapy treatments. It addresses the general requirements for the quality management system (general requirements and requirements related to documentation), the management responsibility, the resource management (human and hardware resources), the preparation and performance of activities allowing the taking into care of a patient from his first consultation to the post-treatment follow-up, the assessment, analysis and improvement of the quality management system

  20. Safety of transrectal ultrasound-guided prostate biopsy in patients affected by Crohn’s disease

    Directory of Open Access Journals (Sweden)

    Lucio Dell'Atti

    2017-06-01

    Full Text Available Purpose: Crohn’s disease (CD is a chronic inflammatory condition of the gastrointestinal tract. It is usually considered a contraindication to transrectal ultrasound-guided prostate biopsy (TRUSBx. The aim of this study was to investigate the safety of TRUSBx in a small cohort of patients with CD. Methods: We queried our institutional database clinical data of patients with a diagnosis of CD undergoing TRUSBx, and a retrospective prospective study of 5 patients was planned. All patients enrolled were in the remission phase of CD and asymptomatic. They received the same antibiotic prophylaxis and a povidone-iodine aqueous solution enema before the procedure. A standardized reproducible technique was used with using a ultrasound machine equipped with a 5-9 MHz multifrequency convex probe “end-fire”. The patients were treated under local anaesthesia, and a 14-core biopsy scheme was performed in each patient as first intention. After the procedure each patient was given a verbal numeric pain scale to evaluate tolerability of TRUSBx. Results: TRUSBx was successfully completed in all patients. The number of biopsy cores was 14 (12-16. Of the 5 biopsy procedures performed 40% revealed prostatic carcinoma (PCa with a Gleason score 6 (3+3. No patients required catheterization or admission to the hospital for adverse events after the procedure. The most frequent adverse event was hematospermia (60%, while hematuria was present in 20% of patients and a minimal rectal bleeding in 20% of the patients. No patients reported severe or unbearable pain (score ≥ 8. Conclusions: This study suggests that CD may not be an absolute contraindication to TRUSBx for prostate cancer detection, but still requires a careful patients selection.

  1. Adrenal neoplasms: Effectiveness and safety of CT-guided ablation of 23 tumors in 22 patients

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, Farrah J.; Dupuy, Damian E.; Machan, Jason T. [Department of Diagnostic Imaging and the Office of Research Administration, Alpert Medical School of Brown University, Rhode Island Hospital, 593 Eddy Street, Providence, RI 02903 (United States); Mayo-Smith, William W., E-mail: wmayo-smith@lifespan.org [Department of Diagnostic Imaging and the Office of Research Administration, Alpert Medical School of Brown University, Rhode Island Hospital, 593 Eddy Street, Providence, RI 02903 (United States)

    2012-08-15

    Purpose: To retrospectively evaluate the effectiveness and safety of image-guided ablation of adrenal tumors. Materials and methods: : This HIPAA-compliant study was IRB approved and informed consent was waived. From 5/1999-6/2008, 20 consecutive adrenal metastases (mean diameter 4.2 cm; range, 2-8) and 3 hormonally active primary adrenal tumors (mean diameter 2.3 cm; range, 1-4), including an aldosteronoma and 2 pheochromocytomas in 22 patients (14 men, 8 women; mean age 61 years; range 40-84) were ablated in 23 sessions. Bilateral metastases were treated in a single patient. Radiofrequency ablation was used to treat 16 adrenal metastases and the 3 hyperfunctioning tumors. Microwave ablation was used to treat 4 metastases. Successful treatment was defined as a lack of both enhancement on follow-up contrast enhanced CT and/or up-take on FDG PET-CT and for functioning tumors, resolution of biochemical abnormalities. Results: Technical success was achieved in all sessions. Mean follow-up was 45.1 months (range, 1-91) Local tumor progression (focal enhancement at ablation site {>=}1 cm in short axis) was detected in 4 of 23 tumors, two of which were identified bilaterally in a single patient prompting re-treatment. Of 19 patients with metastatic disease, 16 had fatal extra-adrenal disease progression, and 3 remain alive. Two of the 3 patients who underwent ablation of hyperfunctioning tumors remain alive, including the patient with an aldosteronoma who had recurrent symptoms 91 months post ablation. Intra-ablative hypertension occurred in 9% (2/23) of sessions and was successfully treated pharmacologically. Conclusion: Ablation of metastatic and hyperfunctioning adrenal tumors is safe and may provide local control and treatment of pathologic biochemical activity.

  2. Adrenal neoplasms: Effectiveness and safety of CT-guided ablation of 23 tumors in 22 patients

    International Nuclear Information System (INIS)

    Wolf, Farrah J.; Dupuy, Damian E.; Machan, Jason T.; Mayo-Smith, William W.

