Water reactor safety research program. A description of current and planned research

International Nuclear Information System (INIS)

1978-07-01

The U.S. Nuclear Regulatory Commission (NRC) sponsors confirmatory safety research on lightwater reactors in support of the NRC regulatory program. The principal responsibility of the NRC, as implemented through its regulatory program is to ensure that public health, public safety, and the environment are adequately protected. The NRC performs this function by defining conditions for the use of nuclear power and by ensuring through technical review, audit, and follow-up that these conditions are met. The NRC research program provides technical information, independent of the nuclear industry, to aid in discharging these regulatory responsibilities. The objectives of NRC's research program are the following: (1) to maintain a confirmatory research program that supports assurance of public health and safety, and public confidence in the regulatory program, (2) to provide objectively evaluated safety data and analytical methods that meet the needs of regulatory activities, (3) to provide better quantified estimates of the margins of safety for reactor systems, fuel cycle facilities, and transportation systems, (4) to establish a broad and coherent exchange of safety research information with other Federal agencies, industry, and foreign organization. Current and planned research toward these goals is described

Bibliography of reports on research sponsored by the NRC Office of Nuclear Regulatory Research, November 1975--June 1976

Energy Technology Data Exchange (ETDEWEB)

Buchanan, J.R.

1976-09-30

A bibliography of 152 reports published by contractors of the NRC Office of Nuclear Regulatory Research during the period November 1975 through June 1976 is presented along with abstracts from the Nuclear Safety Information Center computer file. The bibliography has been sorted into the subject categories used by NRC to organize the research program. Within the subject categories, the reports are sorted by contractor organization and then chronologically. A brief description of the NRC research program precedes the bibliography.

Bibliography of reports on research sponsored by the NRC Office of Nuclear Regulatory Research, November 1975--June 1976

International Nuclear Information System (INIS)

Buchanan, J.R.

1976-01-01

A bibliography of 152 reports published by contractors of the NRC Office of Nuclear Regulatory Research during the period November 1975 through June 1976 is presented along with abstracts from the Nuclear Safety Information Center computer file. The bibliography has been sorted into the subject categories used by NRC to organize the research program. Within the subject categories, the reports are sorted by contractor organization and then chronologically. A brief description of the NRC research program precedes the bibliography

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

International Nuclear Information System (INIS)

1987-05-01

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

NRC/RSR data bank program

International Nuclear Information System (INIS)

Bankert, S.F.; Evans, C.D.; Hardy, H.A.; Litteer, G.L.; Schulz, G.L.; Smith, N.C.

1978-01-01

The United States Nuclear Regulatory Commission (NRC) has established at the Idaho National Engineering Laboratory (INEL) the NRC/Reactor Safety Research (RSR) Data Bank Program. The program is under the direction of EG and G Idaho, Inc., and is intended to provide the means of collecting, processing, and making available experimental data from the many water reactor safety research programs. The NRC/RSR Data Bank Program collects qualified engineering data on a prioritized basis from experimental program data bases, stores the data in a single data bank in a common format, and makes the data available to users. The NRC/RSR Data Bank specializes in water reactor safety experimental data, but it has a number of other scientific applications where large amounts of numeric data are or will be available. As an example of size, a single water reactor safety test may generate 10 million data words. Future examples of the use of a data bank might be in gathering data on low head hydraulics, solar projects, and liquid metal reactor safety data

NRC/RSR Data Bank Program description

International Nuclear Information System (INIS)

Bankert, S.F.

1979-01-01

The United States Nuclear Regulatory Commission (NRC) has established the NRC/Reactor Safety Research (RSR) Data Bank Program to collect, store, and make available data from the many domestic and foreign water reactor safety research programs. Local direction of the program is provided by EG and G Idaho, Inc., at Idaho National Engineering Laboratory. The NRC/RSR Data Bank Program provides a central computer storage mechanism and access software for data to be used by code development and assessment groups in meeting the code and correlation needs of the nuclear industry. The administrative portion of the program provides data entry, documentation, and training and advisory services to users and the NRC. The NRC/RSR Data Bank Program and the capabilities of the data access software are described

Changing emphasis at the NRC's Office of Nuclear Regulatory Research

International Nuclear Information System (INIS)

Remick, F.J.

1994-01-01

One of the major objectives of the Office of Research is to ensure availability of sound technical information for timely decision making in support of the NRC's safety mission. The Office of Research is changing some of its emphasis to better meet the expected needs of the NRC's regulatory offices. Long-standing programs in support of operating reactors are nearing completion. These programs include plant aging and severe accident research for currently operating plants. This meeting will also address the new challenges faced by the NRC in its review of the advanced light water and non-light water reactors. As plant aging and severe accident research programs are nearing completion, the research activities are coming to focus on the emerging technologies, for example, digital instrumentation and control systems, both as replacement equipment for operating plants and as the technology of choice and necessity for the advanced reactors. Necessity, because analog equipment is becoming obsolete. Other examples include the use of new materials in operating plants, human factors considerations in the design and operation of the advanced plants, thermal-hydraulic characteristics of the advanced reactors, and new construction techniques

HTGR safety research concerns at NRC

International Nuclear Information System (INIS)

Minogue, R.B.

1982-01-01

A general discussion of HTGR technical and safety-related problems is given. The broad areas of current research programs specific to the Fort St. Vrain reactor and applicable to HTGR technology are summarized

NRC/AMRMC Resident Research Associateship Program

Science.gov (United States)

2016-04-01

8. Curcumin : A Prototype Anti-inflammatory Therapeutic for Burn Pain and Wound Healing. Burn and Trauma Research Workgroup. BAMMC Burn Center 2014 9...from Burkholderia infection in mice. 9) PUBLICATIONS AND PAPERS RESULTING FROM NRC ASSOCIATESHIP RESEARCH Provide complete citations: author(s), title...PUBLICATIONS AND PAPERS RESULTING FROM NRC ASSOCIATESHIP RESEARCH Provide complete citations: author(s), title, full name of journal, volume number, page

NRC/RSR Data Bank Program

International Nuclear Information System (INIS)

Bankert, S.F.; Evans, C.D.; Hardy, H.A.; Litteer, G.L.; Schulz, G.L.; Smith, N.C.

1978-01-01

The United States Nuclear Regulatory Commission (NRC) has established the NRC/Reactor Safety Research (RSR) Data Bank Program at the Idaho National Engineering Laboratory (INEL). The program provides the means of collecting, storing, and making available experimental data from the many water reactor safety research programs in the United States and other countries. The program collects qualified engineering data on a prioritized basis from experimental program data bases, stores the data in a single data bank in a common format, and makes the data available to users

NRC/AMRMC Resident Research Associateship Program

Science.gov (United States)

2018-05-01

conducted the following activities in support of the subject contract: Outreach and Promotion The promotional schedule to advertise the NRC Research...Approved for Public Release; Distribution Unlimited 13. SUPPLEMENTARY NOTES 14. ABSTRACT During this reporting period, the NRC promoted research...Associateship Programs included the following: 1) attendance at meetings of major scientific and engineering professional societies; 2) advertising in

Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities

International Nuclear Information System (INIS)

Garvin, L.J.

1995-11-01

This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement

NRC confirmatory safety system testing in support of AP600 design review

International Nuclear Information System (INIS)

Rhee, G.S.; Bessette, D.E.; Shotkin, L.M.

1994-01-01

Westinghouse Electric Corporation has submitted the Advanced Passive 600 MWe (AP600) nuclear power plant design to the NRC for design certification. The Office of Nuclear Regulatory Research is proceeding to conduct confirmatory testing to help the NRC staff evaluate the AP600 safety system design. For confirmatory testing, it was determined that the cost-effective route was to modify an existing full-height, full-pressure test facility rather than build a new one. Thus, all the existing integral effects test facilities, both in the US and abroad, were screened to select the best candidate. As a result, the ROSA-V (Rig of Safety Assessment-V) test facility located in the Japan Atomic Energy Research Institute (JAERI) was chosen. However, because of some differences in design between the existing ROSA-V facility and the AP600, the ROSA-V is being modified to conform to the AP600 safety system design. The modification work will be completed by the end of this year. A series of facility characterization tests will then be performed in January 1994 for the modified part of the facility before the main test series is initiated in February 1994. A total of 12 tests will be performed in 1994 under Phase I of this cooperative program with JAERI. Phase II testing is being considered to be conducted in 1995 mainly for beyond-design-basis accident evaluation

NRC safety research in support of regulation

International Nuclear Information System (INIS)

1994-06-01

This report, the ninth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1993. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with sound technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission's rules, regulatory guides, or other guidance

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

NRC/RSR Data Bank Program

International Nuclear Information System (INIS)

Bankert, S.F.; Evans, C.D.; Hardy, H.A.; Litteer, G.L.; Schulz, G.L.; Smith, N.C.

1978-01-01

The United States Nuclear Regulatory Commission (NRC) has established the NRC/Reactor Safety Research (RSR) Data Bank Program to provide a means of collecting, processing, and making available experimental data from the many domestic and foreign water reactor safety research programs. The NRC/RSR Data Bank Program collects qualified engineering data from experimental program data bases, stores the data in a single data bank in a common format, and makes the data available to users. The program is designed to be user oriented to minimize the effort required to obtain and manipulate data of interest. The data bank concept and structure embodied in the data bank processing system are applicable to any program where large quantities of scientific (numeric) data are generated and require compiling, storage, and accessing in order to be collected and made available to multiple users. 3 figures

IRIS and the National Research Council (NRC)

Science.gov (United States)

Since the 2011 National Academies’ National Research Council (NRC) review of the IRIS Program's assessment of Formaldehyde, EPA and NRC have had an ongoing relationship into the improvements of developing the IRIS Assessments.

Continuing the Conversation: Development of the U.S. NRC's Definition of Safety Culture and its Traits

International Nuclear Information System (INIS)

Barnes, Valerie; Koves, Ken

2012-01-01

Val Barnes gave a presentation on behalf of the US NRC and INPO. She summarised the work done by the US NRC to develop the US NRC Policy on Safety Culture. Stakeholder representatives were involved in panel sessions to develop a common definition of safety culture and define the traits of a positive safety culture. A survey-based validation study of the eight traits identified through the panel sessions was then conducted across the 63 US nuclear sites by INPO. The INPO study also examined the correlations between the safety culture traits and safety performance. Strong correlations were found for some factors (for example, the number of unplanned scrams correlated strongly with perceptions on management responsibility). The results of the survey supported the inclusion of an additional safety culture trait (questioning attitude) resulting in the following nine traits: - Leadership Safety Values and Actions. - Problem Identification and Resolution. - Personal Accountability. - Work Process. - Continuous Learning. - Environment for Raising Concerns. - Effective Safety Communication. - Respectful Work Environment. - Questioning Attitude. The US NRC has also issued a safety culture policy statement which provides the following definition: 'Nuclear safety culture is the core values and behaviors resulting from a collective commitment by leaders and individuals to emphasize safety over competing goals to ensure protection of people and the environment'. The US NRC and its regulated communities are now working on implementing the policy statement. It was concluded that the work carried out to develop the safety culture policy statement has helped to develop a common language and understanding amongst stakeholders

Nuclear reactor safety research in Idaho

International Nuclear Information System (INIS)

Zeile, H.J.

1983-01-01

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

Prospects for nuclear safety research

Energy Technology Data Exchange (ETDEWEB)

Beckjord, E.S.

1995-04-01

This document is the text of a paper presented by Eric S. Beckjord (Director, Nuclear Regulatory Research/NRC) at the 22nd Water Reactor Safety Meeting in Bethesda, MD in October 1994. The following topics are briefly reviewed: (1) Reactor vessel research, (2) Probabilistic risk assessment, (3) Direct containment heating, (4) Advanced LWR research, (5) Nuclear energy prospects in the US, and (6) Future nuclear safety research. Subtopics within the last category include economics, waste disposal, and health and safety.

Reactor Safety Research: Semiannual report, July-December 1986

Energy Technology Data Exchange (ETDEWEB)

1987-11-01

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

Reactor Safety Research: Semiannual report, July-December 1986

International Nuclear Information System (INIS)

1987-11-01

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

Coupled processes in NRC high-level waste research

International Nuclear Information System (INIS)

Costanzi, F.A.

1987-01-01

The author discusses NRC research effort in support of evaluating license applications for disposal of nuclear waste and for promulgating regulations and issuing guidance documents on nuclear waste management. In order to do this they fund research activities at a number of laboratories, academic institutions, and commercial organizations. One of our research efforts is the coupled processes study. This paper discusses interest in coupled processes and describes the target areas of research efforts over the next few years. The specific research activities relate to the performance objectives of NRC's high-level waste (HLW) regulation and the U.S. Environmental Protection Agency (EPA) HLW standard. The general objective of the research program is to ensure the NRC has a sufficient independent technical base to make sound regulatory decisions

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

USNRC HTGR safety research program overview

International Nuclear Information System (INIS)

Foulds, R.B.

1982-01-01

An overview is given of current activities and planned research efforts of the US Nuclear Regulatory Commission (NRC) HTGR Safety Program. On-going research at Brookhaven National Laboratory, Oak Ridge National Laboratory, Los Alamos National Laboratory, and Pacific Northwest Laboratory are outlined. Tables include: HTGR Safety Issues, Program Tasks, HTGR Computer Code Library, and Milestones for Long Range Research Plan

Disposition of recommendations of the National Research Council in the report ''Revitalizing Nuclear Safety Research''

International Nuclear Information System (INIS)

1988-06-01

On December 8, 1986, the Committee on Nuclear Safety Research of the National Research Council submitted its report, ''Revitalizing Nuclear Safety Research,'' to the US Nuclear Regulatory Commission (NRC). The Commission and its staff have carefully reviewed the Committee's report and have extensively examined the planning, implementation, and management of NRC research programs in order to respond most effectively to the Committee's recommendations. This report presents the Commission's view of the Committee's report and describes the actions that are under way in response to its recommendations

Overview of NRC PRA research program

International Nuclear Information System (INIS)

Cunningham, M.A.; Drouin, M.T.; Ramey-Smith, A.M.; VanderMolen, M.T.

1997-01-01

The NRC's research program in probabilistic risk analysis includes a set of closely-related elements, from basic research to regulatory applications. The elements of this program are as follows: (1) Development and demonstration of methods and advanced models and tools for use by the NRC staff and others performing risk assessments; (2) Support to agency staff on risk analysis and statistics issues; (3) Reviews of risk assessments submitted by licensees in support of regulatory applications, including the IPEs and IPEEEs. Each of these elements is discussed in the paper, providing highlights of work within an element, and, where appropriate, describing important support and feedback mechanisms among elements

1996 NRC annual report. Volume 13

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-01

This 22nd annual report of the US Nuclear Regulatory Commission (NRC) describes accomplishments, activities, and plans made during Fiscal Year 1996 (FH 1996)--October 1, 1995, through September 30, 1996. Significant activities that occurred early in FY 1997 are also described, particularly changes in the Commission and organization of the NRC. The mission of the NRC is to ensure that civilian uses of nuclear materials in the US are carried out with adequate protection of public health and safety, the environment, and national security. These uses include the operation of nuclear power plants and fuel cycle plants and medical, industrial, and research applications. Additionally, the NRC contributes to combating the proliferation of nuclear weapons material worldwide. The NRC licenses and regulates commercial nuclear reactor operations and research reactors and other activities involving the possession and use of nuclear materials and wastes. It also protects nuclear materials used in operation and facilities from theft or sabotage. To accomplish its statutorily mandated regulatory mission, the NRC issues rules and standards, inspects facilities and operations, and issues any required enforcement actions.

1996 NRC annual report. Volume 13

International Nuclear Information System (INIS)

1997-01-01

This 22nd annual report of the US Nuclear Regulatory Commission (NRC) describes accomplishments, activities, and plans made during Fiscal Year 1996 (FH 1996)--October 1, 1995, through September 30, 1996. Significant activities that occurred early in FY 1997 are also described, particularly changes in the Commission and organization of the NRC. The mission of the NRC is to ensure that civilian uses of nuclear materials in the US are carried out with adequate protection of public health and safety, the environment, and national security. These uses include the operation of nuclear power plants and fuel cycle plants and medical, industrial, and research applications. Additionally, the NRC contributes to combating the proliferation of nuclear weapons material worldwide. The NRC licenses and regulates commercial nuclear reactor operations and research reactors and other activities involving the possession and use of nuclear materials and wastes. It also protects nuclear materials used in operation and facilities from theft or sabotage. To accomplish its statutorily mandated regulatory mission, the NRC issues rules and standards, inspects facilities and operations, and issues any required enforcement actions

Overview of NRC's human factors regulatory research program

International Nuclear Information System (INIS)

Coffman, F.D. Jr.

1989-01-01

The human factors research program is divided into distinct and interrelated program activities: (1) Personnel Performance measurement, (2) Personnel Subsystem, (3) Human-System Interface, (4) Organization and Management, and (5) a group of Reliability Assessment activities. The purpose of the Personnel Performance Measurement activity is to improve the Agency's understanding of the factors influencing personnel performance and the effects on the safety of nuclear operations and maintenance by developing improvements to methods for collecting and managing personnel performance data. Personnel Subsystem research will broaden the understanding of such factors as staffing, qualifications, and training that influence human performance in the nuclear system and will develop the technical basis for regulatory guidance to reduce any adverse impact of these influences on nuclear safety. Research in the Human-System Interface activity will provide the technical basis for ensuring that the interface between the system and the human user supports safe operations and maintenance. Organization and Management research will result in the development of tools for evaluating organization and management issues within the nuclear industry. And finally, the Reliability Assessment group of activities includes multidisciplinary research that will integrate human and hardware considerations for evaluating reliability and risk in NRC licensing, inspection, and regulatory decisions

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

The roles of NRC research in risk-informed, performance-based regulation

International Nuclear Information System (INIS)

Morrison, D.L.; Murphy, J.A.; Hodges, M.W.; Cunningham, M.A.; Drouin, M.T.; Ramey-Smith, A.M.; VanderMolen, H.

1997-01-01

The NRC is expanding the use of probabilistic risk analysis (PRA) throughout the spectrum of its regulatory activities. The NRC's research program in PRA supports this expansion in a number of ways, from performing basic research to developing guidance for regulatory applications. The author provides an overview of the NRC's PRA research program, then focuses on two key activities - the review of individual plant examinations, and the development of guidance for use of PRA in reactor regulation

NRC safety research in support of regulation, FY 1992

International Nuclear Information System (INIS)

1993-05-01

This report, the eighth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1992. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that a number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with the technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission's rules, regulatory guides, or other guidance

U.S. NRC training for research and training reactor inspectors

International Nuclear Information System (INIS)

Sandquist, G.M.; Kunze, J.F.