    2012-01-01

    Purpose: To retrospectively evaluate the effectiveness and safety of image-guided ablation of adrenal tumors. Materials and methods: : This HIPAA-compliant study was IRB approved and informed consent was waived. From 5/1999-6/2008, 20 consecutive adrenal metastases (mean diameter 4.2 cm; range, 2–8) and 3 hormonally active primary adrenal tumors (mean diameter 2.3 cm; range, 1–4), including an aldosteronoma and 2 pheochromocytomas in 22 patients (14 men, 8 women; mean age 61 years; range 40–84) were ablated in 23 sessions. Bilateral metastases were treated in a single patient. Radiofrequency ablation was used to treat 16 adrenal metastases and the 3 hyperfunctioning tumors. Microwave ablation was used to treat 4 metastases. Successful treatment was defined as a lack of both enhancement on follow-up contrast enhanced CT and/or up-take on FDG PET-CT and for functioning tumors, resolution of biochemical abnormalities. Results: Technical success was achieved in all sessions. Mean follow-up was 45.1 months (range, 1–91) Local tumor progression (focal enhancement at ablation site ≥1 cm in short axis) was detected in 4 of 23 tumors, two of which were identified bilaterally in a single patient prompting re-treatment. Of 19 patients with metastatic disease, 16 had fatal extra-adrenal disease progression, and 3 remain alive. Two of the 3 patients who underwent ablation of hyperfunctioning tumors remain alive, including the patient with an aldosteronoma who had recurrent symptoms 91 months post ablation. Intra-ablative hypertension occurred in 9% (2/23) of sessions and was successfully treated pharmacologically. Conclusion: Ablation of metastatic and hyperfunctioning adrenal tumors is safe and may provide local control and treatment of pathologic biochemical activity.

  3. Leveraging Educational, Research and Facility Expertise to Improve Global Seismic Monitoring: Preparing a Guide on Sustainable Networks

    Science.gov (United States)

    Nybade, A.; Aster, R.; Beck, S.; Ekstrom, G.; Fischer, K.; Lerner-Lam, A.; Meltzer, A.; Sandvol, E.; Willemann, R. J.

    2008-12-01

    Building a sustainable earthquake monitoring system requires well-informed cooperation between commercial companies that manufacture components or deliver complete systems and the government or other agencies that will be responsible for operating them. Many nations or regions with significant earthquake hazard lack the financial, technical, and human resources to establish and sustain permanent observatory networks required to return the data needed for hazard mitigation. Government agencies may not be well- informed about the short-term and long-term challenges of managing technologically advanced monitoring systems, much less the details of how they are built and operated. On the relatively compressed time scale of disaster recovery efforts, it can be difficult to find a reliable, disinterested source of information, without which government agencies may be dependent on partial information. If system delivery fails to include sufficient development of indigenous expertise, the performance of local and regional networks may decline quickly, and even data collected during an early high-performance period may be degraded or lost. Drawing on unsurpassed educational capabilities of its members working in close cooperation with its facility staff, IRIS is well prepared to contribute to sustainability through a wide variety of training and service activities that further promote standards for network installation, data exchange protocols, and free and open access to data. Members of the Consortium and staff of its Core Programs together could write a guide on decisions about network design, installation and operation. The intended primary audience would be government officials seeking to understand system requirements, the acquisition and installation process, and the expertise needed operate a system. The guide would cover network design, procurement, set-up, data use and archiving. Chapters could include advice on network data processing, archiving data (including

  4. Definition and means of maintaining the supply ventilation system seismic shutdown portion of the PFP safety envelope. Revision 2

    International Nuclear Information System (INIS)

    Keck, R.D.

    1995-01-01

    This report describes the modifications to the ventilation system for the Plutonium Finishing Plant. Topics discussed in this report include; system functional requirements, evaluations of equipment, a list of drawings showing the safety envelope boundaries; list of safety envelope equipment, functional requirements for individual safety envelope equipment, and a list of the operational, maintenance and surveillance procedures necessary to operate and maintain the system equipment

  5. LISA package user guide. Part II: LISA (Long Term Isolation Safety Assessment) program description and user guide

    International Nuclear Information System (INIS)

    Prado, P.; Saltelli, A.; Homma, T.

    1992-01-01

    This manual is subdivided into three parts. In this second part, this document describes the LISA (Long term Isolation Safety Assessment) Code and its submodels. LISA is a tool for analysis of the safety of an underground disposal of nuclear waste. It has the capability to handle nuclide chain of arbitrary length and to evaluate the migration of nuclide through a geosphere medium composed of an arbitrary number of segments. LISA makes use of Monte Carlo methodology to evaluate the uncertainty in the quantity being assessed (eg dose) arising from the uncertainty in the model input parameters. In the present version LISA is equipped with a very simple source term submodel, a relatively complex geosphere and a simplified biosphere. The code is closely associated with its statistical pre-processor code (PREP), which generates the input Monte Carlo sample from the assigned parameter probability density functions and with its post-processor code (SPOP) which provides useful statistics on the output sample (uncertainty and sensitivity analysis). This report describes the general structure of LISA, its subroutines and submodels, the code input ant output files. It is intended to provide the user with enough information to know and run the code as well as the capacity to incorporate different submodels. 15 refs., 6 figs

  6. Radiation safety in educational, medical and research institutions. Regulatory guide G-121

    International Nuclear Information System (INIS)

    2000-05-01

    This regulatory guide is intended to help educational, medical and research institutions design and implement radiation protection programs that meed regulatory requirements. This guide applied to educational, medical or research institutions that require a licence from the CNSC to posses or use radioactive materials. It describes programs to assure that radioactive materials are used safely during licensed activities. (author)

  7. Seismic Hazard Assessment and Uncertainties Treatment: Discussion on the current French regulation, practices and open issues

    International Nuclear Information System (INIS)

    Berge-Thierry, Catherine

    2014-01-01

    Taking into account the seismic risk in the context of nuclear safety in France is guided by the Fundamental Safety Rule (RFS2001-01) for the assessment of seismic hazard, and by the Guide ASN/2/01 for the design rules of civil engineering structures. These two references have been updated respectively in 2001 and 2006 and validated by the Nuclear Safety Authority. The French approach is anchored on a deterministic approach. We propose to recall the principles of the methodology recommended by the RFS 2001-01, and to illustrate the advantages and limitations highlighted in recent years. Indeed, this regulatory framework is used both in the design stage and for safety reassessment of all nuclear facilities, power reactors and experimental laboratories and factories. We focus on: (i) key parameters of the approach, and their level of knowledge, (ii) key steps and principles that lead to a non-homogeneous approach between various geographic sites, depending on the seismic activity and / or knowledge, (iii) on physical phenomena (such as the geometric extension of the seismic source, the complexity of earthquake rupture on the fault plane) that are not taken into account, or for which (2D and 3D site effects, and non-linear soil behavior under strong motions), the RFS 2001-01 approach does not provide any guidance, (iv) consideration of epistemic and random uncertainties. We discuss also the probabilistic approaches widely implemented both in France as recently to establish the seismic zoning (reference for the regulation of conventional building and classified installations for the environment), used worldwide and strongly supported by the international Atomic Energy Agency references (safety guides and guidelines). The Tohoku earthquake that occurred in Japan on March 11, 2011, triggering the tsunami that itself caused the nuclear accident at Fukushima Daiichi site has resulted in the realization in France of the Complementary Safety Studies as a request of the

  8. Treatment guided by rapid diagnostic tests for malaria in Tanzanian children: safety and alternative bacterial diagnoses

    Directory of Open Access Journals (Sweden)

    Sykes Alma

    2011-10-01

    Full Text Available Abstract Background WHO guidelines for the treatment of young children with suspected malaria have recently changed from presumptive treatment to anti-malarial treatment guided by a blood slide or malaria rapid diagnostic test (RDT. However, there is limited evidence of the safety of this policy in routine outpatient settings in Africa. Methods Children 3-59 months of age with a non-severe febrile illness and no obvious cause were enrolled over a period of one year in a malaria endemic area of Tanzania. Treatment was determined by the results of a clinical examination and RDT result, and blood culture and serum lactate were also collected. RDT-negative children were followed up over 14 days. Results Over the course of one year, 965 children were enrolled; 158 (16.4% were RDT-positive and treated with artemether-lumefantrine and 807 (83.4% were RDT-negative and treated with non-anti-malarial medicines. Compared with RDT-positives, RDT-negative children were on average younger with a lower axillary temperature and more likely to have a history of cough or difficulty in breathing. Six (0.6% children became RDT-positive after enrolment, all of whom were PCR-negative for Plasmodium falciparum DNA at enrolment. In addition, 12 (1.2% children were admitted to hospital, one with possible malaria, none of whom died. A bacterial pathogen was identified in 9/965 (0.9% children, eight of whom were RDT-negative and one was RDT-positive, but slide-negative. Excluding three children with Salmonella typhi, all of the children with bacteraemia were ≤12 months of age. Compared to double-read research slide results RDTs had a sensitivity of 97.8% (95%CI 96.9-98.7 and specificity of 96.3% (95%CI 96.3-98.4. Conclusions Use of RDTs to direct the use of anti-malarial drugs in young children did not result in any missed diagnoses of malaria although new infections soon after a consultation with a negative RDT result may undermine confidence in results. Invasive

  9. Seismic re-evaluation of Mochovce nuclear power plant. Seismic reevaluation of civil structures

    International Nuclear Information System (INIS)

    Podrouzek, P.

    1997-01-01

    In this contribution, an overview of seismic design procedures used for reassessment of seismic safety of civil structures at the Mochovce NPP in Slovak Republic presented. As an introduction, the objectives, history, and current status of seismic design of the NPP have been explained. General philosophy of design methods, seismic classification of buildings, seismic data, calculation methods, assumptions on structural behavior under seismic loading and reliability assessment were described in detail in the subsequent section. Examples of calculation models used for dynamic calculations of seismic response are given in the last section. (author)

  10. Study on seismic design margin based upon inelastic shaking test of the piping and support system

    International Nuclear Information System (INIS)

    Ishiguro, Takami; Eto, Kazutoshi; Ikeda, Kazutoyo; Yoshii, Toshiaki; Kondo, Masami; Tai, Koichi