2011-01-01

Currently, a large number of license activities (Early Site Permits, Combined Operating License, reactor certifications, etc.), are pending for review before the United States Nuclear Regulatory Commission (US NRC). Much of the senior staff at the NRC is now committed to these review and licensing actions. To address this additional workload, the NRC has recruited a large number of new Regulatory Staff for dealing with these and other regulatory actions such as the US Fleet of Research and Test Reactors (RTRs). These reactors pose unusual demands on Regulatory Staff since the US Fleet of RTRs, although few (32 Licensed RTRs as of 2010), they represent a broad range of reactor types, operations, and research and training aspects that nuclear reactor power plants (such as the 104 LWRs) do not pose. The US NRC must inspect and regulate all these entities. This paper addresses selected training topics and regulatory activities provided US NRC Inspectors for US RTRs. (author)

Public meeting on radiation safety for industrial radiographerss: remarks, questions and answers at five NRC regional meetings

International Nuclear Information System (INIS)

1978-11-01

Over the past several years thenumber of radiation overexposures experienced in the radiography industry has been higher than for any other single group of NRC licensees. To inform radiography licensees of NRC's concern fo these recurring overexposure incidents, NRC staff representatives met with licensees in a series of five regional meetings. At these meetings the staff presented prepared remarks and answered questions on NRC regulations and operations. The main purposes of the meetings were to express NRC's concern for the high incidence of overexposures, and to open a line of communication between the NRC and radiography licensees in an effort to achieve the common goal of improved radiation safety. The remarks presented by the staff and subjects discussed at these meetings included: the purpose, scope, findings and goals of the NRC inspection program; ways and means of incorporating safety into radiography operations; and case histories of overexposure incidents, with highlights of the causes and possible preventions. At each of the regional meetings the staff received a request for a copy of the prepared remarks and a consolidation of the questions and answers that were discussed. This document includes that information, and a copy is being provided to each organizaion or firm attending the regional meetings. Requests for other copies should be made in accordance with the directions printed inside the front cover of this document

75 FR 20009 - Development of NRC's Safety Culture Policy Statement: Cancellation of Public Workshops Scheduled...

Science.gov (United States)

2010-04-16

... included the development of the safety culture common terminology effort comprised of: (1) Development of a... alignment on the safety culture common terminology effort. The NRC held the first workshop on February 2-4... of individual industry groups, such as, [[Page 20010

Operational safety reliability research

International Nuclear Information System (INIS)

Hall, R.E.; Boccio, J.L.

1986-01-01

Operating reactor events such as the TMI accident and the Salem automatic-trip failures raised the concern that during a plant's operating lifetime the reliability of systems could degrade from the design level that was considered in the licensing process. To address this concern, NRC is sponsoring the Operational Safety Reliability Research project. The objectives of this project are to identify the essential tasks of a reliability program and to evaluate the effectiveness and attributes of such a reliability program applicable to maintaining an acceptable level of safety during the operating lifetime at the plant

NRC/AMRMC Resident Research Associateship Program

Science.gov (United States)

2015-05-01

maintaining a presence on social media sites such as Facebook . The NRC attended a number of minority focused events in which we maintained exhibit booths...PROPOSAL The Use of Glial Inhibitors to Increase the Efficacy of Opioid Analgesics while Eliminating the Propensity for Addiction 7) SUMMARY OF RESEARCH

Annual technical meeting of the NRC cooperative severe accident research program

International Nuclear Information System (INIS)

Silver, E.G.

1993-01-01

This brief report summarizes the 1992 annual technical meeting of the NRC Cooperative Severe Accident Research Program (CSARP-92) held at the Hyatt Regency Hotel in Bethesda, Maryland, May 4-8, 1992. The report is taken mainly from coverage of the meeting published in the June 5, 1992, issue of Atomic Energy Clearinghouse. Results of this meeting are formalized at the Water Reactor Safety Information Meetings (WRSIM) that are held annually in October. Nuclear Safety summarizes the annual WRSIM meetings and provides a list of the presentations that were given. Interested readers are encouraged to review listed topics to identify specific topic areas in severe accident research. Sessions were held on in-vessel core melt progression; fuel-coolant interactions; fission-product behavior; direct containment heating; and severe accident code development, assessment, and validation. Summaries of the individual technical sessions and the current state of the art in these areas were given by the chairmen

NRC inventory of dams

International Nuclear Information System (INIS)

Lear, G.E.; Thompson, O.O.

1983-01-01

The NRC Inventory of Dams has been prepared as required by the charter of the NRC Dam Safety Officer. The inventory lists 51 dams associated with nuclear power plant sites and 14 uranium mill tailings dams (licensed by NRC) in the US as of February 1, 1982. Of the 85 listed nuclear power plants (148 units), 26 plants obtain cooling water from impoundments formed by dams. The 51 dams associated with the plants are: located on a plant site (29 dams at 15 plant sites); located off site but provide plant cooling water (18 dams at 11 additional plant sites); and located upstream from a plant (4 dams) - they have been identified as dams whose failure, and ensuing plant flooding, could result in a radiological risk to the public health and safety. The dams that might be considered NRC's responsibility in terms of the federal dam safety program are identified. This group of dams (20 on nuclear power plant sites and 14 uranium mill tailings dams) was obtained by eliminating dams that do not pose a flooding hazard (e.g., submerged dams) and dams that are regulated by another federal agency. The report includes the principal design features of all dams and related useful information

NRC's geotechnical engineering research needs for the high-level waste repository program

International Nuclear Information System (INIS)

Gupta, D.C.; Philip, J.; Lorig, L.J.; Chowdhury, A.H.

1992-01-01

To develop the capability for independently assessing the US Department of Energy's (DOE's) geologic repository design within a limited time, the US Nuclear Regulatory Commission (NRC) staff needs to perform certain research well before receiving the license application. The NRC staff is using a number of factors to identify the areas that it needs to research. The staff assigns priorities to the needed research based on programmatic considerations and the significance of the work. In the geotechnical engineering field, the staff is conducting research in the following three areas: response of the repository to repeated strong ground motion, rock-mass sealing, and coupled thermo-hydro-mechanical interactions. In this paper, the NRC staff also presents the areas of additional research needed in the geotechnical engineering field

Stakeholder confidence in effective safety regulation. A regulator's view on the role of independent research capability

International Nuclear Information System (INIS)

Kotra, Janet; Mohanty, Sitakanto

2006-01-01

The authors provided a regulator's view on the role of independent research capability and its relationship to stakeholder confidence. They underscored the NRC's commitment to regulatory openness in its Strategic Plan. A number of actions have been adopted to achieve openness: public access to information about risks, safety and licensee performance that is accurate and timely; enhanced awareness of NRC as an independent regulator; fair and timely process for public involvement in NRC's decision-making; and early public involvement and two-way communication to enhance public confidence in NRC's regulatory process. The presentation reviewed the ways through which the NRC seeks to inspire confidence in its process as independent regulator and in future decisions concerning whether to authorize the U.S. DOE to construct the proposed repository at Yucca Mountain. Key to this will be reviewing all information objectively, making open decisions based on sound, scientific judgements about the facts; and maintaining an open and fair public process, and accessing independent R and D. It was noted that among the requirements for the safety review for the Yucca Mountain repository will be extensive technical and scientific analyses, evaluation of expert judgement, and long-range modelling assessments of expected repository performance. NRC requires independent scientific and engineering analyses to develop technical bases for regulations and guidance; evaluate adequacy of DOE's safety case for a potential repository; assist preparation of NRC Safety Evaluation Report; provide technical support for NRC testimony during licensing hearing; and develop effective outreach and communication tools. Confidence in NRC's use of science and engineering will depend upon: competence; independence; open and fair process; regulatory outcomes that are subject to verification and monitoring. To assist, the NRC established the Centre for Nuclear Waste Regulatory

NRC/AMRMC Resident Research Associateship Program

Science.gov (United States)

2018-05-01

2- 0010 Report Period: 02/06/2012-02/28/2018 4/11/2018, 12:17 PM During the reporting period, the National Academies of Sciences, Engineering , and...to advertise the NRC Research Associateship Programs included the following: 1) attendance at meetings of major scientific and engineering ...professional societies; 2) advertising in programs and career centers for these and other professional society meetings; 3) direct mailing and emailing of

Review and evaluation of the Nuclear Regulatory Commission Safety Research Program for Fiscal Years 1984 and 1985

International Nuclear Information System (INIS)

1983-02-01

This is the sixth report by the Advisory Committee on Reactor Safeguards (ACRS) that has been prepared in response to the Congressional requirement for an annual report on the Nuclear Regulatory Commission (NRC) Reactor Safety Research Program. Part I is a compilation of our general comments and recommendations regarding the NRC Safety Research Program, and includes budget recommendations and an identification of matters of special importance that deserve increased emphasis. It is intended to serve as an Executive Summary. Part II is divided into ten chapters, each of which represents a Decision Unit of the NRC research program. In each chapter, specific comments are included on the research involved in the Decision Unit, an assessment of priorities, and recommendations regarding new directions and levels of funding

NRC safety research in support of regulation. Volume 8, FY 1993

Energy Technology Data Exchange (ETDEWEB)

1994-06-01

This report, the ninth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1993. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with sound technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission`s rules, regulatory guides, or other guidance.

NRC comprehensive records disposition schedule

International Nuclear Information System (INIS)

1983-05-01

Effective January 1, 1982, NRC will institute records retention and disposal practives in accordance with the approved Comprehensive Records Disposition Schedule (CRDS). CRDS is comprised of NRC Schedules (NRCS) 1 to 4 which apply to the agency's program or substantive records and General Records Schedules (GRS) 1 to 24 which apply to housekeeping or facilitative records. NRCS-I applies to records common to all or most NRC offices; NRCS-II applies to program records as found in the various offices of the Commission, Atomic Safety and Licensing Board Panel, and the Atomic Safety and Licensing Appeal Panel; NRCS-III applies to records accumulated by the Advisory Committee on Reactor Safeguards; and NRCS-IV applies to records accumulated in the various NRC offices under the Executive Director for Operations. The schedules are assembled functionally/organizationally to facilitate their use. Preceding the records descriptions and disposition instructions for both NRCS and GRS, there are brief statements on the organizational units which accumulate the records in each functional area, and other information regarding the schedules' applicability

REVIEW OF NRC APPROVED DIGITAL CONTROL SYSTEMS ANALYSIS

International Nuclear Information System (INIS)

Markman, D.W.

1999-01-01

Preliminary design concepts for the proposed Subsurface Repository at Yucca Mountain indicate extensive reliance on modern, computer-based, digital control technologies. The purpose of this analysis is to investigate the degree to which the U. S. Nuclear Regulatory Commission (NRC) has accepted and approved the use of digital control technology for safety-related applications within the nuclear power industry. This analysis reviews cases of existing digitally-based control systems that have been approved by the NRC. These cases can serve as precedence for using similar types of digitally-based control technologies within the Subsurface Repository. While it is anticipated that the Yucca Mountain Project (YMP) will not contain control systems as complex as those required for a nuclear power plant, the review of these existing NRC approved applications will provide the YMP with valuable insight into the NRCs review process and design expectations for safety-related digital control systems. According to the YMP Compliance Program Guidance, portions of various NUREGS, Regulatory Guidelines, and nuclear IEEE standards the nuclear power plant safety related concept would be applied to some of the designs on a case-by-case basis. This analysis will consider key design methods, capabilities, successes, and important limitations or problems of selected control systems that have been approved for use in the Nuclear Power industry. An additional purpose of this analysis is to provide background information in support of further development of design criteria for the YMP. The scope and primary objectives of this analysis are to: (1) Identify and research the extent and precedence of digital control and remotely operated systems approved by the NRC for the nuclear power industry. Help provide a basis for using and relying on digital technologies for nuclear related safety critical applications. (2) Identify the basic control architecture and methods of key digital control

Survey by senior NRC management to obtain veiwpoints on the safety impact of regulatory activities from representative utilities operating and constructing nuclear power plants

International Nuclear Information System (INIS)

1981-08-01

A survey of licensee staff members representing the several organizational elements in different licensee corporate and plant staffs was conducted by senior NRC management to obtain licensee views on the potential safety consequences and impact of NRC regulatory activities. The comments received addressed the full scope of NRC activities and the negative impact of agency actions on licensee resources, staff performance, planning and scheduling, and organizational effectiveness. The findings of the survey is that the pace and nature of regulatory actions have created a potential safety problem which deserves further evaluation by the agency

Development of Safety Review Guidance for Research and Training Reactors

International Nuclear Information System (INIS)

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

2007-01-01

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

Risk-informed inservice test activities at the NRC

International Nuclear Information System (INIS)

Fischer, D.; Cheok, M.; Hsia, A.

1996-01-01

The operational readiness of certain safety-related components is vital to the safe operation of nuclear power plants. Inservice testing (IST) is one of the mechanisms used by licensees to ensure this readiness. In the past, the type and frequency of IST have been based on the collective best judgment of the NRC and industry in an ASME Code consensus process and NRC rulemaking process. Furthermore, IST requirements have not explicitly considered unique component and system designs and contribution to overall plant risk. Because of the general nature of ASME Code test requirements and non-reliance on risk estimates, current IST requirements may not adequately emphasize testing those components that are most important to safety and may overly emphasize testing of less safety significant components. Nuclear power plant licensees are currently interested in optimizing testing by applying resources in more safety significant areas and, where appropriate, reducing measures in less safety-significant areas. They are interested in maintaining system availability and reducing overall maintenance costs in ways that do not adversely affect safety. The NRC has been interested in using probabilistic, as an adjunct to deterministic, techniques to help define the scope, type and frequency of IST. The development of risk-informed IST programs has the potential to optimize the use of NRC and industry resources without adverse affect on safety

Risk-informed inservice test activities at the NRC

Energy Technology Data Exchange (ETDEWEB)

Fischer, D.; Cheok, M.; Hsia, A.

1996-12-01

The operational readiness of certain safety-related components is vital to the safe operation of nuclear power plants. Inservice testing (IST) is one of the mechanisms used by licensees to ensure this readiness. In the past, the type and frequency of IST have been based on the collective best judgment of the NRC and industry in an ASME Code consensus process and NRC rulemaking process. Furthermore, IST requirements have not explicitly considered unique component and system designs and contribution to overall plant risk. Because of the general nature of ASME Code test requirements and non-reliance on risk estimates, current IST requirements may not adequately emphasize testing those components that are most important to safety and may overly emphasize testing of less safety significant components. Nuclear power plant licensees are currently interested in optimizing testing by applying resources in more safety significant areas and, where appropriate, reducing measures in less safety-significant areas. They are interested in maintaining system availability and reducing overall maintenance costs in ways that do not adversely affect safety. The NRC has been interested in using probabilistic, as an adjunct to deterministic, techniques to help define the scope, type and frequency of IST. The development of risk-informed IST programs has the potential to optimize the use of NRC and industry resources without adverse affect on safety.

Spent nuclear fuel project multi-canister overpack, additional NRC requirements

International Nuclear Information System (INIS)

Garvin, L.J.

1998-01-01

The US Department of Energy (DOE), established in the K Basin Spent Nuclear Fuel Project Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel (SNF) Project facilities to achieve nuclear safety equivalency to comparable US Nuclear Regulatory Commission (NRC)-licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Cold Vacuum Drying (CVD) facility or Hot Conditioning System, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNF Project facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements needed, in combination with the existing and applicable DOE requirements, to establish nuclear safety equivalency for the MCO. The background, basic safety issues and general comparison of NRC and DOE requirements for the SNF Project are presented in WHC-SD-SNF-DB-002

CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

International Nuclear Information System (INIS)

Charak, I.; Kier, P.H.

1995-04-01

Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule (section 50.62); (2) station blackout (section 50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term (section 50.34(f) and section 100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements (section 50.55a, etc); (6) ECCS acceptance criteria (section 50.46)(b); (7) combustible gas control (section 50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle (section 51.51); and (11) (standards section 50.55a)

Current perspectives on performance assessment at the NRC

International Nuclear Information System (INIS)

Coplan, S.M.; Eisenberg, N.A.; Federline, M.V.; Randall, J.D.

1992-01-01

The Nuclear Regulatory Commission (NRC) staff is engaging in a number of activities involving performance assessment in order to support NRC's program in high-level waste management. Broad areas of activity include: (1) reactive work responding to products and activities of the Department of Energy (DOE), (2) proactive work, including development of an independent performance assessment capability, development of guidance for DOE, support for technical and programmatic integration, (3) a program of regulatory research, and (4) participation in a number of international activities. As the U.S. high-level waste program continues to mature, performance assessment is seen as playing a more prominent role in evaluating safety and focussing technical activities

NRC regulation of DOE facilities

International Nuclear Information System (INIS)

Buhl, A.R.; Edgar, G.; Silverman, D.; Murley, T.

1997-01-01

The US Department of Energy (DOE), its contractors, and the Nuclear Regulatory Commission (NRC) are in for major changes if the DOE follows through on its intentions announced December 20, 1996. The DOE is seeking legislation to establish the NRC as the regulatory agency with jurisdiction over nuclear health, safety, and security at a wide range of DOE facilities. At this stage, it appears that as many as 200 (though not all) DOE facilities would be affected. On March 28, 1997, the NRC officially endorsed taking over the responsibility for regulatory oversight of DOE nuclear facilities as the DOE had proposed, contingent upon adequate funding, staffing resources, and a clear delineation of NRC authority. This article first contrasts the ways in which the NRC and the DOE carry out their basic regulatory functions. Next, it describes the NRC's current authority over DOE facilities and the status of the DOE's initiative to expand that authority. Then, it discusses the basic changes and impacts that can be expected in the regulation of DOE facilities. The article next describes key lessons learned from the recent transition of the GDPs from DOE oversight to NRC regulation and the major regulatory issues that arose in that transition. Finally, some general strategies are suggested for resolving issues likely to arise as the NRC assumes regulatory authority over DOE facilities

NRC nuclear waste geochemistry 1983

International Nuclear Information System (INIS)

Alexander, D.H.; Birchard, G.F.

1984-05-01

The purpose of the meeting was to present results from NRC-sponsored research and to identify regulatory research issues which need to be addressed prior to licensing a high-level waste repository. Important summaries of technical issues and recommendations are included with each paper. The issue reflect areas of technical uncertainty addressed by the NRC Research program in geochemistry. The objectives of the NRC Research Program in geochemistry are to provide a technical basis for waste management rulemaking, to provide the NRC Waste Management Licensing Office with information that can be used to support sound licensing decisions, and to identify investigations that need to be conducted by DOE to support a license application. Individual papers were processed for inclusion in the Energy Data Base

Summary of ORNL work on NRC-sponsored HTGR safety research, July 1974-September 1980

International Nuclear Information System (INIS)

Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Delene, J.G.; Harrington, R.M.; Hatta, M.; Hedrick, R.A.; Johnson, L.G.; Sanders, J.P.

1982-03-01

A summary is presented of the major accomplishments of the Oak Ridge National Laboratory (ORNL) research program on High-Temperature Gas-Cooled Reactor (HTGR) safety. This report is intended to help the nuclear Regulatory Commission establish goals for future research by comparing the status of the work here (as well as at other laboratories) with the perceived safety needs of the large HTGR. The ORNL program includes extensive work on dynamics-related safety code development, use of codes for studying postulated accident sequences, and use of experimental data for code verification. Cooperative efforts with other programs are also described. Suggestions for near-term and long-term research are presented

NRC research on the application of advanced I and C technology to commercial nuclear power plants

International Nuclear Information System (INIS)

Gollei, K.R.; Hon, A.L.