    2009-01-01

    In Japan, according to the revised Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, September 2006, criteria of design basis earthquakes of Nuclear Power Reactor Facilities become more severe. Then, evaluating seismic design margin took on a great importance and it has been profoundly discussed. Since seismic safety is one of the major key issues of nuclear power plant safety, it has been demonstrated that nuclear piping system possesses large safety margins by various durability test reports for piping in ultimate conditions. Though the knowledge of safety margin has been accumulated from these reports, there still remain some technical uncertainties about the phenomenon when both piping and support structures show inelastic behavior in extremely high seismic excitation level. In order to obtain the influences of inelastic behavior of the support structures to the whole piping system response when both piping and support structures show inelastic behavior, we examined seismic proving tests and we conducted simulation analyses for the piping system which focused on the inelastic behavior of the support to the whole piping system response. This paper introduces major results of the seismic shaking tests of the piping and support system and the simulation analyses of these tests. (author)

  11. Schedules of Provisions of the IAEA Regulations for the Safe Transport of Radioactive Material (2009 Ed.). Safety Guide

    International Nuclear Information System (INIS)

    2014-01-01

    This Safety Guide is issued in support of Regulations for the Safe Transport of Radioactive Material (IAEA Safety Standards Series No. TS-R-1, 2009 Edition). It lists the paragraph numbers of the Transport Regulations that are relevant for specified types of consignment, classified according to their UN numbers. It does not provide additional recommendations. The intended users are consignors and consignees, carriers, shippers, regulators, and end users involved in the transport of radioactive material. A person or organization intending to transport a particular type of consignment of radioactive material must meet requirements in all sections of the Transport Regulations. This Safety Guide aids users by providing a listing of the relevant requirements of the Transport Regulations for each type of radioactive material, package or shipment. Once a consignor has classified the radioactive material to be shipped, the appropriate UN number can be assigned and the paragraph numbers of the requirements that apply for the shipment can be found in the corresponding schedule

  12. Design report on the guide box-reactivity and safety control plates for MPR reactor under normal operation conditions

    International Nuclear Information System (INIS)

    Markiewicz, M.

    1999-01-01

    The reactivity control system for the MPR reactor (Multi Purpose Reactor) is a critical component regarding safety, it must ensure a fast shut down, maintaining the reactor in subcritical condition under normal or accidental operation condition. For this purpose, this core component must be designed to maintain its operating capacity during all the residence time and under any foreseen operation condition. The mechanical design of control plates and guide boxes must comply with structural integrity, maintaining its geometric and dimensional stability within the pre-established limits to prevent interferences with other core components. For this, the heat generation effect, mechanical loads and environment and irradiation effects were evaluated during the mechanical design. The reactivity control system is composed of guide boxes, manufactured from Aluminium alloy, located between the fuel elements, and control absorber plates of Ag-In-Cd alloy hermetically enclosed by a cladding of stainless steel sliding inside de guide boxes. The upward-downward movement is transmitted by a rod from the motion device located at the reactor lower part. The design requirements, criteria and limits were established to fulfill with the normal and abnormal operation conditions. The design verifications were performed by analytical method, estimating the guide box and control plates residence time. The result of the analysis performed, shows that the design of the reactivity control system and the material selected, are appropriate to fulfill the functional requirements, with no failures attributed to the mechanical design. (author)

  13. Percutaneous Image-guided radiofrequency ablation of tumors in inoperable patients - immediate complications and overall safety

    Directory of Open Access Journals (Sweden)

    Anubha Sahay

    2016-01-01

    Conclusions: Percutaneous image-guided RFA is an option in patients where most other tumor management modalities have been exhausted or rejected. RFA may not be free from side effects such as postablation syndrome, pain, and there may be other serious complications such as bleeding, but based on our observations, percutaneous image-guided RFA of tumors is a safe palliative and therapeutic treatment option.

  14. Radiation protection and safety guide no. GRPB-G-5: safe use of x-rays

    International Nuclear Information System (INIS)

    Schandorf, C.; Darko, O.; Yeboah, J.; Osei, E.K.; Asiamah, S.D.

    1998-01-01

    If properly utilized, the use of x-rays can be instrumental in the improvement of the health and welfare of the public. This regulatory guide was developed to assist and encourage registrants in the safe and constructive use of x-rays and to prohibit and prevent exposure to ionizing radiation in amounts which are or may be detrimental to health. The present guide applies to the use of x-rays for diagnostic, therapeutic, and non medical purposes

  15. Seismic fragility capacity of equipment

    International Nuclear Information System (INIS)

    Iijima, Toru; Abe, Hiroshi; Suzuki, Kenichi

    2006-01-01

    Seismic probabilistic safety assessment (PSA) is an available method to evaluate residual risks of nuclear plants that are designed on definitive seismic conditions. From our preliminary seismic PSA analysis, horizontal shaft pumps are important components that have significant influences on the core damage frequency (CDF). An actual horizontal shaft pump and some kinds of elements were tested to evaluate realistic fragility capacities. Our test results showed that the realistic fragility capacity of horizontal shaft pump would be at least four times as high as a current value, 1.6 x 9.8 m/s 2 , used for our seismic PSA. We are going to incorporate the fragility capacity data that were obtained from those tests into our seismic PSA analysis, and we expect that the reliability of seismic PSA should increase. (author)

  16. L-038: EPR-First Responders: Forces / safety equipment. Action Guides

    International Nuclear Information System (INIS)

    2011-01-01

    This conference is about the actions carry out by the forces and the safety equipment in a radiological emergency. The security area, the victims, the hospitals, the police vehicles area, the safety cordon, the evacuation, the contamination level and the risk of life are important aspects to be considered by the first responders.