1983-01-01

The operational safety and efficiency of commercial nuclear power plants (NPP's) could possibly be enhanced by utilizing advanced instrumentation and control technology developed by other industries. The NRC is interested in learning about new I and C technology that probably will or could be applied to new or existing plants. This would enable the NRC to be better prepared to evaluate the application without undue delays. It would also help identify any appropriate changes in NRC regulations or guidance necessary to facilitate the application of advanced IandC technology to NPP's. The NRC has initiated a project to work cooperatively with the advanced technology industry, power industry, EPRI, and technical organizations such as ISA toward this goal. This paper describes the objectives and plans of this cooperative effort. It summarizes the highlights of some of the advanced technology already being evaluated by NRC such as microprocessor applications, instruments to detect inadequate core cooling and other two-phase flow measurements, reactor noise surveillance and diagnostic techniques. This paper also suggests potential candidates for consideration such as utilization of advanced instruments for LOCA experiments. It also identifies some of the potential challenges facing the application of advanced technology to NPP's. It concludes that close cooperation between NRC and industry is essential for the success of such applications

In vivo monitoring of nuclear research centre (NRC) workers. Vol. 4

Energy Technology Data Exchange (ETDEWEB)

Gomaa, M A; Ali, E M; Taha, T M [Radiation Protection Departion, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

Occupational workers of Nuclear research center (NRC) of atomic energy authority (AEA) working with radio-nuclides such as (Cs-137, I-131, and I-125) are monitored yearly. Individuals involved in routine work in hot laboratories are examined every three months. When an incident occurs, the workers involved are examined promptly. In order to measure internal contamination, the modified NRC-AEA whole body counter is used. More than hundred occupational workers had been examined in 1994. Results indicated that the annual limit of intake as recommended by ICRP was not exceeded. 3 figs., 3 tabs.

NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Presentation on NRC Regulatory Positions and guidelines

International Nuclear Information System (INIS)

Russell, W.T.

1994-01-01

The NRC staff recognizes the potential for enhanced safety and reliability that digital systems bring to the nuclear industry. The staff also recognizes the challenges to safety that are unique to digital systems implementation

Introductory user's manual for the US Nuclear Regulatory Commission Reactor Safety Research Data Bank

International Nuclear Information System (INIS)

Scofield, N.R.; Hardy, H.A.; Laats, E.T.

1984-03-01

The United States NRC has established the NRC/Division of Accident Evaluation (DAE) Data Bank Program to collect, store, and make available data from the many domestic and foreign water reactor safety research programs. Local direction of the program is provided by EG and G Idaho, Inc., prime contractor for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). The NRC/DAE Data Bank Program provides a central computer storage mechanism and access software for data that is to be used by code development and assessment groups in meeting the code and correlation needs of the nuclear industry. The administrative portion of the program provides data entry, documentation, training, and advisory services to users and the NRC. The NRC/DAE Data Bank and the capabilities of the data access software are described in this document

Introductory user's manual for the US Nuclear Regulatory Commission Reactor Safety Research Data Bank

International Nuclear Information System (INIS)

Scofield, N.R.; Hardy, H.A.; Laats, E.T.

1983-02-01

The United States Nuclear Regulatory Commission (NRC) has established the NRC/Division of Accident Evaluation (DAE) Data Bank Program to collect, store, and make available data from the many domestic and foreign water safety research programs. Local direction of the program is provided by E G and G Idaho, Inc., prime contractor for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). The NRC/DAE Data Bank Program provides a central computer storage mechanism and access software for data that is to be used by code development and assessment groups in meeting the code and correlation needs of the nuclear industry. The administrative portion of the program provides data entry, documentation, training, and advisory services to users and the NRC. The NRC/DAE Data Bank and the capabilities of the data access software are described

Overview of the NRC performance monitoring program

International Nuclear Information System (INIS)

Jordan, E.L.

1987-01-01

In response to the accident at Three Mile Island, the NRC developed the Systematic Assessment of Licensee Performance (SALP) Program to aid in the identification of those licensees that were more likely than others to have safety problems and to provide a rational basis for allocation of inspection resources. The NRC also has an ongoing program of screening and evaluating operating reactor event reports on a daily basis for promptly identifying safety problems. Although the SALP and event report evaluation programs have been successful in identifying potential performance problems, a concern developed recently about the adequacy and timeliness of NRC programs to detect poor or declining performance. The performance indicator program as approved by the commission is in the implementation phase. The program is expected to undergo refinements as new indicators are developed and experience is gained in the use of indicators

Reports distributed under the NRC Light-Water Reactor Safety Research Foreign Technical Exchange Program. Volume III, January--June 1977

International Nuclear Information System (INIS)

Sharp, D.S.; Cottrell, W.B.

1977-01-01

Lists of documents exchanged during the first half of 1977 under agreements between the U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research and the governments of France, Federal Republic of Germany, and Japan are presented. During this period, the NRC received 41 reports from France, 29 from F. R. Germany, and 24 from Japan, and in return sent 107 U.S. reports to each of these three countries

Investigation and consideration on the framework of oversight-based safety regulation. U.S. NRC 'Risk-Informed, Performance-Based' Regulation

International Nuclear Information System (INIS)

Saji, Gen

2001-01-01

Regulation on safety, environment and health in Japan has before today been intended to correspond with an accident at forms of reinforcement of national standards and monitoring, if any. However, as it was thought that such regulation reinforcement was afraid to bring some social rigidity, and to weaken independent responsibility, as a result, because of anxiety of losing peoples' merits inversely, some fundamental directivity such as respect of self-responsibility principle' and 'necessary and least limit of regulation' were selected as a part of political innovation. On the other hand, at a background of wide improvements on various indexing values showing operation results of nuclear power stations in U.S.A., private independent effort on upgrading of safety is told to largely affect at beginning of INPO (Institute of Nuclear Power Operations), without regulation reinforcement of NRC side. This is a proof of concrete effect of transfer to oversight-based safety regulation. Here were introduced on nuclear safety in U.S.A. at a base of some references obtained on entering the 'MIT summer specialist program. Nuclear system safety', on focussing at new safety regulation of NRC and its effect and so on, and adding some considerations based on some knowledge thereafter. (G.K.)

Amalgamation of performance indicators to support NRC senior management reviews

International Nuclear Information System (INIS)

Wreathall, J.; Schurman, D.; Modarres, M.; Mosleh, A.; Anderson, N.; Reason, J.

1991-01-01

The purpose of this project is to develop a methodology for amalgamating performance indicators to provide an overall perspective on plant safety, as one input to Nuclear Regulatory Commission's (NRC) senior management reviews of plant safety. These reviews are used to adjust the level of oversight by NRC. Work completed to date includes the development of frameworks for relating indicator measures to safety, a classification scheme for performance indicators, and a mapping process to portray indicators in the frameworks

NRC regulatory information conference: Proceedings

International Nuclear Information System (INIS)

1989-09-01

This volume of the report provides the proceedings from the Nuclear Regulatory Commission (NRC) Regulatory Information Conference that was held at the Mayflower Hotel, Washington, DC, on April 18, 19, and 20, 1989. This conference was held by the NRC and chaired by Dr. Thomas E. Mosley, Director, Office of Nuclear Reactor Regulations (NRR) and coordinated by S. Singh Bajwa, Chief, Technical Assistance Management Section, NRR. There were approximately 550 participants from nine countries at the conference. The countries represented were Canada, England, Italy, Japan, Mexico, Spain, Taiwan, Yugoslavia, and the United States. The NRC staff discussed with nuclear industry its regulatory philosophy and approach and the bases on which they have been established. Furthermore, the NRC staff discussed several initiatives that have been implemented recently and their bases as well as NRC's expectations for new initiatives to further improve safety. The figures contained in Appendix A to the volume correspond to the slides that were shown during the presentations. Volume 2 of this report contains the formal papers that were distributed at the beginning of the Regulatory Information Conference and other information about the conference

Reports distributed under the NRC Light-Water Reactor Safety Research Foreign Technical Exchange Program. Volume III, January--June 1977

Energy Technology Data Exchange (ETDEWEB)

Sharp, D.S.; Cottrell, W.B.

1977-09-19

Lists of documents exchanged during the first half of 1977 under agreements between the U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research and the governments of France, Federal Republic of Germany, and Japan are presented. During this period, the NRC received 41 reports from France, 29 from F. R. Germany, and 24 from Japan, and in return sent 107 U.S. reports to each of these three countries.

Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-04-01

This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee`s annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs.

Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

International Nuclear Information System (INIS)

1997-04-01

This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee's annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs

Safeguards at NRC licensed facilities: Are we doing enough

International Nuclear Information System (INIS)

Asselstine, J.K.

1986-01-01

Safeguards at the Nuclear Regulatory Commission (NRC) facilities are discussed in this paper. The NRC is pursuing a number of initiatives in the safeguards area. The Commission is conducting a reassessment of its safeguards design basis threat statements to consider the possible implications of an explosive-laden vehicle for U.S. nuclear safeguards and to examine the comparability of safeguards features at NRC-licensed and DOE facilities. The Commission is also completing action on measures to protect against the sabotage threat from an insider at NRC-licensed facilities, and is examining the potential safety implications of safeguards measures. Finally, the NRC has developed measures to reduce the theft potential for high-enriched uranium

US NRC research on the integrity of piping in nuclear reactor primary systems

International Nuclear Information System (INIS)

Serpan, C.Z. Jr.

1983-01-01

This paper has attempted to provide a ''snapshot'' of the activities underway in NRC on the subject of LWR piping integrity as of the summer and fall of 1983. The paper is necessarily vague on certain topics of policy because they are either under review or are under development and the outcome cannot be accurately forecast at this time. Particularly in the area of BWR pipe cracking, events are very rapid so that positions and actions described in this paper may well be obsolete by the time it is published. Nevertheless, the activities and positions are as accurate as possible at the time of writing. Certainly the longer-range aspects of the research program represent the current direction and intent of NRC; nevertheless, as results come in and actions occur in the licensing and regulation arena of operating reactors, the emphasis of the research programs will necessarily shift to accommodate them so as to remain as relevant as possible. Thus, this paper is useful to show the intentions of NRC in the area of research for LWR piping, and it is also useful to document the status of the regulations on piping for which the research is being performed. (orig.)

Reports distributed under the NRC Light-Water Reactor Safety Research Foreign Technical Exchange Program. Volume IV, July--December 1977

International Nuclear Information System (INIS)

Sharp, D.S.; Cottrell, W.B.

1978-01-01

Lists of documents exchanged during the second half of 1977 under agreements between the U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research and the governments of France, Federal Republic of Germany, Japan, and the United Kingdom are presented. During this period, the NRC received 1 report from France, 40 from the Federal Republic of Germany, and 11 from Japan, and in return sent 112 U.S. reports to each of these three countries and 23 reports to the United Kingdom

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

Preparation for Scaling Studies of Ice-Crystal Icing at the NRC Research Altitude Test Facility

Science.gov (United States)

Struk, Peter M.; Bencic, Timothy J.; Tsao, Jen-Ching; Fuleki, Dan; Knezevici, Daniel C.

2013-01-01

This paper describes experiments conducted at the National Research Council (NRC) of Canadas Research Altitiude Test Facility between March 26 and April 11, 2012. The tests, conducted collaboratively between NASA and NRC, focus on three key aspects in preparation for later scaling work to be conducted with a NACA 0012 airfoil model in the NRC Cascade rig: (1) cloud characterization, (2) scaling model development, and (3) ice-shape profile measurements. Regarding cloud characterization, the experiments focus on particle spectra measurements using two shadowgraphy methods, cloud uniformity via particle scattering from a laser sheet, and characterization of the SEA Multi-Element probe. Overviews of each aspect as well as detailed information on the diagnostic method are presented. Select results from the measurements and interpretation are presented which will help guide future work.

NRC performance indicator program

International Nuclear Information System (INIS)

Singh, R.N.

1987-01-01

The performance indicator development work of the US Nuclear Regulatory Commission (NRC) interoffice task group involved several major activities that included selection of candidate indicators for a trial program, data collection and review, validation of the trial indicators, display method development, interactions with the industry, and selection of an optimum set of indicators for the program. After evaluating 27 potential indicators against certain ideal attributes, the task group selected 17 for the trial program. The pertinent data for these indicators were then collected from 50 plants at 30 sites. The validation of the indicators consisted of two primary processes: logical validity and statistical analysis. The six indicators currently in the program are scrams, safety system actuations, significant events, safety system failures, forced outage rate, and equipment forced outages per 100 critical hours. A report containing data on the six performance indicators and some supplemental information is issued on a quarterly basis. The NRC staff is also working on refinements of existing indicators and development of additional indicators as directed by the commission

The use of U.S. NRC licensing practices for VVERs

International Nuclear Information System (INIS)

Popp, D.M.

2000-01-01

The licensing process for the upgraded Temelin I and C and Fuel designs were enhanced with the introduction of U.S. Nuclear Regulatory Commission, NRC practices. Specifically, the use of the NRC Regulatory Guide 1.70, 'Standard Format and Content Guide for Safety Analyses Reports' and NRC NUREG 0800, 'Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants', were beneficial in the development and review of Temelin licensing documentation. These standards have been used for the preparation and review of Safety Analysis Reports in the United States and also in a large number of licensing applications around the world. Both Regulatory Guide 1.70 and NUREG 0800 were developed to provide a predictable and structured approach to licensing. This paper discusses this approach and identifies the benefits to designers, writers of licensing documentation and reviewers of licensing documents. (author)

Plan for research to improve the safety of light-water nuclear power plants

International Nuclear Information System (INIS)

1978-03-01

This is the U.S. Nuclear Regulatory Commission's first annual report to Congress on recommendations for research on improving the safety of light-water nuclear power plants. Suggestions for reactor safety research were identified in, or received from, various sources, including the Advisory Committee on Reactor Safeguards, the NRC regulatory staff, and the consultants to the Research Review Group. After an initial screening to eliminate those not related to improved reactor safety, all the suggestions were consolidated into research topics. It is recommended that the following research projects be carried out: alternate containment concepts, especially vented containments; alternate decay heat removal concepts, especially add-on bunkered systems; alternate emergency core cooling concepts; improved in-plant accident response; and advanced seismic designs

NRC regulatory uses of PSA

International Nuclear Information System (INIS)

Murley, T.E.

1991-01-01

The publication in 1975 of WASH-1400, with its new probabilistic safety assessment (PSA) methodology, had the effect of presenting a pair of eyeglasses to a man with poor eyesight. Suddenly, it gave us a view of nuclear safety with a new clarity, and it allowed us to sort out the important safety issues from the unimportant. In the intervening years, PSA insights have permeated the fabric of nearly all our safety judgments. This acceptance can be seen from the following list of broad areas where the Nuclear Regulatory Commission (NRC) staff uses PSA insights and methodology: evaluating the safety significance of operating events and recommending safety improvements where warranted; requesting licensees to systematically look for design vulnerabilities in each operating reactor; evaluating the safety significance of design weaknesses or non-compliances when judging the time frame for necessary improvements; conducting sensitivity analyses to judge where safety improvements are most effective; assessing the relative safety benefits of design features for future reactors. In judging where PSA methodology can be improved to give better safety insights, it is believed that the following areas need more attention: better modeling of cognitive errors; more comprehensive modeling of accident sequences initiated from conditions other than full power; more comprehensive modeling of inter-system loss of coolant accident (ISLOCA) sequences. Although PSA is widely used in the staff's regulatory activities, the NRC deliberately chooses not to include probabilistic prescriptions in regulations or guidance documents. The staff finds the bottom line risk estimates to be one of the least reliable products of a PSA. The reason for this view is that PSA cannot adequately address cognitive errors nor assess the effects of a pervasive poor safety attitude

Interactive Computerized Based Training, In Radiation Protection at NRC-Negev

International Nuclear Information System (INIS)

Sberlo, E.; Krumbein, H.; Ankri, D.; Ben-Shachar, B.; Laichter, Y.; Weizer, G.; Adorarn, D.

1999-01-01

According to the rules of safety at the working places in Israel, all radiation employees in Israel should receive once a year a refreshing course in several areas of safety. At the NRC-Negev there are two kinds of radiation employees: the ''hot area'' employees, who work in an environment of radioactive materials or radiation machines and the ''old area'' employees (all the other employees in the NRC-Negev). One of the main goals of the Department of Human Resources Development and Training at the NRC-Negev was to organize safety refresher courses. All ''hot area'' employees received a training program of two days in safety subjects, each year. The ''cold area'' employees received the same course, each second year. The former training program included several lectures in radiation protection, health physics, biological effects of ionizing radiation, etc., as well as same lectures in industrial safety, fast aid, fee fighting, emergency procedures, etc. The safety refresher courses were given by Rental lectures. There were a lot of disadvantages in these frontal lectures: The lecturers are employees of the NRCN who had to stop their routine work in order to lecture; the lecturers had to carry out identical training for each course for a large group of workers; there was a lack of testing methods or any other certification for the employees. Recently, seven safety courseware were developed by the NRC-Negev and the CET (Centre for Educational Technology), in order to perform these safety refresher courses. The courseware are based on an interactive computerized training including tutorials and quiz. The tutorial is an interactive course in each subject. The employee gets a simple and clear explanation (including pictures). After each Morial there is a quiz which includes 7 American style questions. The first two courseware are for all the employees, the next 4 courseware for the ''hot area'' employees, and the seventh for the ''cold area'' employees (the seventh is a

Examples of NRC research products used in regulation

International Nuclear Information System (INIS)

Anderson, N.R.

1987-01-01

The key to effective research is a close relationship between information needs and research results. This can only be achieved by close cooperation between the researchers and the regulators. At the NRC, this relationship has matured over the years until now the researchers participate in definition of the information needs and the regulators help define the research programs. The more formal means of ensuring a close match between needs and results include joint research groups, oversight working groups, and a system of Research Information Letters (RILs). On an informal basis there are many day to day discussions and meetings on the various programs which ensure effective program guidance and early identification of significant findings. This paper describes both the formal and informal researcher/regulation interface and discusses some examples of how specific research programs are utilized in the regulatory process. Specific programs described are the pressurized thermal shock program, the seismic margins program and the Category 1 structures program. Other examples cited are the aging and life extension programs

Research on the management and endorsement of nuclear safety standards in the United States and its revelation for China

Science.gov (United States)

Liu, Ting; Tian, Yu; Yang, Lili; Gao, Siyi; Song, Dahu

2018-01-01

This paper introduces the American standard system, the Nuclear Regulatory Commission (NRC)’s responsibility, NRC nuclear safety regulations and standards system, studies on NRC’s standards management and endorsement mode, analyzes the characteristics of NRC standards endorsement management, and points out its disadvantages. This paper draws revelation from the standard management and endorsement model of NRC and points suggestion to China’s nuclear and radiation safety standards management.The issue of the “Nuclear Safety Law”plays an important role in China’s nuclear and radiation safety supervision. Nuclear and radiation safety regulations and standards are strong grips on the implementation of “Nuclear Safety Law”. This paper refers on the experience of international advanced countriy, will effectively promote the improvement of the endorsed management of China’s nuclear and radiation safety standards.