  17. Guide for preparing annual reports on radiation-safety testing of electronic products (general)

    International Nuclear Information System (INIS)

    1987-10-01

    For manufacturers of electronic products other than those for which a specific guide has been issued, the guide replaces the Guide for the Filing of Annual Reports (21 CFR Subchapter J, Section 1002.11), HHS Publication FDA 82-8127. The electronic product (general) annual reporting guide is applicable to the following products: products intended to produce x radiation (accelerators, analytical devices, therapy x-ray machines); microwave diathermy machines; cold-cathode discharge tubes; and vacuum switches and tubes operating at or above 15,000 volts. To carry out its responsibilities under Public Law 90-602, the Food and Drug Administration's Center for Devices and Radiological Health (CDRH) has issued a series of regulations contained in Title 21 of the Code of Federal Regulations (CFR). Part 1002 of 21 CFR deals with records and reports. Section 1002.61 categorizes electronic products into Groups A through C. Section 1002.30 requires manufacturers of products in Groups B and C to establish and maintain certain records, while Section 1002.11 requires such manufacturers to submit an Annual Report summarizing the contents of the required records. Section 1002.7 requires that reports conform to reporting guides issued by CDRH unless an acceptable justification for an alternate format is provided

  18. Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

    Energy Technology Data Exchange (ETDEWEB)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

    2005-03-15

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments.

  19. Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

    International Nuclear Information System (INIS)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

    2005-03-01

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments

  20. Application of the Management System for Facilities and Activities. Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication provides guidance for following the requirements for management systems that integrate safety, health, security, quality assurance and environmental objectives. A successful management system ensures that nuclear safety matters are not dealt with in isolation but are considered within the context of all these objectives. The aim of this publication is to assist Member States to establish and implement effective management systems that coherently integrate all aspects of managing nuclear facilities and activities.

  1. Seismic analysis - what goal

    International Nuclear Information System (INIS)

    Tagart, S.W.

    1978-01-01

    The seismic analysis of nuclear components is characterized today by extensive engineering computer calculations in order to satisfy both the component standard codes such as ASME III as well as federal regulations and guides. The current nuclear siesmic design procedure has envolved in a fragmented fashion and continues to change its elements as improved technology leads to changing standards and guides. The dominant trend is a monotonic increase in the overall conservation with time causing a similar trend in costs of nuclear power plants. Ironically the improvements in the state of art are feeding a process which is eroding the very incentives that attracted us to nuclear power in the first place. This paper examines the cause of this process and suggests that what is needed is a realistic goal which appropriately addresses the overall uncertainty of the seismic design process. (Auth.)

  2. Seismic Ecology

    Science.gov (United States)

    Seleznev, V. S.; Soloviev, V. M.; Emanov, A. F.

    The paper is devoted to researches of influence of seismic actions for industrial and civil buildings and people. The seismic actions bring influence directly on the people (vibration actions, force shocks at earthquakes) or indirectly through various build- ings and the constructions and can be strong (be felt by people) and weak (be fixed by sensing devices). The great number of work is devoted to influence of violent seismic actions (first of all of earthquakes) on people and various constructions. This work is devoted to study weak, but long seismic actions on various buildings and people. There is a need to take into account seismic oscillations, acting on the territory, at construction of various buildings on urbanized territories. Essential influence, except for violent earthquakes, man-caused seismic actions: the explosions, seismic noise, emitted by plant facilities and moving transport, radiation from high-rise buildings and constructions under action of a wind, etc. can exert. Materials on increase of man- caused seismicity in a number of regions in Russia, which earlier were not seismic, are presented in the paper. Along with maps of seismic microzoning maps to be built indicating a variation of amplitude spectra of seismic noise within day, months, years. The presence of an information about amplitudes and frequencies of oscillations from possible earthquakes and man-caused oscillations in concrete regions allows carry- ing out soundly designing and construction of industrial and civil housing projects. The construction of buildings even in not seismically dangerous regions, which have one from resonance frequencies coincident on magnitude to frequency of oscillations, emitted in this place by man-caused objects, can end in failure of these buildings and heaviest consequences for the people. The practical examples of detail of engineering- seismological investigation of large industrial and civil housing projects of Siberia territory (hydro power

  3. Dispersion of radioactive material in air and water and consideration of population distribution in site evaluation for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    The IAEA issues Safety Requirements and Safety Guides pertaining to nuclear power plants and activities in the field of nuclear energy, on the basis of its Safety Fundamentals publication on The Safety of Nuclear Installations. The present Safety Guide, which supplements the Code on the Safety of Nuclear Power Plants: Siting, concerns the effects of a nuclear power plant on the surrounding region and the consideration of population distribution in the siting of a plant. This Safety Guide makes recommendations on how to meet the requirements of the Code on the Safety of Nuclear Power Plants: Siting, on the basis of knowledge of the mechanisms for the dispersion of effluents discharged into the atmosphere and into surface water and groundwater. Relevant site characteristics and safety considerations are discussed. Population distribution, the projected population growth rate, particular geographical features, the capabilities of local transport networks and communications networks, industry and agriculture in the region, and recreational and institutional activities in the region should be considered in assessing the feasibility of developing an emergency response plan. In the selection of a site for a facility using radioactive material, such as a nuclear power plant, account should be taken of any local features that might be affected by the facility and of the feasibility of off-site intervention, including emergency response and protective actions. This is in addition to the evaluation of any features of the site itself that might affect the safety of the facility. This Safety Guide recommends methods for the assessment of regional and local characteristics. This Safety Guide supersedes four earlier IAEA Safety Guides, namely: Atmospheric Dispersion in Nuclear Power Plant Siting (Safety Series No. 50-SG-S3 (1980)); Site Selection and Evaluation for Nuclear Power Plants with Respect to Population Distribution (Safety Series No. 50-SG-S4 (1980)); Hydrological

  4. Dispersion of radioactive material in air and water and consideration of population distribution in site evaluation for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The IAEA issues Safety Requirements and Safety Guides pertaining to nuclear power plants and activities in the field of nuclear energy, on the basis of its Safety Fundamentals publication on The Safety of Nuclear Installations. The present Safety Guide, which supplements the Code on the Safety of Nuclear Power Plants: Siting, concerns the effects of a nuclear power plant on the surrounding region and the consideration of population distribution in the siting of a plant. This Safety Guide makes recommendations on how to meet the requirements of the Code on the Safety of Nuclear Power Plants: Siting, on the basis of knowledge of the mechanisms for the dispersion of effluents discharged into the atmosphere and into surface water and groundwater. Relevant site characteristics and safety considerations are discussed. Population distribution, the projected population growth rate, particular geographical features, the capabilities of local transport networks and communications networks, industry and agriculture in the region, and recreational and institutional activities in the region should be considered in assessing the feasibility of developing an emergency response plan. In the selection of a site for a facility using radioactive material, such as a nuclear power plant, account should be taken of any local features that might be affected by the facility and of the feasibility of off-site intervention, including emergency response and protective actions. This is in addition to the evaluation of any features of the site itself that might affect the safety of the facility. This Safety Guide recommends methods for the assessment of regional and local characteristics. This Safety Guide supersedes four earlier IAEA Safety Guides, namely: Atmospheric Dispersion in Nuclear Power Plant Siting (Safety Series No. 50-SG-S3 (1980)). Site Selection and Evaluation for Nuclear Power Plants with Respect to Population Distribution (Safety Series No. 50-SG-S4 (1980)). Hydrological

  5. Advisory material for the IAEA regulations for the safe transport of radioactive material. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    Since the first edition in 1961, the Regulations for the Safe Transport of Radioactive Material of the IAEA (IAEA Regulations) have served as the basis of safety for the transport of radioactive material worldwide. In the discussions leading to the first edition of the IAEA Regulations, it was realized that there was need for a publication to supplement the Regulations which could give information of individual provisions as to their purpose, their scientific background and how to apply them in practice. In response, the Agency published Safety Series No. 7, entitled, in its first edition in 1961, 'Notes on Certain Aspects of the Regulations'. An additional source of information on the Regulations, providing advice on 'how' the user should comply with them which could be augmented from time to time in the light of latest experience, was provided by the Agency, initially in relation to the 1973 edition of the Regulations. This was entitled 'Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material' and designated Safety Series No. 37. This document is the result of combining the two Safety Series in a single publication. Thus the primary purpose of this publication is to provide guidance to users on proven and acceptable ways of complying with the Regulations. This Advisory Material is not a stand-alone text and it only has significance when used as a companion to the IAEA Safety Standards Series No. ST-1, Regulations for the Safe Transport of Radioactive Material (1996 edition)

  6. Main features of the unit 1-4 building complex, Kozloduy NPP in respect to seismic safety

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M; Boncheva, H; Stafanov, D [Central Laboratory for Seismic Mechanics and Earthquake Engineering, Bulgarian Academy of Sciences, Sofia (Bulgaria)

    1993-07-01

    The Units 1 and 2 of Kozloduy NPP were originally designed to resist a IV-V degree MSK earthquake. They have been subsequently upgraded for a VII degree earthquake. Since that structure basically do not meet the safety requirements to resist the new earthquake with a maximum acceleration of 0.2 g and very broad spectrum. The performed analyses are clearly pointing out that an upgrading for the new earthquake level is possible. The problems common for all the structures of Kozloduy NPP are summarized in this presentation.

  7. Main features of the unit 1-4 building complex, Kozloduy NPP in respect to seismic safety

    International Nuclear Information System (INIS)

    Kostov, M.; Boncheva, H.; Stafanov, D.

    1993-01-01

    The Units 1 and 2 of Kozloduy NPP were originally designed to resist a IV-V degree MSK earthquake. They have been subsequently upgraded for a VII degree earthquake. Since that structure basically do not meet the safety requirements to resist the new earthquake with a maximum acceleration of 0.2 g and very broad spectrum. The performed analyses are clearly pointing out that an upgrading for the new earthquake level is possible. The problems common for all the structures of Kozloduy NPP are summarized in this presentation

  8. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  9. Challenges Ahead for Nuclear Facility Site-Specific Seismic Hazard Assessment in France: The Alternative Energies and the Atomic Energy Commission (CEA) Vision

    Science.gov (United States)

    Berge-Thierry, C.; Hollender, F.; Guyonnet-Benaize, C.; Baumont, D.; Ameri, G.; Bollinger, L.