Review and evaluation of the Nuclear Regulatory Commission safety research program for Fiscal Year 1983. Report to the Congress

International Nuclear Information System (INIS)

1982-02-01

Public Law 95-209 includes a requirement that the Advisory Committee on Reactor Safeguards submit an annual report to Congress on the safety research program of the Nuclear Regulatory Commission. This report presents the results of the ACRS review and evaluation of the NRC safety research program for Fiscal Year 1983. The report contains a number of comments and recommendations

NRC Bulletin No. 87-02, Supplement 1: Fastener testing to determine conformance with applicable material specifications

International Nuclear Information System (INIS)

Rossi, C.E.

1992-01-01

Item 5 of NRC Compliance Bulletin 87-02 requested that all holders of operating licenses or construction permits for nuclear power reactors information regarding the identity of the suppliers and manufacturers of the safety-related and non-safety-related fasteners selected for testing. After further consideration, the NRC has determined that it needs information regarding the identity of all vendors from which safety-related and non-safety-related fasteners have been obtained within the past 10 years, a reasonable period which will not put undue burden on addressees. This information will assist the NRC in determining whether nuclear facility fasteners in use have been supplied in accordance with their intended use. In addition, this information is needed so that the NRC can properly coordinate information with other government agencies concerned with problems identified in the quality of fasteners

Safeguards at NRC licensed facilities: Are we doing enough

International Nuclear Information System (INIS)

Asselstine, J.K.

1986-01-01

The Nuclear Regulatory Commission is pursuing a number of initiatives in the safeguards area. The Commission is conducting a reassessment of its safeguards design basis threat statements to consider the possible implications of an explosive-laden vehicle for U.S. nuclear safeguards and to examine the comparability of safeguards features at NRC-licensed and DOE facilities. The Commission is also completing action on measures to protect against the sabotage threat from an insider at NRC-licensed facilities, and is examining the potential safety implications of safeguards measures. Finally, the NRC has developed measures to reduce the theft potential for high-enriched uranium

NRC staff review of licensee responses to pressure-locking and thermal-binding issue

Energy Technology Data Exchange (ETDEWEB)

Rathbun, H.J.

1996-12-01

Commercial nuclear power plant operating experience has indicated that pressure locking and thermal binding represent potential common mode failure mechanisms that can cause safety-related power-operated gate valves to fail in the closed position, thus rendering redundant safety-related systems incapable of performing their safety functions. In Generic Letter (GL) 95-07, {open_quotes}Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves,{close_quotes} the U.S. Nuclear Regulatory Commission (NRC) staff requested that nuclear power plant licensees take certain actions to ensure that valves susceptible to pressure locking or thermal binding are capable of performing their safety functions within the current licensing bases of the facility. The NRC staff has received summary information from licensees in response to GL 95-07 describing actions they have taken to prevent the occurrence of pressure locking and thermal binding. The NRC staff has developed a systematic process to help ensure uniform and consistent review of licensee submittals in response to GL 95-07.

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

NRC perspectives on the digital system review process

International Nuclear Information System (INIS)

Mauck, J.L.

1998-01-01

Since about 1988, the USNRC has been involved in the review of digital retrofits to instrumentation and control (I and C) systems in nuclear power plants. Initially, this involvement was limited but with the advent of the 1990s, NRC involvement has become greater because of increased interest in and application of digital systems as existing analog systems become obsolete. Criteria for the design of such systems to ensure safety has been promulgated over the years and the USNRC has been actively involved both nationally and internationally with this effort. With the publication of the Zion Eagle 21 Safety Evaluation Report in 1992, Generic Letter 95-02 in April 1995 which endorses EPRI guidance document TR-102348 on digital upgrades and the latest revision to Regulatory Guide 1.152 which endorses IEEE 7.4.3.2-1993; a basic digital system review process was established. The NRC supplemented this review process with recently issued inspection procedures for use by NRC inspectors when conducting onsite reviews of digital modifications. In addition, the NRC undertook a major effort to codify the above guidance and the experience gained from digital system reviews of both operating plant modifications and advanced reactor designs, over these years into a revision to Standard Review Plan, (SRP), NUREG-0800, Chapter 7, Instrumentation and Control. This SRP revision was published in June, 1997, and included new SRP sections, branch technical positions and six new regulatory guides endorsing IEEE standards on software quality. The NRC staff believes that a stable digital system review process is now in place. (author)

NRC sponsored rotating equipment vibration research: a program description and progress report

International Nuclear Information System (INIS)

Nitzel, M.E.

1986-01-01

The Idaho National Engineering Laboratory (INEL) is currently involved in a research project sponsored by the United States Nuclear Regulatory Commission (NRC) regarding operational vibration in rotating equipment. The object of this program is to assess the nature of vibrational failures and the effect that improved qualification standards may have in reducing the incidence of failure. In order to limit the scope of the initial effort, safety injection (SI) pumps were chosen as the component group for concentrated study. The task has been oriented to addressing the issues of whether certain SI pumps experience more failures than others, examining the dynamic environments in operation, examining the adequacy of current qualification standards, and examining what performance parameters could be used more efficiently to predict degradation or failure. Results of a literature search performed to survey SI pump failures indicate that failures are due to a diversity of causes, many of which may not be influenced by qualification criteria. Cooperative efforts have been undertaken with a limited number of nuclear utilities to describe the variety of possible operating environments and to analyze available data. The results of this analysis as they apply to the research issues are presented and possibilities for the future direction of the program are discussed

Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Oh, Dongseok; Shin, Changhwan; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others.

Berkeley Lab Pilot on External Regulation of DOE National Laboratories by the U.S. NRC

International Nuclear Information System (INIS)

Zeman, Gary H.

1999-01-01

The US Department of Energy and the US Nuclear Regulatory Commission entered into an agreement in November 1997 to pursue external regulation of radiation safety at DOE national laboratories through a Pilot Program of simulated regulation at 6-10 sites over a 2 year period. The Ernest Orlando Lawrence Berkeley National Laboratory (Berkeley Lab), the oldest of the DOE national laboratories, volunteered and was selected as the first Pilot site. Based on the similarities and linkages between Berkeley Lab and nearby university research laboratories, Berkeley Lab seemed a good candidate for external regulation and a good first step in familiarizing NRC with the technical and institutional issues involved in regulating laboratories in the DOE complex. NRC and DOE team members visited Berkeley Lab on four occasions between October 1997 and January 1998 to carry out the Pilot. The first step was to develop a detailed Work Plan, then to carry out both a technical review of the radiation safety program and an examination of policy and regulatory issues. The Pilot included a public meeting held in Oakland, CA in December 1997. The Pilot concluded with NRC's assessment that Berkeley Lab has a radiation protection program adequate to protect workers, the public and the environment, and that it is ready to be licensed by the NRC with minor programmatic exceptions. A draft final report of the Pilot was prepared and circulated for comment as a NUREG document (dated May 7, 1998). The report's recommendations include extending NRC regulatory authority to cover all ionizing radiation sources (including accelerators, x-ray units, NARM) at Berkeley Lab. Questions remaining to be resolved include: who should be the licensee (DOE, the Lab, or both)?; dealing with legacy issues and NRC D and D requirements; minimizing dual oversight; quantifying value added in terms of cost savings, enhanced safety, and improved public perception; extrapolating results to other national laboratories; and

Safety reviews of next-generation light-water reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

NRC policy on future reactor designs

International Nuclear Information System (INIS)

1985-07-01

On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions

Initial experience with the NRC significance determination process

International Nuclear Information System (INIS)

Madison, A.L.

2001-01-01

The U.S. Nuclear Regulatory Commission (NRC) has revamped its inspection, assessment, and enforcement programs for commercial nuclear power plants. The new oversight process uses more objective, timely, and safety-significant criteria in assessing performance, while seeking to more effectively and efficiently regulate the industry. The NRC tested the new process at thirteen reactors at nine sites across the country on a pilot basis in 1999 to identify what things worked well and what improvements were called for before beginning Initial Implementation at all US nuclear power plants on April 2, 2000. After a year of experience has been gained with the new oversight process at all US plants, the NRC anticipates making further improvements based on this wider experience. (author)

Initial experience with the NRC significance determination process

Energy Technology Data Exchange (ETDEWEB)

Madison, A.L. [Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission (United States)

2001-07-01

The U.S. Nuclear Regulatory Commission (NRC) has revamped its inspection, assessment, and enforcement programs for commercial nuclear power plants. The new oversight process uses more objective, timely, and safety-significant criteria in assessing performance, while seeking to more effectively and efficiently regulate the industry. The NRC tested the new process at thirteen reactors at nine sites across the country on a pilot basis in 1999 to identify what things worked well and what improvements were called for before beginning Initial Implementation at all US nuclear power plants on April 2, 2000. After a year of experience has been gained with the new oversight process at all US plants, the NRC anticipates making further improvements based on this wider experience. (author)

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

New trends in the evaluation and implementation of the safety-related operating experience associated with NRC-licensed reactors

International Nuclear Information System (INIS)

Michelson, C.; Heltemes, C.J.

1981-01-01

This article is an overview of the Nuclear Regulatory Commission program for the evaluation and dissemination of the safety-related operating experience associated with all NRC-licensed reactors. It discusses the historical background and past problems that led to the recent formation of NRC's Office for Analysis and Evaluation of Operational Data (AEOD) and details its activities, organization, staffing, and proposed analysis and evaluation methodology. The programs of industry organizations and nuclear plant licensees and the integration of foreign operating experience are included in the overview. The problems and limitations of the Licensee Event Report (LER) program and the Nuclear Plant Reliability Data system program are discussed. The AEOD analysis and evaluation methodology program includes some new improvements in the assessment of safety-related operating experience. Of particular note is the sequence coding and search procedure being developed by AEOD under a contract with the Nuclear Safety Information Center at the Oak Ridge National Laboratory. This computer-based retrieval system will have markedly improved search strategy capability for such items as commoncause failures or complex system interactions involving various failure sequences and other relationships associated with an event. The system retrieves failure data and information on the principal LER occurrence and on related component and system responses. The computer-generated Power Reactor Watch List enables AEOD to monitor all critical or unusual situations warranting close attention because of potential public health and safety. This listing is supported by a preestablished computer search strategy of the historical data base permitting identification of all past events and statistical information that are applicable to the situation being watched

International exchange of safety and licensing information

International Nuclear Information System (INIS)

Lafleur, J.D. Jr.; Hauber, R.D.; Chenier, D.M.

1977-01-01

A network of formal and informal bilateral arrangements for the exchange of nuclear safety information is being established by the U.S. Nuclear Regulatory Commission. For developing countries, such arrangements can provide ready access to the extensive, fully documented safety analyses and safety research results that NRC has accumulated. NRC has been receiving foreign visitors at a rate of about 500 per year, largely for discussions of safety and licensing questions related to light water reactors. Exchanges also are taking place on the safety of advanced reactors. A special interest of the NRC is in providing for reciprocal communicaion, at the earliest possible time, of important problems, decisions and other actions on nuclear safety matters. For example, it is essential that a newly-discovered problem in a nuclear reactor be brought immediately to the attention of other governments which are responsible for the safety of similar reactors. Definite progress has been made in the U.S. Freedom of Information Act. Certain exchanges have taken place on this basis. Experience in the establishment and operation of NRC's bilateral exchange arrangements is summarized. A typical exchange with the regulatory authority of country building its first power reactor is described

NRC program of inspection and enforcement

International Nuclear Information System (INIS)

LeDoux, J.C.; Rehfuss, C.

1978-01-01

The Nuclear Regulatory Commission (NRC) regulates civilian uses of nuclear materials to ensure the protection of the public health and safety and the environment. The Office of Inspection and Enforcement (IE) develops and implements the inspection, investigation, and enforcement programs for the NRC. The IE conducts inspection programs for reactors under construction and in operation, nuclear industry vendors, fuel facilities and users of nuclear materials, and all aspects of the safeguarding of facilities and materials. Recently the IE began implementing a program that will place inspectors on site at nuclear power reactors and will provide for national appraisal of licensee performance and for an evaluation of the effectiveness of the inspection programs

Respirator studies for the Nuclear Regulatory Commission (NRC)

International Nuclear Information System (INIS)

Skaggs, B.J.; Fairchild, C.I.; DeField, J.D.; Hack, A.L.

1985-01-01

A project of the Health, Safety and Environment Division is described. The project provides the NRC with information of respiratory protective devices and programs for their licensee personnel. The following activities were performed during FY 1983: selection of alternate test aerosols for quality assurance testing of high-efficiency particulate air respirator filters; evaluation of MAG-1 spectacles for use with positive and negative-pressure respirators; development of a Manual of Respiratory Protection in Emergencies Involving Airborne Radioactive Materials, and technical assistance to NRC licensees regarding respirator applications. 2 references, 1 figure

NRC Information Notice No. 91-29, Supplement 1: Deficiencies identified during electrical distribution system functional inspections

International Nuclear Information System (INIS)

Rossi, C.E.

1993-01-01

During multidiscipline inspections such as safety system functional inspections (SSFIs) or safety system outage modification inspections (SSOMIs), the NRC identified a number of deficiencies related to the electrical distribution system (EDS). As a result of these deficiencies, the NRC developed the EDSFI to specifically evaluate the EDS. Since 1989, the NRC has performed over 50 EDSFIs, and found design weaknesses in the following generic areas: (1) undervoltage relay setpoints for degraded grid conditions; (2) interrupting capacity of fault protection devices; (3) improper coordination of fault protection devices; (4) analysis of emergency diesel generator (EDG) capacity to power safety-related loads during postulated accidents; and (5) EDG mechanical interfaces. Each of these issues are discussed

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-30

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

Nuclear Safety Research Review Committee

International Nuclear Information System (INIS)

Todreas, N.E.

1990-01-01

The Nuclear Safety Research Review Committee has had a fundamental difficulty because of the atmosphere that has existed since it was created. It came into existence at a time of decreasing budgets. For any Committee the easiest thing is to tell the Director what additional to do. That does not really help him a lot in this atmosphere of reduced budgets which he reviewed for you on Monday. Concurrently the research arm of Nuclear Regulatory Commission has recognized that the scope of its activity needed to be increased rather than decreased. In the last two-and-a-half-year period, human factors work was reinstated, radiation and health effects investigations were reinvigorated, research in the waste area was given significant acceleration. Further, accident management came into being, and the NRC finally got back into the TMI-2 area. So with all of those activities being added to the program at the same time that the research budget was going down, the situation has become very strained. What that leads to regarding Committee membership is a need for technically competent generalists who will be able to sit as the Division Directors come in, as the contractors come in, and sort the wheat from the chaff. The Committee needs people who are interested in and have a broad perspective on what regulatory needs are and specifically how safety research activities can contribute to them. The author summarizes the history of the Committee, the current status, and plans for the future

Introductory user's manual for the US Nuclear Regulatory Commission Reactor Safety Research Data Bank. Revision 3

International Nuclear Information System (INIS)

Hardy, H.A.; Laats, E.T.

1985-03-01

The United States Nuclear Regulatory Commission (USNRC) has established the NRC/Division of Accident Evaluation (DAE) Data Bank Program to collect, store, and make available data from the many domestic and foreign water reactor safety research programs. Local direction of the program is provided by EG and G Idaho, Inc., prime contractor for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). The NRC/DAE Data Bank Program provides a central computer storage mechanism and access software for data to be used by code development and assessment groups in meeting the code correlation needs of the nuclear industry. The administrative portion of the program provides data entry, documentation, training, and advisory services to users and the USNRC. The NRC/DAE Data Bank and the capabilities of the data access software are described in this document

Summary of LWR safety research in the USA

International Nuclear Information System (INIS)

Murley, T.E.; Tong, L.S.; Bennett, G.L.

1977-01-01

quantification of reactor behaviour under off-normal conditions. The reactor operational safety research is just starting initially, it addresses fire protection, component aging and human engineering. To date, the NRC water reactor safety research program has greatly expanded the safety data base which in turn aids the further quantification of the inherent safety margins in nuclear power plants

Independent safety organization

International Nuclear Information System (INIS)

Kato, W.Y.; Weinstock, E.V.; Carew, J.F.; Cerbone, R.J.; Guppy, J.G.; Hall, R.E.; Taylor, J.H.

1985-01-01

Brookhaven National Laboratory has conducted a study on the need and feasibility of an independent organization to investigate significant safety events for the Office for Analysis and Evaluation of Operational Data, USNRC. The study consists of three parts: the need for an independent organization to investigate significant safety events, alternative organizations to conduct investigations, and legislative requirements. The determination of need was investigated by reviewing current NRC investigation practices, comparing aviation and nuclear industry practices, and interviewing a spectrum of representatives from the nuclear industry, the regulatory agency, and the public sector. The advantages and disadvantages of alternative independent organizations were studied, namely, an Office of Nuclear Safety headed by a director reporting to the Executive Director for Operations (EDO) of NRC; an Office of Nuclear Safety headed by a director reporting to the NRC Commissioners; a multi-member NTSB-type Nuclear Safety Board independent of the NRC. The costs associated with operating a Nuclear Safety Board were also included in the study. The legislative requirements, both new authority and changes to the existing NRC legislative authority, were studied. 134 references

Application of a methodology to determine priorities for nuclear power plant safety issues

International Nuclear Information System (INIS)

Daling, P.M.

1988-01-01

The Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES) is sponsoring a research program to determine priorities of nuclear power plant safety issues. A methodology has been developed at the Pacific Northwest Laboratory (PNL) to provide technical assistance in the development of risk and cost estimates for implementing resolutions to the safety issues. The information development methods are intended to provide the NRC with a consistent level of information for use in ranking the issues. The NRC uses this information, along with judgmental factors, to rank the issues for further consideration by the NRC staff. The primary purpose of the priority rankings are to assist in the allocation of resources to issues that have high potential for reducing public risk as well as to remove issues from further consideration that have little safety significance

Status report on NRC's current below regulatory concern activities

International Nuclear Information System (INIS)

Dragonette, K.S.

1988-01-01

The concept of below regulatory concern (BRC) is not new to the Nuclear Regulatory Commission (NRC) or its predecessor agency, the Atomic Energy Commission. The regulations and licensing decisions have involved limited and de facto decisions on BRC since the beginning. For example, consumer products containing radioactive materials have been approved for distribution to persons exempt from licensing for some time and procedures for survey and release of equipment have traditionally been a part of many licensees' radiation safety programs. However, these actions have generally been ad hoc decisions in response to specific needs and have not been necessarily consistent. The need to deal with this regulatory matter has been receiving attention from both Congress and the NRC Commissioners. NRC response has grown from addressing specific waste streams, to generic rulemaking for wastes, and finally to efforts to develop a broad generic BRC policy. Section 10 of the Low-Level Radioactive Waste Policy Amendments Act of 1985 addressed NRC actions on specific waste streams. In response, NRC issued guidance on rulemaking petitions for specific wastes. NRC also issued an advance notice of proposed rulemaking indicating consideration of Commission initiated regulations to address BRC wastes in a generic manner. The Commissioners have directed staff to develop an umbrella policy for all agency decisions concerning levels of risk or dose that do not require government regulation

U.S. NRC's generic issues program

International Nuclear Information System (INIS)

Kauffman, J.V.; Foster, J.W.

2008-01-01

The United States Nuclear Regulatory Commission (NRC) has a Generic Issues Program (GIP) to address Generic Issues (GI). A GI is defined as 'a regulatory matter involving the design, construction, operation, or decommissioning of several, or a class of, NRC licensees or certificate holders that is not sufficiently addressed by existing rules, guidance, or programs'. This rather legalistic definition has several practical corollaries: First, a GI must involve safety. Second, the issue must involve at least two plants, or it would be a plant-specific issue rather than a GI. Third, the potential safety question must not be covered by existing regulations and guidance (compliance). Thus, the effect of a GI is to potentially change the body of regulations and associated guidance (e.g., regulatory guides). The GIP was started in 1976, thus it is a relatively mature program. Approximately 850 issues have been processed by the program to date. More importantly, even after 30 years, new GIs continue to be proposed. The entire set of Generic Issues (GIs) is updated annually in NUREG-0933, 'A Prioritization of Generic Safety Issues'. GIs normally involve complex questions of safety and regulation. Efficient and effective means of addressing these issues are very important for regulatory effectiveness. If an issue proves to pose a genuine, significant safety question, then swift, effective, enforceable, and cost-effective action needs to be taken. Conversely, if an issue is of little safety significance, the issue should be dismissed in an expeditious manner, avoiding unnecessary expenditure of resources and regulatory burden or uncertainty. This paper provides a summary of the 5-stage program, from identification through the regulatory assessment stage. The paper also includes a discussion of the program's seven criteria, sources of proposed GIs, recent improvements, publicly available information, historical performance, and status of current GIs. (authors)

Current role of the USNRC safety research program in support of the regulatory process

International Nuclear Information System (INIS)

Levine, S.

1979-01-01

The current role of the USNRC's safety research program is shown. Some aspects of this role in the wake of the TMI accident are discused as well as some historical perspective on the development of USNRC's program, its relationship with the NRC mission, an overview of the program activities and some recent research results, and finally the impact of the TMI accident in clarifiying needs for expedited and new research activities, including the need for a greatly enhanced use of probabilistic analysis techniques to improve the coherence of its regulatory process. (author)

NRC perspective and experience on valve testing

International Nuclear Information System (INIS)

Eapen, P.K.

1990-01-01

Testing of safety related valves is one of the major activities at commercial nuclear power plants. In addition to Technical Specification, valve testing is required in 10 CFR 50.55a and 10 CFR 50 Appendix J. NRC inspectors (both resident and specialists) spend a considerable amount of time in following the valve test activities as part of their routine business. In the past, depending on a licensee's organizational structure, a valve could be tested more than three times to verify conformance with Technical Specifications, 10 CFR 50.55a, and 10 CFR 50 Appendix J. The regulatory reviewers were isolated from each other. Licensee test personnel were also not communicating among themselves. As a result, NRC inspectors found that certain valves in the IST program were inadequately tested. The typical licensee response was to say that this valve is exempted from testing under Appendix J. Others would say that the technical specification does not require fast closure of a valve in question. In addition to the above, the inspectors had to deal with exemption requests that were not dispositioned by the NRC. In the seventies there was a gentlemen's agreement to allow the licensee to do the testing in accordance with the exception, without waiting for the NRC approval. Needless to say when the new NRC inspection procedure was issued in March 1989 for implementation, the Regional inspectors had extremely difficult time to cope with the gray areas of valve testing. In August 1987, NRC Region I was reorganized and the special test program section was established to perform inspections in the IST area. This section was chartered to optimize resources and develop a meaningful inspection plan. The perspectives and insights used in the development of a detailed inspection plan is discussed below

Safety evaluation status report for the prototype license application safety analysis report

International Nuclear Information System (INIS)

1989-07-01

The US Nuclear Regulatory Commission (NRC) staff and consultants reviewed a Prototype License Application Safety Analysis Report (PLASAR) submitted by the US Department of Energy (DOE) for the earth-mounded concrete bunker (EMCB) alternative method of low-level radioactive waste disposal. The NRC reviewers relied extensively on the Standard Review Plan (SRP), Rev.1 (NUREG-1200), to evaluate the acceptability of the information provided in the EMCB PLASAR. The NRC staff selected certain review areas in the PLASAR for development of safety evaluation report input to provide examples of safety assessments that are necessary as part of a licensing review. Because of the fictitious nature of the assumed disposal site, and the decision to limit the review to essentially first-round review status, the NRC staff report is labeled a ''Safety Evaluation Status Report'' (SESR). Appendix A comprises the NRC review comments and questions on the information that DOE submitted in the PLASAR. The NRC concentrated its review on the design and operations-related portions of the EMCB PLASAR

NRC review of passive reactor design certification testing programs: Overview, progress, and regulatory perspective

Energy Technology Data Exchange (ETDEWEB)

Levin, A.E.

1995-09-01

New reactor designs, employing passive safety systems, are currently under development by reactor vendors for certification under the U.S. Nuclear Regulatory Commission`s (NRC`s) design certification rule. The vendors have established testing programs to support the certification of the passive designs, to meet regulatory requirements for demonstration of passive safety system performance. The NRC has, therefore, developed a process for the review of the vendors` testing programs and for incorporation of the results of those reviews into the safety evaluations for the passive plants. This paper discusses progress in the test program reviews, and also addresses unique regulatory aspects of those reviews.

Compilation of fastener testing data received in response to NRC Compliance Bulletin 87-02

International Nuclear Information System (INIS)

Cwalina, G.C.; Conway, J.T.; Parker, L.B.

1989-06-01

On November 6, 1987, the Nuclear Regulatory Commission (NRC) issued Bulletin 87-02, ''Fastener Testing to Determine Conformance With Applicable Material Specifications,'' to all holders of operating licenses or construction permits for nuclear power reactors (licensees). The bulletin was issued so that the NRC staff could gather data to determine whether counterfeit fasteners are a problem in the nuclear power industry. The bulletin requested nuclear power plant owners to determine whether fasteners obtained from suppliers and/or manufacturers for use in their facilities meet the mechanical and chemical specifications stipulated in the procurement documents. The licensees were requested to sample a minimum of 10 safety-related and 10 non-safety-related fasteners (studs, bolts, and/or cap screws) and a sample of typical nuts that would be used with each fastener and to report the testing results to the NRC. The results of this study did not indicate a safety concern relating to the use of mismarked or counterfeit fasteners in the nuclear industry, but they did indicate a nonconformance rate of 8 to 12 percent for fasteners. The NRC staff is considering taking action to improve the effectiveness of receipt inspection and testing programs for all materials at nuclear power plants

Cohort Profile : The National Academy of Sciences-National Research Council Twin Registry (NAS-NRC Twin Registry)

NARCIS (Netherlands)

Gatz, Margaret; Harris, Jennifer R.; Kaprio, Jaakko; McGue, Matt; Smith, Nicholas L.; Snieder, Harold; Spiro, Avron; Butler, David A.

The National Academy of Sciences-National Research Council Twin Registry (NAS-NRC Twin Registry) is a comprehensive registry of White male twin pairs born in the USA between 1917 and 1927, both of the twins having served in the military. The purpose was medical research and ultimately improved

Assessment of operational safety data in the Nuclear Regulatory Commission (NRC)

International Nuclear Information System (INIS)

Michelson, C.; Heltemes, C.J.

1981-01-01

The collection, assessment, and dissemination of operational safety data, including the Licensee Event Reports (LERs) is the principal activity of the NRC's Office for Analysis and Evaluation of Operational Data (AEOD). This office was recently formed to provide a dedication to this activity. It has been staffed and fully operational since April 1980. The office programs are evolving and include some new ideas and techniques to aid in the assessment of LERs. For example, the office is managing the development of a computer-based Sequence Coding and Search (SCS) procedure which will have the capability to store and retrieve the detail and individual sequences associated with each LER. Such events may be rather complex and involve a number of isolated or unrelated happenings associated with multiple systems and components with various causes, failure modes, and failure effects. Thus, the SCS system is particularly useful because it documents in a computer-retrievable form not only the principal occurrence, but also related component and system responses which precede, accompany, follow, or result from the principal occurrence. Also noteworthy is the Power Reactor Watch List which is being developed and monitored as a part of the AEOD program

NRC as referee (reactor licensing following the Three Mile Island accident)

International Nuclear Information System (INIS)

Eisenhut, D.G.

1984-01-01

In this article, the NRC's licensing director reports on the progress made by US utilities in complying with the key regulations stemming from the Three Mile Island accident. Over 130 items must be improved at more than 65 reactors. The actions taken by France in response to its own analysis of the accident are discussed. New NRC requirements with regard to operational safety, design, and emergency-response capability are outlined. Nearly all the training, or software, items in Nureg-0737 (''Clarification of TMI Action Plan Requirements'') and more than half of the mechanical, or hardware, items have been completed at plants with operating reactors. The Committee to Review Generic Requirements was created to develop means for controlling the number and nature of NRC requirements placed on licensees. Probabilistic risk-assessment techniques were not widely used by the NRC until after the Three Mile Island accident. The NRC has directed licensees and applicants for operating licenses to conduct control-room design reviews to identify and correct human-engineering discrepancies. Includes 2 tables

Status of safety issues at licensed power plants: TMI action plan requirements, unresolved safety issues, generic safety issues

International Nuclear Information System (INIS)

1991-12-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program was established whereby an annual NUREG report would be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was compiled and reported in three NUREG volumes. Volume 1, published in March 1991, addressed the status of of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). This annual NUREG report combines these volumes into a single report and provides updated information as of September 30, 1991. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. This report is to provide a comprehensive description of the implementation and verification status of TMI Action Plan Requirements, safety issues designated as USIs, and GSIs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. An additional purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

NRC inspections of licensee activities to improve the performance of motor-operated valves

International Nuclear Information System (INIS)

Scarbrough, T.G.

1992-01-01

The NRC regulations require that components important to the safe operation of a nuclear power plant be treated in a manner that provides assurance of their proper performance. Despite these regulatory requirements, operating experience and research programs have raised concerns regarding the performance of motor-operated valves (MOVs) in nuclear power plants. In June 1990, the staff issued NUREG-1352, Action Plans for Motor-Operated Valves and Check Valves, which contains planned actions to organize the activities aimed at resolving the concerns about MOV performance. A significant task of the MOV action plan is the staff's review of the implementation of Generic Letter (GL) 89-10 (June 28, 1989), 'Safety-Related Motor-Operated Valve Testing and Surveillance,' and its supplements, by nuclear power plant licensees. The NRC staff has issued several supplements to GL 89-10 to provide additional guidance for use by licensees in responding to the generic letter. The NRC staff has conducted initial inspections of the GL 89-10 programs at most licensee facilities. This paper outlines some of the more significant findings of those inspections. For example, licensees who have begun differential pressure and flow testing have found some MOVs to require more thrust to operate than predicted by the standard industry equation with typical valve factors assumed in the past. The NRC staff has found weaknesses in licensee procedures for conducting the differential pressure and flow tests, the acceptance criteria for the tests in evaluating the capability of the MOV to perform its safety function under design basis conditions, and feedback of the test results into the methodology used by the licensee in predicting the thrust requirements for other MOVs. Some licensees have not made adequate progress toward resolving the MOV issue for their facilities within the recommended schedule of GL 89-10

Research on Integration of NPP Operational Safety Management Performance Systems

International Nuclear Information System (INIS)

Chi, Miao; Shi, Liping

2014-01-01

The operational safety management of Nuclear Power Plants demands systematic planning and integrated control. NPPs are following the well-developed safety indicator systems proposed by IAEA Operational Safety Performance Indicator Programme, NRC Reactor Oversight Process or the other institutions. Integration of the systems is proposed to benefiting from the advantages of both systems and avoiding improper application into the real world. The authors analyzed the possibility and necessity for system integration, and propose an indicator system integrating method

Reports distributed in 1976 under the NRC Light-Water Reactor Safety Technical Exchange. Volume II

International Nuclear Information System (INIS)

Sharp, D.S.; Cottrell, W.B.

1977-01-01

Lists of documents exchanged in 1976 under agreement between the U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research and the governments of France, Federal Republic of Germany, and Japan are presented. In 1976 the NRC received 25 reports from France, 74 from F.R. Germany, and 39 from Japan, and in return sent 119 U.S. reports to France, 154 to F.R. Germany, and 155 to Japan

NRC closing remarks

International Nuclear Information System (INIS)

Coffman, F.

1994-01-01

This section contains the edited transcript of the NRC closing remarks made by Mr. Franklin Coffman (Chief, Human Factors Branch, Office of Nuclear Regulatory Research) and Dr. Cecil Thomas (Deputy Director, Division of Reactor Controls and Human Factors, Office of Nuclear Reactor Regulation). This editing consisted of minimal editing to correct grammar and remove extraneous references to microphone volume, etc

MHTGR demonstration role in the NRC design certification process

International Nuclear Information System (INIS)

Kelley, A.P. Jr.; Jones, G.

1986-01-01

A modular high-temperature gas-cooled reactor (MHTGR) design is being developed by the US HTGR Program. Because of the small size of the individual modules that would make up a commercial facility, it appears feasible to design and construct a single-module demonstration plant within the funding constraints on the public and private-sector program participants. Furthermore, the safety margins that can be made inherent to the design permit full-scale testing that could supply a new basis for demonstrating investment protection and safety adequacy to the public, the US Nuclear Regulatory Commission (NRC), and potential users. With this in mind, a Project Definition Study was sponsored by Gas-Cooled Reactor Associates and the Tennessee Valley Authority to study the potential benefits of undertaking such a demonstration project. One of the areas investigated was the potential benefits of such a facility in supporting the NRC design certification process, which is envisioned as a necessary commercialization step for the MHTGR

General statement of policy and procedures for NRC enforcement actions: Enforcement policy. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-05-01

This document includes the US Nuclear Regulatory Commission`s (NRC`s or Commission`s) revised General Statement of Policy and Procedure for Enforcement Actions (Enforcement Policy) as it was published in the Federal Register on May 13, 1998 (63 ER 26630). The Enforcement Policy is a general statement of policy explaining the NRC`s policies and procedures in initiating enforcement actions, and of the presiding officers and the Commission in reviewing these actions. This policy statement is applicable to enforcement matters involving the radiological health and safety of the public, including employees` health and safety, the common defense and security, and the environment.

General statement of policy and procedures for NRC enforcement actions: Enforcement policy. Revision 1

International Nuclear Information System (INIS)

1998-05-01

This document includes the US Nuclear Regulatory Commission's (NRC's or Commission's) revised General Statement of Policy and Procedure for Enforcement Actions (Enforcement Policy) as it was published in the Federal Register on May 13, 1998 (63 ER 26630). The Enforcement Policy is a general statement of policy explaining the NRC's policies and procedures in initiating enforcement actions, and of the presiding officers and the Commission in reviewing these actions. This policy statement is applicable to enforcement matters involving the radiological health and safety of the public, including employees' health and safety, the common defense and security, and the environment

Staged licensing: An essential element of the NRC's revised regulations

International Nuclear Information System (INIS)

Echols, F.S.

1997-01-01

Over the past several years, Congress has directed the Department of Energy (DOE), the Nuclear Regulatory Commission (NRC), and the Environmental Protection Agency (EPA) to abandon their efforts to assess an array of potential candidate geologic repository sites for the permanent disposal of spent nuclear reactor fuel and high level radioactive waste, to develop generally applicable requirements for licensing geologic repositories, and to develop generally applicable radiation protection standards for geologic repositories, and instead to focus their efforts to determine whether a single site located at Yucca Mountain, Nevada can be developed as a geologic repository which providing reasonable assurance that public health and safety and the environment will be adequately protected. If the Yucca Mountain site is found to be suitable for development as a geologic repository, then at each stage of development DOE will have to provide the NRC with progressively more detailed information regarding repository design and long-term performance. NRC regulations reflect the fact that it will not be until the repository has been operated for a number of years that the NRC will be able to make a final determination as to long-term repository performance. Nevertheless, the NRC will be able to allow DOE to construct and operate a repository, provided that the NRC believes that the documented results of existing studies, together with the anticipated results from continuing and future studies, will enable the NRC to make a final determination that it has reasonable assurance that the repository system's long-term performance will not cause undue risk to the public. Thus, in its efforts to revise its current regulations to assure that the technical criteria are specifically applicable to the Yucca Mountain site, the NRC should also make sure that it preserves and clarifies the concept of staged repository development

Recent developments in NRC guidelines for atmosphere cleanup systems

International Nuclear Information System (INIS)

Bellamy, R.R.

1976-01-01

The Nuclear Regulatory Commission (NRC) maintains the policy of updating when necessary, its published guidance for the design of engineered safety feature (ESF) and normal ventilation systems. The guidance is disseminated by means of issuing new, or revisions to, existing Regulatory Guides, Standard Review Plans, Branch Technical Positions and Technical Specifications. A revised Regulatory Guide, new Technical Specifications and new Standard Review Plans with Branch Technical Positions for atmosphere cleanup systems are discussed. Regulatory Guide 1.52, ''Design, Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants,'' was issued in July 1973. The major comments received from the nuclear industry since the guide was issued, NRC's experience in implementing the guide in recent license applications, status of operating plants in meeting the guidelines and NRC's continuing assessment of operating data and laboratory tests to assure that the guide reflects the latest technology are discussed

NRC review of passive reactor design certification testing programs: Overview and regulatory perspective

International Nuclear Information System (INIS)

Levin, A.E.

1993-01-01

Reactor vendors are developing new designs for future deployment, including open-quotes passiveclose quotes light water reactors (LWRs), such as General Electric's (G.E.'s) simplified boiling water reactor (SBWR) and Westinghouse's AP600, which depend primarily on inherent processes, such as national convection and gravity feed, for safety injection and emergency core cooling. The U.S. Nuclear Regulatory Commission (NRC) has implemented a new process, certification of standardized reactor designs, for licensing these Plants. Part 52 of Title 10 of the Code of Federal Regulations (10CFR52) contains the requirements that vendors must meet for design certification. One important section, 10CFR52.47, reads open-quotes Certification of a standard design which . . . utilizes simplified, inherent, passive, or other innovative means to accomplish its safety functions will be granted only if: (1) The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof; (2) Interdependent effects among the safety features have been found acceptable by analysis, appropriate test programs, experience, or a combination thereof; and (3) Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses. . . . The vendors have initiated programs to test innovative features of their designs and to develop data bases needed to validate their analytical codes, as required by the design certification rule. Accordingly, the NRC is reviewing and evaluating the vendors programs to ensure that they address adequately key issues concerning safety system performance. This paper provides an overview of the NRC's review process and regulatory perspective

NRC wants plant-specific responses on Thermo-Lag

International Nuclear Information System (INIS)

Anon.