    2017-09-01

    Seismic analysis in the context of nuclear safety in France is currently guided by a pure deterministic approach based on Basic Safety Rule ( Règle Fondamentale de Sûreté) RFS 2001-01 for seismic hazard assessment, and on the ASN/2/01 Guide that provides design rules for nuclear civil engineering structures. After the 2011 Tohohu earthquake, nuclear operators worldwide were asked to estimate the ability of their facilities to sustain extreme seismic loads. The French licensees then defined the `hard core seismic levels', which are higher than those considered for design or re-assessment of the safety of a facility. These were initially established on a deterministic basis, and they have been finally justified through state-of-the-art probabilistic seismic hazard assessments. The appreciation and propagation of uncertainties when assessing seismic hazard in France have changed considerably over the past 15 years. This evolution provided the motivation for the present article, the objectives of which are threefold: (1) to provide a description of the current practices in France to assess seismic hazard in terms of nuclear safety; (2) to discuss and highlight the sources of uncertainties and their treatment; and (3) to use a specific case study to illustrate how extended source modeling can help to constrain the key assumptions or parameters that impact upon seismic hazard assessment. This article discusses in particular seismic source characterization, strong ground motion prediction, and maximal magnitude constraints, according to the practice of the French Atomic Energy Commission. Due to increases in strong motion databases in terms of the number and quality of the records in their metadata and the uncertainty characterization, several recently published empirical ground motion prediction models are eligible for seismic hazard assessment in France. We show that propagation of epistemic and aleatory uncertainties is feasible in a deterministic approach, as in a

  10. Regulatory Guide 1.79 safety injection recirculation test requirements, fact or fiction

    International Nuclear Information System (INIS)

    Roberts, J.K.

    1976-01-01

    The overwhelming concern of the general public in this day of state nuclear initiatives is the basic question, ''is nuclear power safe.'' Much of this concern has focused on the emergency core cooling systems. This public attention spotlights the testing organization's responsibility during startup of proving the operation and reliability of the emergency core cooling systems. The standard established by the Nuclear Regulatory Commission for testing emergency core cooling systems is Regulatory Guide 1.79 ''Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors''. The nuclear industry must satisfy the testing requirements of Regulatory Guide 1.79 to meet their responsibility to the public; and to prevent future embarrassment when questioned on the adequacy of emergency core cooling systems

  11. Safety of Ultrasound-Guided Botulinum Toxin Injections for Sialorrhea as Performed by Pediatric Otolaryngologists.

    Science.gov (United States)

    Shariat-Madar, Bahbak; Chun, Robert H; Sulman, Cecille G; Conley, Stephen F

    2016-05-01

    To evaluate incidence of complications and hospital readmission as a result of ultrasound-guided botulinum toxin injections to manage sialorrhea. Case series with chart review. Children's Hospital of Wisconsin. A case series with chart review was performed of all cases of ultrasound-guided injection of botulinum toxin by pediatric otolaryngologists from March 5, 2010, to September 26, 2014,. Primary outcomes included complications such as dysphagia, aspiration pneumonia, and motor paralysis. Secondary outcomes included hospitalization, intubation, and nasogastric tube placement. There were 48 patients, 111 interventions, and 306 intraglandular injections identified. Botulinum toxin type A and type B were utilized in 4 and 107 operative interventions, respectively. Type A was injected into 4 parotid and 4 submandibular glands, utilizing doses of 20 U per parotid and 30 U per submandibular gland. Type B was injected into 98 parotid and 200 submandibular glands, with average dosing of 923 U per parotid and 1170 U per submandibular gland, respectively. There were 2 instances of subjectively worsening of baseline dysphagia that self-resolved. No cases were complicated by aspiration pneumonia or motor paralysis. No patients required hospital readmission, intubation, or nasogastric tube placement. Prior published data indicated 16% complication incidence with ultrasound-guided injection of botulinum toxin. Our study found a low complication rate (0.6%) with ultrasound-guided injections of botulinum toxin to manage sialorrhea, without cases of aspiration pneumonia or motor paralysis. Of 306 intraglandular injections, there were 2 cases of worsening baseline subjective dysphagia that self-resolved. © American Academy of Otolaryngology—Head and Neck Surgery Foundation 2016.

  12. Guide for reviewing safety analysis reports for packaging: Review of quality assurance requirements

    International Nuclear Information System (INIS)

    Moon, D.W.

    1988-10-01

    This review section describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging which are important to safety. The design effort, operation's plans, and quality assurance requirements should be integrated to achieve a system in which the independent QA program is not overly stringent and the application of QA requirements is commensurate with safety significance. The reviewer must verify that the applicant's QA section in the SARP contains package-specific QA information required by DOE Orders and federal regulations that demonstrate compliance. 8 refs

  13. Environment, Safety, and Health (ES&H) self-assessment guide

    Energy Technology Data Exchange (ETDEWEB)

    Reese, R.T.; Golden, N.L.; Romero, J.R.; Yesner, S.