1994-01-01

Dissatisfied with recent industry-backed efforts to assure fire safety at nuclear power plants, the Nuclear Regulatory Commission announced on November 24 that it would direct all nuclear plant owners to specify the actions they would take to assure that the use of the Thermo-Lag 330 fire barrier material would not lead to insufficient protection of electrical cables connected to safe-shutdown systems. Previously, the NRC had been content to let the matter wait until tests sponsored by the Nuclear Management and Resources Council (Numarc) could show whether Thermo-Lag, used and installed in certain ways, would provide sufficient protection, but the NRC and Numarc have disagreed over the test methodology, and the Numarc tests are now considered to be several months behind schedule

Safety standards and safety record of nuclear power plants

International Nuclear Information System (INIS)

Davis, A.B.

1984-01-01

This paper focuses on the use of standards and the measurement and enforcement of these standards to achieve safe operation of nuclear power plants. Since a discussion of the safety standards that the Nuclear Regulatory Commission (NRC) uses to regulate the nuclear power industry can be a rather tedious subject, this discussion will provide you with not only a description of what safety standards are, but some examples of their application, and various indicators that provide an overall perspective on safety. These remarks are confined to the safety standards adopted by the NRC. There are other agencies such as the Environmental Protection Agency, the Occupational Safety and Health Administration, and the state regulatory agencies which impact on a nuclear power plant. The NRC has regulatory authority for the commercial use of the nuclear materials and facilities which are defined in the Atomic Energy Act of 1954 to assure that the public health and safety and national security are protected

Development of the NRC's Human Performance Investigation Process (HPIP)

International Nuclear Information System (INIS)

Paradies, M.; Unger, L.; Haas, P.; Terranova, M.

1993-10-01

The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause events at nuclear power plants. This document, Volume III, is a detailed documentation of the development effort and the pilot training program

An appropriate level for reactor regulatory research

International Nuclear Information System (INIS)

Budnitz, R.J.; Kouts, H.J.C.

1986-01-01

In this brief paper the appropriate role for NRC's research program is set down, based on the experience of the last decade; and the broad outlines of an appropriate research program for the next decade are discussed. The two authors bring to this topic direct personal experience: they are two of only three former directors of NRC's research program in its history (the other was the last Saul Levine). The discussion considers only reactor safety research, although NRC's research program covers several other areas as well. Safeguards, environmental and health studies, waste management, and transportation are not covered

Systematic evaluation program review of NRC Safety Topic VI-10.A associated with the electrical, instrumentation and control portions of the testing of reactor trip system and engineered safety features, including response time for the Dresden station, Unit II nuclear power plant

International Nuclear Information System (INIS)

St Leger-Barter, G.

1980-11-01

This report documents the technical evaluation and review of NRC Safety Topic VI-10.A, associated with the electrical, instrumentation, and control portions of the testing of reactor trip systems and engineered safety features including response time for the Dresden II nuclear power plant, using current licensing criteria

Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

International Nuclear Information System (INIS)

Carter, R.E.

1985-01-01

During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

Licensing considerations in converting NRC-licensed non-power reactors from high-enriched to low-enriched uranium fuels

Energy Technology Data Exchange (ETDEWEB)

Carter, R E

1985-07-01

During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non-power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public comment that would require all NRC-licensed non-power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modifications, and does not involve an unreviewed safety question, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regulatory method of achieving this principle has not been finalized. (author)

Recommendations for NRC policy on shift scheduling and overtime at nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Lewis, P.M.

1985-07-01

This report contains the Pacific Northwest Laboratory's (PNL's) recommendations to the US Nuclear Regulatory Commission (NRC) for an NRC policy on shift scheduling and hours of work (including overtime) for control room operators and other safety-related personnel in nuclear power plants. First, it is recommended that NRC make three additions to its present policy on overtime: (1) limit personnel to 112 hours of work in a 14-day period, 192 hours in 28 days, and 2260 hours in one year; exceeding these limits would require plant manager approval; (2) add a requirement that licensees obtain approval from NRC if plant personnel are expected to exceed 72 hours of work in a 7-day period, 132 hours in 14 days, 228 hours in 28 days, and 2300 hours in one year; and (3) make the policy a requirement, rather than a nonbinding recommendation. Second, it is recommended that licensees be required to obtain NRC approval to adopt a routine 12-hour/day shift schedule. Third, it is recommended that NRC add several nonbinding recommendations concerning routine 8-hour/day schedules. Finally, because additional data can strengthen the basis for future NRC policy on overtime, five methods are suggested for collecting data on overtime and its effects. 44 refs., 10 tabs.

Recommendations for NRC policy on shift scheduling and overtime at nuclear power plants

International Nuclear Information System (INIS)

Lewis, P.M.

1985-07-01

This report contains the Pacific Northwest Laboratory's (PNL's) recommendations to the US Nuclear Regulatory Commission (NRC) for an NRC policy on shift scheduling and hours of work (including overtime) for control room operators and other safety-related personnel in nuclear power plants. First, it is recommended that NRC make three additions to its present policy on overtime: (1) limit personnel to 112 hours of work in a 14-day period, 192 hours in 28 days, and 2260 hours in one year; exceeding these limits would require plant manager approval; (2) add a requirement that licensees obtain approval from NRC if plant personnel are expected to exceed 72 hours of work in a 7-day period, 132 hours in 14 days, 228 hours in 28 days, and 2300 hours in one year; and (3) make the policy a requirement, rather than a nonbinding recommendation. Second, it is recommended that licensees be required to obtain NRC approval to adopt a routine 12-hour/day shift schedule. Third, it is recommended that NRC add several nonbinding recommendations concerning routine 8-hour/day schedules. Finally, because additional data can strengthen the basis for future NRC policy on overtime, five methods are suggested for collecting data on overtime and its effects. 44 refs., 10 tabs

NRC regulatory agenda

International Nuclear Information System (INIS)

1990-10-01

The Regulatory Agenda is a quarterly compilation of all rules on which the NRC has recently completed action or has proposed, or is considering action and of all petitions for rulemaking that the NRC has received that are pending disposition

NRC regulatory agenda

International Nuclear Information System (INIS)

1990-04-01

The Regulatory Agenda is a quarterly compilation of all rules on which the NRC has recently completed action or has proposed, or is considering action and of all petitions for rulemaking that the NRC has received that are pending disposition

Applications of probabilistic techniques at NRC

International Nuclear Information System (INIS)

Thadani, A.; Rowsome, F.; Speis, T.

1984-01-01

The NRC is currently making extensive use of probabilistic safety assessment in the reactor regulation. Most of these applications have been introduced in the regulatory activities in the past few years. Plant Probabilistic Safety Studies are being utilized as a design tool for applications for standard designs and for assessment of plants located in regions of particularly high population density. There is considerable motivation for licenses to perform plant-specific probabilistic studies for many, if not all, of the existing operating nuclear power plants as a tool for prioritizing the implementation of the many outstanding licensing actions of these plants as well as recommending the elimination of a number of these issues which are judged to be insignificant in terms of their contribution to safety and risk. Risk assessment perspectives are being used in the priorization of generic safety issues, development of technical resolution of unresolved safety issues, assessing safety significance of proposed new regulatory requirements, assessment of safety significance of some of the occurrences at operating facilities and in environmental impact analyses of license applicants as required by the National Environmental Policy Act. (orig.)

Safety evaluation report by the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission for U.S. Energy Research and Development Administration Light Water Breeder Reactor. Special project No. 561

International Nuclear Information System (INIS)

1976-07-01

The Safety Evaluation Report is presented for the Light Water Breeder Reactor (LWBR). The LWBR core is to be installed in the Shippingport reactor at the Shippingport Atomic Power Station. The Safety Evaluation Report is the result of an NRC staff review of the LWBR Safety Analysis Report submitted by the Division of Naval Reactors, U. S. Energy Research and Development Administration. As a result of its review, the NRC staff has recommended that: (1) a diverse trip signal, such as containment high pressure, be included in a 2-out-of-3 logic for initiation of safety injection; (2) power be locked out from the pressurizer surge isolation valve during normal operation; and (3) a chlorine monitor be installed in the main control room

Summary of Chernobyl followup research activities

International Nuclear Information System (INIS)

1992-06-01

In NUREG-1251, ''Implications of the Accident at Chernobyl for Safety Regulation of Commercial Nuclear Power Plants in the United States,'' April 1989, the NRC staff concluded that no immediate changes in NRC's regulations regarding design or operation of US commercial reactors were needed; however, it recommended that certain issues be considered further. NRC's Chernobyl followup research program consisted of the research tasks undertaken in response to the recommendations in NUREG-1251. It included 23 tasks that addressed potential lessons to be learned from the Chernobyl accident. This report presents summaries of NRC's Chernobyl followup research tasks. For each task, the Chernobyl-related issues are indicated, the work is described, and the staff's findings and conclusions are presented. More detailed reports concerning the work are referenced where applicable. This report closes out NRC's Chernobyl followup research program as such, but additional research will be conducted on some issues as needed. The report includes remarks concerning significant further activity with respect to the issues addressed

76 FR 57006 - Proposed Generic Communications; Draft NRC Regulatory Issue Summary 2011-XX; NRC Regulation of...

Science.gov (United States)

2011-09-15

... NUCLEAR REGULATORY COMMISSION 10 CFR Parts 30 and 150 [NRC-2011-0146] Proposed Generic Communications; Draft NRC Regulatory Issue Summary 2011-XX; NRC Regulation of Military Operational Radium-226... published for public comment the proposed draft RIS 2011-XX; NRC Regulation of Military Operational Radium...

NRC program for the resolution of generic issues related to nuclear power plants. (Includes plans for the resolution of ''unresolved safety issues'' pursuant to Section 210 of the Energy Reorganization Act of 1974, as amended)

International Nuclear Information System (INIS)

1977-12-01

This report provides a description of the Nuclear Regulatory Commission's Program for the Resolution of Generic Issues Related to Nuclear Power Plants. The NRC program is of considerably broader scope than the ''Unresolved Safety Issues Plan'' required by Section 210. The NRC program does include plans for the resolution of ''Unresolved Safety Issues''; however, in addition, it includes generic tasks for the resolution of environmental issues, for the development of improvements in the reactor licensing process and for consideration of less conservative design criteria or operating limitations in areas where over conservatisms may be unnecessarily restrictive or costly

78 FR 29392 - Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied...

Science.gov (United States)

2013-05-20

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0098] Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied on for Safety AGENCY: Nuclear Regulatory Commission. ACTION... (NRC) is issuing for public comment Draft Regulatory Issue Summary (RIS) 2013-XX, ``Embedded Digital...

Comparison between Radiology Science Laboratory, Brazil (LCR) and National Research Council, Canada (NRC) of the absorbed dose in water using Fricke dosimetry

International Nuclear Information System (INIS)

Salata, Camila; David, Mariano Gazineu; Almeida, Carlos Eduardo de

2014-01-01

The absorbed dose to water standards for HDR brachytherapy dosimetry developed by the Radiology Science Laboratory, Brazil (LCR) and the National Research Council, Canada (NRC), were compared. The two institutions have developed absorbed dose standards based on the Fricke dosimetry system. There are significant differences between the two standards as far as the preparation and readout of the Fricke solution and irradiation geometry of the holder. Measurements were done at the NRC laboratory using a single Ir-192 source. The comparison of absorbed dose measurements was expressed as the ratio Dw(NRC)/Dw(LCR), which was found to be 1.026. (author)

Development of the NRC's Human Performance Investigation Process (HPIP)

International Nuclear Information System (INIS)

Paradies, M.; Unger, L.; Haas, P.; Terranova, M.

1993-10-01

The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause events at nuclear power plants. This document, Volume I is a concise description of the need for the human performance investigation process, the process' components, the methods used to develop the process, the methods proposed to test the process, and conclusions on the process' usefulness

Development of the NRC's Human Performance Investigation Process (HPIP)

International Nuclear Information System (INIS)

Paradies, M.; Unger, L.; Haas, P.; Terranova, M.

1993-10-01

The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause event at nuclear power plants. This document, Volume II, is a field manual for use by investigators when performing event investigations. Volume II includes the HPIP Procedure, the HPIP Modules, and Appendices that provide extensive documentation of each investigation technique

Program management plan for the conduct of a research, development, and demonstration program for improving the safety of nuclear powerplants

International Nuclear Information System (INIS)

1981-12-01

Congress passed Public Law 96-567, Nuclear Safety Research, Development, and Demonstration Act of 1980, (hereafter referred to as the Act) to provide for an accelerated and coordinated program of light water reactor safety research, development, and demonstration to be carried out by the Department of Energy. In order to assure that this program would be compatible with the needs of Nuclear Regulatory Commission (NRC) and industry, the Department of Energy (DOE) initiated its response to Section 4 of the Act by conducting individual information gathering meetings with NRC and a wide cross section of the nuclear industry. The Department received recommendations on needs of what type of activities would and would not be appropriate for the Department to assist in satisfying these needs. Based on the evaluation of these inputs, it is concluded that the Department's ongoing Light Water Reactor (LWR) safety program is responsive to the Act. Specifically, the Department's ongoing program includes tasks in the areas of regulatory assessment, risk assessment, fission product source term, and emergency preparedness as well as providing technical assistance to the Institute of Nuclear Power Operations (INPO) to improve training of nuclear power personnel. These were among the very high priority efforts that were identified as necessary and appropriate for support by the Department

NRC regulatory agenda

International Nuclear Information System (INIS)

1990-01-01

The NRC Regulatory Agenda is a compilation of all rules on which the NRC has proposed or is considering action and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

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)

Proceedings of the Digital Systems Reliability and Nuclear Safety Workshop

Energy Technology Data Exchange (ETDEWEB)

Wallace, D. R.; Cuthill, B. B.; Ippolito, L. M. [National Inst. of Standards and Technology, Gaithersburg, MD (United States); Beltracchi, L. [Nuclear Regulatory Commission, Washington, DC (United States) ed.

1994-03-01

The United States Nuclear Regulatory Commission (NRC), in cooperation with the National Institute of Standards and Technology conducted the.Digital Systems Reliability and Nuclear Safety Workshop on September 13--14, 1993, in Rockville, Maryland. The workshop provided a forum for the exchange of information among experts within the nuclear industry, experts from other industries, regulators and academia. The information presented at this workshop provided in-depth exposure of the NRC staff and the nuclear industry to digital systems design safety issues and also provided feedback to the NRC from outside experts regarding identified safety issues, proposed regulatory positions, and intended research associated with the use of digital systems in nuclear power plants. Technical presentations provided insights on areas where current software engineering practices may be inadequate for safety-critical systems, on potential solutions for development issues, and on methods for reducing risk in safety-critical systems. This report contains an analysis of results of the workshop, the papers presented panel presentations, and summaries of, discussions at this workshop. The individual papers have been cataloged separately.

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

NRC influences on nuclear training

International Nuclear Information System (INIS)

Hannon, J.N.

1987-01-01

NRC influences on utility training programs through prescriptive requirements and evaluation of industry self-initiatives are discussed. NRC regulation and industry initiatives are complimentary and in some instances industry initiatives are replacing NRC requirements. Controls and feedback mechanisms designed to enhance positive NRC influences and minimize or eliminate negative influences are discussed. Industry and NRC efforts to reach an acceptable mix between regulator oversight and self-initiatives by the industry are recognized. Problem areas for continued cooperation to enhance training and minimize conflicting signals to industry are discussed. These areas include: requalification examination scope and content, depth of training and examination on emergency procedures; improved learning objectives as the basis for training and examination, and severe accident training

Status of safety issues at licensed power plants: TMI Action Plan requirements; unresolved safety issues; generic safety issues; other multiplant action issues

International Nuclear Information System (INIS)

1993-12-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. This third annual NUREG report, Supplement 3, presents updated information as of September 30, 1993. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

NRC regulatory agenda

International Nuclear Information System (INIS)

1991-04-01

The NRC Regulatory Agenda is a compilation of all rules on which the NRC has recently completed action or has proposed, or is considering action and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

NRC Regulatory Agenda

International Nuclear Information System (INIS)

1991-08-01

The NRC Regulatory Agenda is a compilation of all rules on which the NRC has recently completed action or has proposed, or is considering action and all petitions for rulemaking which have been received by the commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

NRC Regulatory Agenda

International Nuclear Information System (INIS)

1991-10-01

The NRC Regulatory Agenda is a compilation of all rules on which the NRC has recently completed action, or has proposed action, or is considering action, and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

NRC regulatory agenda

International Nuclear Information System (INIS)

1993-04-01

The NRC Regulatory Agenda is a compilation of all rules on which the NRC has recently completed action, or has proposed action, or is considering action, and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

Status of safety issues at licensed power plants

International Nuclear Information System (INIS)

1991-06-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG report will be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirement areas. This report, the third volume of a three-volume series, addresses the status of generic safety issues (GSIs) at licensed plants. Volume 1 addressed the status of Three Mile Island Action Plan requirements and was published in March 1991. Volume 2 addressed the status of implementation and verification of unresolved safety issues and was published in May 1991. The annual NUREG report will combine these three areas in a single volume to be published in late 1991. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. The purpose of this report is to provide a comprehensive description of the status of implementation and verification of the 34 GSIs and sub-issues that have been resolved by the NRC and involve implementation of an action or actions by licensees. This NUREG report also serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until a request for action by licensees is issued by NRC. 3 figs., 6 tabs

NRC regulatory agenda

International Nuclear Information System (INIS)

1993-07-01

The NRC Regulatory Agenda is a compilation of all rules on which the NRC has recently completed action, or has proposed action, or is considering action, and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter. The rules on which final action has been taken since March 31, 1993 are: Repeal of NRC standards of conduct; Fitness-for-duty requirements for licensees who possess, use, or transport Category I material; Training and qualification of nuclear power plant personnel; Monitoring the effectiveness of maintenance at nuclear power plants; Licensing requirements for land disposal of radioactive wastes; and Licensees' announcements of safeguards inspections

NRC Support for the Kalinin (VVER) probabilistic risk assessment

International Nuclear Information System (INIS)

Bley, D.; Diamond, D.J.; Chu, T.L.; Azarm, A.; Pratt, W.T.; Johnson, D.; Szukiewicz, A.; Drouin, M.; El-Bassioni, A.; Su, T.M.

1998-01-01

The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA

Implementation and interpretation of the NRC/IAEA transport regulations in the United States

International Nuclear Information System (INIS)

Shappert, L.B.

1987-01-01

For those materials used in the construction of spent fuel casks that are covered by the many available codes (ASME, ASTM, etc.), their acceptance by NRC can be expected to be straightforward. When materials are used that are not recognized by the codes, NRC has taken a more cautious attitude in their acceptance and approval. As mentioned earlier, they have initiated a number of studies involving ferritic steels, including DCI. More recently, they are organizing a panel of independent experts to consider all the data available and provide guidance as to whether these ferritic materials meet the safety requirements of the regulations and, if so, under what conditions. They will also be considering the testing that has been applied to casks fabricated from DCI and the likely margins of safety they possess. (orig./DG)

Status of NRC approval of EPRI electromagnetic interference susceptibility testing guidelines for digital equipment

International Nuclear Information System (INIS)

James, R.W.; Shank, J.W.; Yoder, C.

1996-01-01

Historically, nuclear power plants installing digital equipment have been required to conduct expensive, site-specific electromagnetic interference (EMI) surveys to demonstrate that EMI will not affect the operation of sensitive electronic equipment. Consequently, EPRI formed a Utility Working Group which developed a set of generic EMI susceptibility testing guidelines, which were published as an EPRI report in September 1994. These guidelines are based upon EMI survey data obtained from several different plants and include criteria for determining their applicability. The Working Group interacted with NRC staff to obtain NRC approval. In April 1996, the NRC issued a Safety Evaluation Report (SER) endorsing the guidelines as a valid means of demonstrating EMI compatibility. The issuance of this SER was conditional on issuing a revision to the EPRI EMI Guidelines. This paper summarizes the guidelines, the NRC SER, and the current status of Revision 1 to the report

Safety culture management: The importance of organizational factors

International Nuclear Information System (INIS)

Haber, S.B.; Shurberg, D.A.; Jacobs, R.; Hofmann, D.

1995-01-01

The concept of safety culture has been used extensively to explain the underlying causes of performance based events, both positive and negative, across the nuclear industry. The work described in this paper represents several years of effort to identify, define and assess the organizational factors important to safe performance in nuclear power plants (NPPs). The research discussed in this paper is primarily conducted in support of the US Nuclear Regulatory Commission's (NRC) efforts in understanding the impact of organizational performance on safety. As a result of a series of research activities undertaken by numerous NRC contractors, a collection of organizational dimensions has been identified and defined. These dimensions represent what is believed to be a comprehensive taxonomy of organizational elements that relate to the safe operation of nuclear power plants. Techniques were also developed by which to measure these organizational dimensions, and include structured interview protocols, behavioral checklists, and behavioral anchored rating scales (BARS). Recent efforts have focused on devising a methodology for the extraction of information related to the identified organizational dimensions from existing NRC documentation. This type of effort would assess the applicability of the organizational dimensions to existing NRC inspection and evaluation reports, refine the organizational dimensions previously developed so they are more relevant to the task of retrospective analysis, and attempt to rate plants based on the review of existing NRC documentation using the techniques previously developed for the assessment of organizational dimensions

Status of safety issues at licensed power plants: TMI Action Plan requirements, unresolved safety issues, generic safety issues, other multiplant action issues. Supplement 4

International Nuclear Information System (INIS)

1994-12-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. Supplement 3 gives status as of September 30, 1993. This annual report, Supplement 4, presents updated information as of September 30, 1994. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

Resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants

International Nuclear Information System (INIS)

Johnson, R.E.

1990-06-01

This report describes the US Nuclear Regulatory Commission's (NRC's) Generic Safety Issue 29, ''Bolting Degradation or Failure in Nuclear Power Plants,'' including the bases for establishing the issue and its historical highlights. The report also describes the activities of the Atomic Industrial Forum (AIF) relevant to this issue, including its cooperation with the Materials Properties Council (MPC) to organize a task group to help resolve the issue. The Electric Power Research Institute, supported by the AIF/MPC task group, prepared and issued a two-volume document that provides, in part, the technical basis for resolving Generic Safety Issue 29. This report presents the NRC's review and evaluation of the two-volume document and NRC's conclusion that this document, in conjunction with other information from both industry and NRC, provides the bases for resolving this issue

Status of safety issues at licensed power plants

International Nuclear Information System (INIS)

1991-05-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG report will be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirement areas. This report, the second volume of a three-volume series, addresses the status of unresolved safety issues (USIs) at licensed plants. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. The purpose of this report is to provide a comprehensive description of the status of implementation and verification of the 27 safety issues designated as USIs and to make this information available to other interested parties, including the public. A corollary purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants. 3 figs., 4 tabs

Status of safety issues at licensed power plants

International Nuclear Information System (INIS)

1991-03-01

As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG series report will be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirement areas. The data contained in this report are a product of the NRC's Safety Issues Management System database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by personnel in the NRC regions. This report has been prepared in order to provide a comprehensive description of the implementation and verification status of all the TMI Action Plan requirements at licensed reactors, and to make this information available to other interested parties, including the public. A corollary purpose of this report is for it to serve as a follow-on to NUREG-0933, ''A Prioritization of Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed facilities

NRC licensing speedup. Hearing before a Subcommittee of the Committee on Government Operations, House of Representatives, Ninety-Seventh Congress, First Session, June 18, 1981

International Nuclear Information System (INIS)

Anon.

1982-01-01

Joseph Hendrie, Acting Chairman of the Nuclear Regulatory Commission (NRC), was the principal witness at a June 18, 1981 hearing on whether the NRC is responsible for idling and delaying nuclear plants at a high cost to consumers. Congress is considering an interim licensing procedure that will allow plants to operate before safety hearings are completed. NRC evidence suggests that most plant delays are only projected ones, and that utilities must bear the blame for slippage in most construction schedules. Mr. Hendrie discussed the status of the nine impacted plants on the NRC list and outlined steps the Commission is taking to improve efficiency without impairing safety. His testimony and letters, statements, and other materials submitted for the record make up the hearing report

NRC/UBC Node

Energy Technology Data Exchange (ETDEWEB)

Ellis-Perry, B. [Univ. of British Columbia, Vancouver, British Columbia (Canada); Yogendran, Y. [NRC Inst. for Fuel Cell Innovation, Vancouver, British Columbia (Canada)

2004-07-01

'Full text:' In the search for cleaner, more sustainable energy sources, many of the most promising breakthroughs have been in hydrogen technology. However, this promise will remain unfulfilled without public interest and enthusiasm, and without the infrastructure to support the technology. In order to get there, we have to test, perfect, and demonstrate technology that is safe and affordable, and we must do so in practical, familiar settings. Ideally, such settings should be easily accessible to the engineers, planners, and architects of tomorrow while providing a showcase for hydrogen technology that will attract the general public. This place is the NRC/UBC Hydrogen Node. The UBC campus in Point Grey is home to leading edge, internationally recognized researchers in a range of disciplines, both within the University and at the NRC Institute for Fuel Cell Innovation. On average, 40,000 students, faculty, and staff use the campus every day; UBC graduates go on to leadership positions in communities around the globe. Its spectacular setting makes UBC a popular destination for thousands of visitors from around the world. In 2006 UBC will host the World Urban Forum, and in 2010 it will be one of the sites for the Vancouver-Whistler Olympic Games. UBC and its South Campus neighbourhoods are developing as a model sustainable community, offering an excellent opportunity to develop and showcase hydrogen infrastructure and technology in a real-life, attractive setting that will be seen by thousands of people around the world. UBC's facilities, location, and Trek 2010 commitment to excellence in learning, research, and sustainability make it an ideal location for such a project. The H2 Village at UBC will be an integrated hydrogen demonstration project, linked to the hydrogen highway. This project is bringing together leading companies, researchers, and government agencies committed to making the refinement and early adoption of safe hydrogen technology a

NRC/UBC Node

International Nuclear Information System (INIS)

Ellis-Perry, B.; Yogendran, Y.

2004-01-01

'Full text:' In the search for cleaner, more sustainable energy sources, many of the most promising breakthroughs have been in hydrogen technology. However, this promise will remain unfulfilled without public interest and enthusiasm, and without the infrastructure to support the technology. In order to get there, we have to test, perfect, and demonstrate technology that is safe and affordable, and we must do so in practical, familiar settings. Ideally, such settings should be easily accessible to the engineers, planners, and architects of tomorrow while providing a showcase for hydrogen technology that will attract the general public. This place is the NRC/UBC Hydrogen Node. The UBC campus in Point Grey is home to leading edge, internationally recognized researchers in a range of disciplines, both within the University and at the NRC Institute for Fuel Cell Innovation. On average, 40,000 students, faculty, and staff use the campus every day; UBC graduates go on to leadership positions in communities around the globe. Its spectacular setting makes UBC a popular destination for thousands of visitors from around the world. In 2006 UBC will host the World Urban Forum, and in 2010 it will be one of the sites for the Vancouver-Whistler Olympic Games. UBC and its South Campus neighbourhoods are developing as a model sustainable community, offering an excellent opportunity to develop and showcase hydrogen infrastructure and technology in a real-life, attractive setting that will be seen by thousands of people around the world. UBC's facilities, location, and Trek 2010 commitment to excellence in learning, research, and sustainability make it an ideal location for such a project. The H2 Village at UBC will be an integrated hydrogen demonstration project, linked to the hydrogen highway. This project is bringing together leading companies, researchers, and government agencies committed to making the refinement and early adoption of safe hydrogen technology a reality

76 FR 54986 - NRC Enforcement Policy

Science.gov (United States)

2011-09-06

... NUCLEAR REGULATORY COMMISSION 10 CFR Chapter I [NRC-2011-0209] NRC Enforcement Policy AGENCY: Nuclear Regulatory Commission. ACTION: Proposed enforcement policy revision; request for comment. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC or the Commission) is soliciting comments from interested...

76 FR 76192 - NRC Enforcement Policy

Science.gov (United States)

2011-12-06

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0273] NRC Enforcement Policy AGENCY: Nuclear Regulatory Commission. ACTION: Proposed enforcement policy revision; request for comment. SUMMARY: The U.S. Nuclear... licensees, vendors, and contractors), on proposed revisions to the NRC's Enforcement Policy (the Policy) and...

Systematic evaluation program review of NRC Safety Topic VI-7.3 associated with the electrical, instrumentation and control portions of the ECCS actuation system for the Dresden II Nuclear Power Plant

International Nuclear Information System (INIS)

St Leger-Barter, G.

1980-11-01

This report documents the technical evaluation and review of NRC Safety Topic VI-7.A.3, associated with the electrical, instrumentation, and control portions of the classification of the ECCS actuation system for the Dresden II nuclear power plant, using current licensing criteria

NRC licensing requirements: DOD options

International Nuclear Information System (INIS)

Pike, W.J.; O'Reilly, P.D.

1982-09-01

This report describes the licensing process (both safety and environmental) that would apply if the Department of Defense (DOD) chooses to obtain licenses from the US Nuclear Regulatory Commission (NRC) for using nuclear energy for power and luminous sources. The specific nuclear energy sources being considered include: small or medium-size nuclear power reactors; radioisotopic thermoelectric generators with 90 Sr or 238 Pu; radioisotopic dynamic electric generators with 90 Sr or 238 Pu; and applications of radioisotopes for luminous sources (lights) with 3 H, 85 Kr, or 147 Pm. The steps of the licensing process are summarized in the following sections, with particular attention given to the schedule and level of effort necessary to support the process

U.S. Nuclear Regulatory Commission Safety Culture Oversight

International Nuclear Information System (INIS)

Sieracki, D. J.

2016-01-01

The NRC recognises that it is important for all organizations performing or overseeing regulated activities to establish and maintain a positive safety culture commensurate with the safety and security significance of their activities and the nature and complexity of their organizations and functions. The NRC’s approach to safety culture is based on the premise that licencees bear the primary responsibility for safety. The NRC provides oversight of safety culture through expectations detailed in policy statements, safety culture assessor training for NRC inspectors, the oversight process, and the Allegations and Enforcement Programs. The NRC’s Safety Culture Policy Statement (SCPS) sets forth the Commission’s expectation that individuals and organizations establish and maintain a positive safety culture commensurate with the safety and security significance of their activities and the nature and complexity of their organizations and functions. The SCPS is not a regulation. It applies to all licencees, certificate holders, permit holders, authorisation holders, holders of quality assurance program approvals, vendors and suppliers of safety-related components, and applicants for a licence, certificate, permit, authorisation, or quality assurance program approval, subject to NRC authority.

Unresolved safety issues: where do we go from here

International Nuclear Information System (INIS)

Aycock, M.B.

1980-01-01

Section 210 of the Energy reorganization Act of 1974, as amended requires the NRC to develop a program for resolving Unresolved Safety Issues related to nuclear power plants. Seventeen Unresolved Safety Issues were identified by the NRC in 1978 and by early 1979 the NRC Unresolved Safety Issues Program was quickly becoming a well defined and manageable effort. Although, the Three Mile Island accident caused the momentum developed in early 1979 to be lost, efforts on ongoing generic tasks were continued by a special NRC Task Force established in June 1979. The momentum that was lost must be regained, however, if the Congressional mandate in Section 210 is to be met. With increased industry involvement and the marriage of the Unresolved Safety Issues Program with the improved and broader safety program development, audit and evaluation activities of the new NRR Division of Safety Technology, this should be possible

Psychology in nuclear power plants: an integrative approach to safety - general statement

International Nuclear Information System (INIS)

Shikiar, R.

1983-08-01

Since the accident at the Three Mile Island nuclear power plant on March 28, 1979, the commercial nuclear industry in the United States has paid increasing attention to the role of humans in overall plant safety. As the regulatory body with primary responsibility for ensuring public health and safety involving nuclear operations, the United States Nuclear Regulatory Commission (NRC) has also become increasingly involved with the ''human'' side of nuclear operations. The purpose of this symposium is to describe a major program of research and technical assistance that the Pacific Northwest Laboratory is performing for the NRC that deals with the issues of safety at nuclear power plants (NPPs). This program addresses safety from several different levels of analysis, which are all important within the context of an integrative approach to system safety

NRC antitrust licensing actions, 1978--1996

International Nuclear Information System (INIS)

Mayer, S.J.; Simpson, J.J.

1997-09-01

NUREG-0447, Antitrust Review of Nuclear Power Plants, was published in May 1978 and includes a compilation and discussion of U.S. Nuclear Regulatory Commission (NRC) proceedings and activity involving the NRC's competitive review program through February 1978, NUREG-0447 is an update of an earlier discussion of the NRC's antitrust review of nuclear power plants, NR-AIG-001, The US Nuclear Regulatory Commission's Antitrust Review of Nuclear Power Plants: The Conditioning of Licenses, which reviewed the Commission's antitrust review function from its inception in December 1970 through April 1976. This report summarizes the support provided to NRC staff in updating the compilation of the NRC's antitrust licensing review activities for commercial nuclear power plants that have occurred since February 1978. 4 refs., 4 tabs

Safety evaluation report related to the construction permit and operating license for the research reactor at the University of Texas (Docket No. 50-602)

International Nuclear Information System (INIS)

1992-01-01

The Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC) has prepared Supplement 1 to NUREG-1135, ''Safety Evaluation Report Related to the Construction Permit and Operating License for the Research Reactor at the University of Texas'' (SER) May 1985. The reactor facility is owned by The University of Texas at Austin (UT, the applicant) and is located at the University's Balcones Research Center in Austin, Texas. This supplement to the SER (SSER) describes the changes to the reactor facility design from the description in the SER. The SER and SSER together reflect the facility as built. The SSER also documents the reviews that the NRC has completed regarding the applicant's emergency plan, security plan, and technical specifications that were identified as open in the SER

NRC regulatory agenda: Quarterly report, April-June 1987

International Nuclear Information System (INIS)

1987-07-01

This compilation summarizes significant enforcement actions that have been resolved during one quarterly period and includes copies of letters, Notices, and Orders sent by the Nuclear Regulatory Commission to licensees with respect to these enforcement actions. It is anticipated that the information in this publication will be widely disseminated to managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoiding future violations similar to those described in this publication

Safety goals for commercial nuclear power plants

International Nuclear Information System (INIS)

Roe, J.W.

1988-01-01

In its official policy statement on safety goals for the operation of nuclear power plants, the Nuclear Regulatory Commission (NRC) set two qualitative goals, supported by two quantitative objectives. These goals are that (1) individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health; and (2) societal risks to life and health from nuclear power plant operation should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risks. As an alternative, this study proposes four quantitative safety goals for nuclear power plants. It begins with an analysis of the NRC's safety-goal development process, a key portion of which was devoted to delineating criteria for evaluating goal-development methods. Based on this analysis, recommendations for revision of the NRC's basic benchmarks for goal development are proposed. Using the revised criteria, NRC safety goals are evaluated, and the alternative safety goals are proposed. To further support these recommendations, both the NRC's goals and the proposed goals are compared with the results of three major probabilistic risk assessment studies. Finally, the potential impact of these recommendations on nuclear safety is described

A review of NRC staff uses of probabilistic risk assessment

Energy Technology Data Exchange (ETDEWEB)

1994-03-01

The NRC staff uses probabilistic risk assessment (PRA) and risk management as important elements its licensing and regulatory processes. In October 1991, the NRC`s Executive Director for Operations established the PRA Working Group to address concerns identified by the Advisory Committee on Reactor Safeguards with respect to unevenness and inconsistency in the staff`s current uses of PRA. After surveying current staff uses of PRA and identifying needed improvements, the Working Group defined a set of basic principles for staff PRA use and identified three areas for improvements: guidance development, training enhancements, and PRA methods development. For each area of improvement, the Working Group took certain actions and recommended additional work. The Working Group recommended integrating its work with other recent PRA-related activities the staff completed and improving staff interactions with PRA users in the nuclear industry. The Working Group took two key actions by developing general guidance for two uses of PRA within the NRC (that is, screening or prioritizing reactor safety issues and analyzing such issues in detail) and developing guidance on basic terms and methods important to the staff`s uses of PRA.

A review of NRC staff uses of probabilistic risk assessment

International Nuclear Information System (INIS)

1994-03-01

The NRC staff uses probabilistic risk assessment (PRA) and risk management as important elements its licensing and regulatory processes. In October 1991, the NRC's Executive Director for Operations established the PRA Working Group to address concerns identified by the Advisory Committee on Reactor Safeguards with respect to unevenness and inconsistency in the staff's current uses of PRA. After surveying current staff uses of PRA and identifying needed improvements, the Working Group defined a set of basic principles for staff PRA use and identified three areas for improvements: guidance development, training enhancements, and PRA methods development. For each area of improvement, the Working Group took certain actions and recommended additional work. The Working Group recommended integrating its work with other recent PRA-related activities the staff completed and improving staff interactions with PRA users in the nuclear industry. The Working Group took two key actions by developing general guidance for two uses of PRA within the NRC (that is, screening or prioritizing reactor safety issues and analyzing such issues in detail) and developing guidance on basic terms and methods important to the staff's uses of PRA

NRC Information No. 89-89: Event notification worksheets

International Nuclear Information System (INIS)

Rossi, C.E.

1992-01-01

The NRC ''Event Notification Worksheet,'' NRC Form 361, has been revised to assist the NRC Headquarters Operations Officers in obtaining adequate information for evaluation of significant events reported to the NRC Operations Center. The new forms more accurately reflect the event classifications and the 10 CFR 50.72 categories that must be reported. A copy of the new worksheet is enclosed for your reference. NRC Form 361 can be ordered from the NRC Information and Records Management Branch

NRC antitrust licensing actions, 1978--1996

Energy Technology Data Exchange (ETDEWEB)

Mayer, S.J.; Simpson, J.J.

1997-09-01

NUREG-0447, Antitrust Review of Nuclear Power Plants, was published in May 1978 and includes a compilation and discussion of U.S. Nuclear Regulatory Commission (NRC) proceedings and activity involving the NRC`s competitive review program through February 1978, NUREG-0447 is an update of an earlier discussion of the NRC`s antitrust review of nuclear power plants, NR-AIG-001, The US Nuclear Regulatory Commission`s Antitrust Review of Nuclear Power Plants: The Conditioning of Licenses, which reviewed the Commission`s antitrust review function from its inception in December 1970 through April 1976. This report summarizes the support provided to NRC staff in updating the compilation of the NRC`s antitrust licensing review activities for commercial nuclear power plants that have occurred since February 1978. 4 refs., 4 tabs.