    1997-06-01

    This document has been prepared as a guide for conducting self-assessments of ES&H functional programs and organizational (line) implementation of these programs. This guide is intended for use by individuals and/or teams involved in or familiar with ES&H programs and line operations (e.g., the {open_quotes}self{close_quotes}in self-assessment). Essential elements of the self-assessment process are described including: schedule and priorities, scope and approach, assessment criteria (e.g., performance objectives and measures), information gathering and analysis techniques, and documentation of planning efforts and results. The appendices in this guide include: (1) an assessment prioritization process, (2) generic performance objectives for line implementation and for ES&H functional programs, (3) sources for ES&H assessment information, (4) systemic factors (developed for SNL`s root cause analysis program), (5) Lockheed Martin audit questions for management systems, compliance and validation, and specific areas and concerns, (6) DOE facility representatives checklist, and (7) assessment tools and resources developed at SNL and other DOE/Lockheed Martin sites. This document is a product of the efforts associated with the SNL ES&H Oversight Pilot Project conducted from June 1995 to January 1997. This Pilot was part of the overall initiative by DOE to reduce burdensome agency oversight by placing greater reliance on contractor self-assessment.

  14. Study on structural seismic margin and probabilistic seismic risk. Development of a structural capacity-seismic risk diagram

    International Nuclear Information System (INIS)

    Nakajima, Masato; Ohtori, Yasuki; Hirata, Kazuta

    2010-01-01

    Seismic margin is extremely important index and information when we evaluate and account seismic safety of critical structures, systems and components quantitatively. Therefore, it is required that electric power companies evaluate the seismic margin of each plant in back-check of nuclear power plants in Japan. The seismic margin of structures is usually defined as a structural capacity margin corresponding to design earthquake ground motion. However, there is little agreement as to the definition of the seismic margin and we have no knowledge about a relationship between the seismic margin and seismic risk (annual failure probability) which is obtained in PSA (Probabilistic Safety Assessment). The purpose of this report is to discuss a definition of structural seismic margin and to develop a diagram which can identify a relation between seismic margin and seismic risk. The main results of this paper are described as follows: (1) We develop seismic margin which is defined based on the fact that intensity of earthquake ground motion is more appropriate than the conventional definition (i.e., the response-based seismic margin) for the following reasons: -seismic margin based on earthquake ground motion is invariant where different typed structures are considered, -stakeholders can understand the seismic margin based on the earthquake ground motion better than the response-based one. (2) The developed seismic margin-risk diagram facilitates us to judge easily whether we need to perform detailed probabilistic risk analysis or only deterministic analysis, given that the reference risk level although information on the uncertainty parameter beta is not obtained. (3) We have performed numerical simulations based on the developed method for four sites in Japan. The structural capacity-risk diagram differs depending on each location because the diagram is greatly influenced by seismic hazard information for a target site. Furthermore, the required structural capacity

  15. Seismic fragility analyses

    International Nuclear Information System (INIS)

    Kostov, Marin

    2000-01-01

    In the last two decades there is increasing number of probabilistic seismic risk assessments performed. The basic ideas of the procedure for performing a Probabilistic Safety Analysis (PSA) of critical structures (NUREG/CR-2300, 1983) could be used also for normal industrial and residential buildings, dams or other structures. The general formulation of the risk assessment procedure applied in this investigation is presented in Franzini, et al., 1984. The probability of failure of a structure for an expected lifetime (for example 50 years) can be obtained from the annual frequency of failure, β E determined by the relation: β E ∫[d[β(x)]/dx]P(flx)dx. β(x) is the annual frequency of exceedance of load level x (for example, the variable x may be peak ground acceleration), P(fI x) is the conditional probability of structure failure at a given seismic load level x. The problem leads to the assessment of the seismic hazard β(x) and the fragility P(fl x). The seismic hazard curves are obtained by the probabilistic seismic hazard analysis. The fragility curves are obtained after the response of the structure is defined as probabilistic and its capacity and the associated uncertainties are assessed. Finally the fragility curves are combined with the seismic loading to estimate the frequency of failure for each critical scenario. The frequency of failure due to seismic event is presented by the scenario with the highest frequency. The tools usually applied for probabilistic safety analyses of critical structures could relatively easily be adopted to ordinary structures. The key problems are the seismic hazard definitions and the fragility analyses. The fragility could be derived either based on scaling procedures or on the base of generation. Both approaches have been presented in the paper. After the seismic risk (in terms of failure probability) is assessed there are several approaches for risk reduction. Generally the methods could be classified in two groups. The

  16. Control of Orphan Sources and Other Radioactive Material in the Metal Recycling and Production Industries. Specific Safety Guide (Arabic Edition)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-09-01

    Accidents involving orphan sources and other radioactive material in the metal recycling and production industries have resulted in serious radiological accidents as well as in harmful environmental, social and economic impacts. This Safety Guide provides recommendations, the implementation of which should prevent such accidents and provide confidence that scrap metal and recycled products are safe. Contents: 1. Introduction; 2. Protection of people and the environment; 3. Responsibilities; 4. Monito