Safety of Research Reactors. Safety Requirements

International Nuclear Information System (INIS)

2010-01-01

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

77 FR 30332 - Mr. James Chaisson; Order Prohibiting Involvement in NRC-Licensed Activities

Science.gov (United States)

2012-05-22

... industrial radiographic operations in accordance with conditions specified therein. The license was... NRC conducted a safety and security inspection of the use of byproduct material for industrial... requirements. Mr. Chaisson chose to store a radiographic exposure device at a leased facility on Elk Street in...

The role of research in nuclear regulation: Opening remarks

International Nuclear Information System (INIS)

Taylor, J.M.

1997-01-01

More than 20 years ago, the Energy Reorganization Act of 1974 created the USNRC and that same act provided for an office of nuclear regulatory research. It's what is called a statutory office within the NRC. In providing for an NRC research program, our Congress had several things to say about the character of the research that would be performed. First, NRC should perform such research as is necessary for the effective performance of the Commission's licensing and related regulatory functions. Second, the research may be characterized as confirmatory reassessment related to the safe operation and the protection of commercial reactors and other nuclear materials. Third, the NRC should have an independent capability for developing and analyzing technical information related to reactor safety, safeguards, and environmental protection in support of both the licensing and regulatory processes. Fourth, the research should not go beyond the need for confirmatory assessment, because the NRC should never be place in a position of having generated and then having to defend basic design data of its own. This has been and continues to be the role of research at the NRC. Somewhat different purposes might apply for regulatory agencies in other countries. Several regulatory agencies are represented here on this panel, so some of these difference may be discussed

The evolution of the structure and application of U.S. NRC regulations and standards

International Nuclear Information System (INIS)

Murley, T.E.; Rosztoczy, Z.R.; McPherson, G.D.

1991-01-01

NRC regulations and standards and their implementation have evolved from early adaptations of conventional engineering practices to a mature, cohesive set of regulations that govern NRC regulation of nuclear power plant safety in the United States. From a simple set of rules and design criteria and from the standards of the professional engineering societies, a hierarchy of practices, standards, guides, rules and goals has developed. Resting on a foundation of industrial practices, this hierarchy rises through levels of national standards, regulatory guides and standard review plans, policy statements and NRC regulations. The licensing process is evolving today toward one that permits both site approval and standard design certification before the plant is constructed. At the present time, NRC is reviewing five standard designs for certification for a period of 15 years. NRC focuses its regulation of operating nuclear plants on inspections conducted from five regional offices. Resident inspectors, specialist inspectors, and multi-disciplinary inspection teams examine specific plant situations. The results of all these inspections are used to develop a complete understanding of a plant's physical condition, its operation, maintenance and management. To improve safe operation of nuclear plants in the U.S., a most important program, the Systematic Assessment of Licensee Performance, measures operational performance, using a broad spectrum of functional areas. (orig.)

NRC's license renewal regulations

International Nuclear Information System (INIS)

Akstulewicz, Francis

1991-01-01

In order to provide for the continuity of the current generation of nuclear power plant operating licenses and at the same time ensure the health and safety of the public, and the quality of the environment, the US Nuclear Regulatory Commission (NRC) established a goal of developing and issuing regulations and regulatory guidance for license renewal in the early 1990s. This paper will discuss some of those activities underway to achieve this goal. More specifically, this paper will discuss the Commission's regulatory philosophy for license renewal and the two major license renewal rule makings currently underway. The first is the development of a new Part 54 to address procedural and technical requirements for license renewal; the second is a revision to existing Part 51 to exclude environmental issues and impacts from consideration during the license renewal process. (author)

RSAS: a Reactor Safety Assessment System

International Nuclear Information System (INIS)

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

1985-01-01

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

Small Business Innovation Research Program. Program solicitation FY 1984

International Nuclear Information System (INIS)

1984-01-01

The Nuclear Regulatory Commission (NRC) invites science-based and high-technology small business firms to submit research proposals under this program solicitation entitled Small Business Innovation Research (SBIR). Firms with strong research capabilities in science or engineering in any of the following topic areas are encouraged to participate. NRC will support high-quality research proposals on important scientific or engineering problems and opportunities that could lead to significant advancement in the safety of nuclear operations or nuclear power plants. Objectives of the solicitation include stimulating technological innovation in the private sector, strengthening the role of small business in meeting Federal research and development needs, increasing the commercial application of NRC-supported research results, and improving the return on investment from Federally funded research for economic and social benefits to the Nation

75 FR 60485 - NRC Enforcement Policy Revision

Science.gov (United States)

2010-09-30

... NUCLEAR REGULATORY COMMISSION [NRC-2008-0497] NRC Enforcement Policy Revision AGENCY: Nuclear Regulatory Commission. ACTION: Policy statement. SUMMARY: The Nuclear Regulatory Commission (NRC or Commission) is publishing a major revision to its Enforcement Policy (Enforcement Policy or Policy) to...

Methods and criteria for safety analysis (FIN L2535)

International Nuclear Information System (INIS)

1992-12-01

In response to the NRC request for a proposal dated October 20, 1992, Westinghouse Savannah River Company (WSRC) submit this proposal to provide contractural assistance for FIN L2535, ''Methods and Criteria for Safety Analysis,'' as specified in the Statement of Work attached to the request for proposal. The Statement of Work involves development of safety analysis guidance for NRC licensees, arranging a workshop on this guidance, and revising NRC Regulatory Guide 3.52. This response to the request for proposal offers for consideration the following advantages of WSRC in performing this work: Experience, Qualification of Personnel and Resource Commitment, Technical and Organizational Approach, Mobilization Plan, Key Personnel and Resumes. In addition, attached are the following items required by the NRC: Schedule II, Savannah River Site - Job Cost Estimate, NRC Form 189, Project and Budget Proposal for NRC Work, page 1, NRC Form 189, Project and Budget Proposal for NRC Work, page 2, Project Description

Computerized based training in nuclear safety in the nuclear research center Negev

International Nuclear Information System (INIS)

Ben-Shachar, B.; Krubain, H.; Sberlo, E.

2002-01-01

The Department of Human Resources and Training in the Nuclear Research Center, Negev, in collaboration with the Department of Radiation Protection and Safety used to organize different kinds of training and refresher courses for different aspects of safety in nuclear centers (radiation safety, biological effects of ionizing radiation, industrial safety, fire fighting, emergency procedures, etc.). All radiation workers received a training program of several days in all these subjects, each year. The administrative employees received a shorter training, each second year. The training included only frontal lectures and no quiz or exams were done. No feedback of the employees was received after the training, as well. Recently, a new training program was developed by the NRC-Negev and the CET (Center for Educational Technology), in order to perform the refresher courses. The training includes CBT-s (Computer Based Training), e.g. tutorials and quiz. The tutorial is an interactive course in one subject, including animations, video films and photo stills. The employee gets a simple and clear explanation (including pictures). After each tutorial there is a quiz which includes 7 American style questions. In the following lecture different parts from two of the tutorials used for the refresher courses, will be presented

NRC/UBC fuelling station with intelligent compression

International Nuclear Information System (INIS)

Dada, A.; Boyd, B.; Law, L.; Semczyszyn, D.

2004-01-01

BOC Canada Ltd. will design, integrate and construct the second fueling station on the Hydrogen Highway. This station will be located at the National Research Council's Institute for Fuel Cell Innovation on the campus of the University of British Columbia. BOC's design will bring together an existing alkaline electrolyser, new compression, storage and dispensing. The station will be designed to serve fuel cell passenger vehicles using 350-bar storage. However, the flexible design concept will allow for many other user needs including the potential for servicing larger vehicles, as well as filling portable storage systems for use at satellite stations. The novel station design also offers the potential to fuel from multiple hydrogen sources. Together with NRC, this fueling station will be used to increase public, consumer and investor awareness of hydrogen technologies. Design and construction of this facility will assist in the development of industry codes and standards and familiarize authorities having jurisdiction with hydrogen fueling. The system concept offers the utmost attention to safety, novelty and flexibility. (author)

Reactor safety: a discussion by officials of the Nuclear Regulatory Commission

International Nuclear Information System (INIS)

Anders, W.A.; Rusche, B.C.; Stello, V. Jr.; Minogue, R.B.

1976-01-01

William A. Anders, Chairman of the Nuclear Regulatory Commission (NRC), and several senior officials spoke to the Joint Committee on Atomic Energy on the subject of nuclear safety, improvements in reactor plant safety, and quality assurance. The NRC, during its first year of organization, has developed new initiatives to improve safety and safeguards regulations. Anders stressed that NRC is not stifling internal discussion of opposing views, that it has been honest with the public, and that operating reactors are meeting rigorous safety standards. Other speakers discussed comparative safety of old and new reactors. Backfitting of older plants with new features is done when substantial safety protection can be added, but detuning an integrated system is not done indiscriminately. Officials of NRC do not agree with former General Electric employees, who testified that the regulatory procedure is inadequate. Safety improvements since August 28, 1962 and outlines of the review process are included in the Appendixes

US nuclear safety review and experience

International Nuclear Information System (INIS)

Gilinsky, V.

1977-01-01

The nuclear safety review of commercial nuclear power reactors has changed over the years from the relatively simple review of Dresden 1 in 1955 to the highly complex and sophisticated regulatory process which characterizes today's reviews. Four factors have influenced this evolution: (1) maturing of the technology and industry; (2) development of the regulatory process and associated staff; (3) feedback of operating experience; and (4) public awareness and participation. The NRC's safety review responsibilities start before an application is tendered and end when the plant is decommissioned. The safety review for reactor licensing is a comprehensive, two-phase process designed to assure that all the established conservative acceptance criteria are satisfied. Operational safety is assured through a strong inspection and enforcement program which includes shutting down operating facilities when necessary to protect the health and safety of the public. The safety of operating reactors is further insured through close regulation of license changes and selective backfitting of new regulatory requirements. An effective NRC standards development program has been implemented and coordinates closely with the national standards program. A confirmatory safety research program has been developed. Both of these efforts are invaluable to the nuclear safety review because they provide the staff with key tools needed to carry out its regulatory responsibilities. Both have been given increased emphasis since the formation of the NRC in 1975. The safety review process will continue to evolve, but changes will be slower and more deliberate. It will be influenced by standardization, early site reviews and development of advanced reactor concepts. New legislation may make possible changes which will simplify and shorten the regulatory process. Certainly the experience provided by the increasing number and types of operating plants will have a very strong impact on future trends in the

Assessment of the NRC Enforcement Program

International Nuclear Information System (INIS)

Lieberman, J.; Coblentz, L.

1995-04-01

On May 12, 1994, the Executive Director for Operations (EDO) established a Review Team composed of senior NRC managers to re-examine the NRC enforcement program. A copy of the Review Team's charter is enclosed as Appendix A. This report presents the Team's assessment. The purpose of this review effort are: (1) to perform an assessment of the NRC's enforcement program to determine whether the defined purposes of the enforcement program are appropriate; (2) to determine whether the NRC's enforcement practices and procedures for issuing enforcement actions are consistent with those purposes; and (3) to provide recommendations on any changes the Review Team believes advisable. In accordance with its charter, the Review Team considered the following principal issues in conducting its assessment of the enforcement program: the balance between providing deterrence and incentives (both positive and negative) for the identification and correction of violations; the appropriateness of NRC sanctions; whether the commission should seek statutory authority to increase the amount of civil penalties; whether the NRC should use different enforcement policies and practices for different licensees (e.g., materials licensees in contrast to power reactors or large fuel facilities); and whether the commission should establish open enforcement conferences as the normal practice

NRC [Nuclear Regulatory Commission] staff evaluation of the General Electric Company Nuclear Reactor Study (''Reed Report'')

International Nuclear Information System (INIS)

1987-07-01

In 1975, the General Electric Company (GE) published a Nuclear Reactor Study, also referred to as ''the Reed Report,'' an internal product-improvement study. GE considered the document ''proprietary'' and thus, under the regulations of the Nuclear Regulatory Commission (NRC), exempt from mandatory public disclosure. Nonetheless, members of the NRC staff reviewed the document in 1976 and determined that it did not raise any significant new safety issues. The staff also reached the same conclusion in subsequent reviews. However, in response to recent inquiries about the report, the staff reevaluated the Reed Report from a 1987 perspective. This re-evaluation, documented in this staff report, concluded that: (1) there are no issues raised in the Reed Report that support a need to curtail the operation of any GE boiling water reactor (BWR); (2) there are no new safety issues raised in the Reed Report of which the staff was unaware; and (3) although certain issues addressed by the Reed Report are still being studied by the NRC and the industry, there is no basis for suspending licensing and operation of GE BWR plants while these issues are being resolved

Impacts of NRC programs on state and local governments

International Nuclear Information System (INIS)

Nussbaumer, D.A.; Lubenau, J.O.

1983-12-01

This document reports the results of an NRC staff examination of the impacts of NRC regulatory programs on State and local governments. Twenty NRC programs are identified. For each, the source of the program (e.g., statutory requirement) and NRC funding availability are described and the impacts upon State and local governments are assessed. Recommendations for NRC monitoring and assessing impacts and for enhancing NRC staff awareness of the impacts are offered

The NRC and utility finances

International Nuclear Information System (INIS)

Byus, L.C.

1992-01-01

In a speech before the National Association of Regulatory Utility Commissioners in November 1991, Nuclear Regulatory Commission Chairman Ivan Selin presented what he called an open-quotes expansion of the concept of safety beyond our previous narrow bounds.close quotes He went on to explain, open-quotes To be seen as successful and safe operators of nuclear facilities, utilities must have safe and predictable cash flows.close quotes While there is little disagreement that the concepts of successful plant operations and financial strength go hand in hand, the relationship between the two is not clear. Which came first, successful operation of generating plants or financial strength? Selin's views on NRC involvement in financial aspects of utility operation in the United States are sure to stimulate debate on the issue

NRC [Nuclear Regulatory Commission] perspective of software QA [quality assurance] in the nuclear history

International Nuclear Information System (INIS)

Weiss, S.H.

1988-01-01

Computer technology has been a part of the nuclear industry since its inception. However, it is only recently that computers have been integrated into reactor operations. During the early history of commercial nuclear power in the United States, the US Nuclear Regulatory Commission (NRC) discouraged the use of digital computers for real-time control and monitoring of nuclear power plant operation. At the time, this position was justified since software engineering was in its infancy, and horror stories on computer crashes were plentiful. Since the advent of microprocessors and inexpensive computer memories, significant advances have been made in fault-tolerant computer architecture that have resulted in highly reliable, durable computer systems. The NRC's requirement for safety parameter display system (SPDS) stemmed form the results of studies and investigations conducted on the Three Mile Island Unit 2 (TMI-2) accident. An NRC contractor has prepared a handbook of software QA techniques applicable to the nuclear industry, published as NUREG/CR-4640 in August 1987. Currently, the NRC is considering development of an inspection program covering software QA. Future efforts may address verification and validation as applied to expert systems and artificial intelligence programs

10 CFR 2.709 - Discovery against NRC staff.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Discovery against NRC staff. 2.709 Section 2.709 Energy... Rules for Formal Adjudications § 2.709 Discovery against NRC staff. (a)(1) In a proceeding in which the NRC staff is a party, the NRC staff will make available one or more witnesses, designated by the...

Nuclear regulation. NRC's security clearance program can be strengthened

International Nuclear Information System (INIS)

Fultz, Keith O.; Kruslicky, Mary Ann; Bagnulo, John E.

1988-12-01

Because of the national security implications of its programs, the Nuclear Regulatory Commission (NRC) investigates the background of its employees and consultants as well as others to ensure that they are reliable and trustworthy. If the investigation indicates that an employee will not endanger national security, NRC grants a security clearance that allows access to classified information, material, and facilities. NRC also requires periodic checks for some clearance holders to ensure their continued clearance eligibility. The Chairman, Subcommittee on Environment, Energy, and Natural Resources, House Committee on Government Operations, asked GAO to review NRC's personnel security clearance program and assess the procedures that NRC uses to ensure that those who operate nuclear power plants do not pose a threat to the public. The Atomic Energy Act of 1954 requires NRC to conduct background investigations of its employees and consultants as well as others who have access to classified information, material, or facilities. To do this, NRC established a personnel security clearance program. Under NRC policies, a security clearance is granted after the Office of Personnel Management (OPM) or the Federal Bureau of Investigation checks the background of those applying for an NRC clearance. NRC also periodically reassesses the integrity of those holding the highest level clearance. NRC employees, consultants, contractors, and licensees as well as other federal employees hold approximately 10,600 NRC clearances. NRC does not grant clearances to commercial nuclear utility employees unless they require access to classified information or special nuclear material. However, the utilities have voluntarily established screening programs to ensure that their employees do not pose a threat to nuclear plants. NRC faces a dilemma when it hires new employees. Although its policy calls for new hires to be cleared before they start work, the security clearance process takes so long

Evaluation of reliability assurance approaches to operational nuclear safety

International Nuclear Information System (INIS)

Mueller, C.J.; Bezella, W.A.

1984-01-01

This report discusses the results of research to evaluate existing and/or recommended safety/reliability assurance activities among nuclear and other high technology industries for potential nuclear industry implementation. Since the Three Mile Island (TMI) accident, there has been increased interest in the use of reliability programs (RP) to assure the performance of nuclear safety systems throughout the plant's lifetime. Recently, several Nuclear Regulatory Commission (NRC) task forces or safety issue review groups have recommended RPs for assuring the continuing safety of nuclear reactor plants. 18 references

Systematic evaluation program review of NRC safety topic VII-2 associated with the electrical, instrumentation and control portions of the ESF system control logic and design for the Dresden Station, Unit II nuclear power plant

International Nuclear Information System (INIS)

St Leger-Barter, G.

1980-11-01

This report documents the technical evaluation and review of NRC Safety Topic VII-2, associated with the electrical, instrumentation, and control portions of the ESF system control logic and design for the Dresden Station Unit II nuclear power plant, using current licensing criteria

Regulatory instrument review: Management of aging of LWR [light water reactor] major safety-related components

International Nuclear Information System (INIS)

Werry, E.V.

1990-10-01

This report comprises Volume 1 of a review of US nuclear plant regulatory instruments to determine the amount and kind of information they contain on managing the aging of safety-related components in US nuclear power plants. The review was conducted for the US Nuclear Regulatory Commission (NRC) by the Pacific Northwest Laboratory (PNL) under the NRC Nuclear Plant Aging Research (NPAR) Program. Eight selected regulatory instruments, e.g., NRC Regulatory Guides and the Code of Federal Regulations, were reviewed for safety-related information on five selected components: reactor pressure vessels, steam generators, primary piping, pressurizers, and emergency diesel generators. Volume 2 will be concluded in FY 1991 and will also cover selected major safety-related components, e.g., pumps, valves and cables. The focus of the review was on 26 NPAR-defined safety-related aging issues, including examination, inspection, and maintenance and repair; excessive/harsh testing; and irradiation embrittlement. The major conclusion of the review is that safety-related regulatory instruments do provide implicit guidance for aging management, but include little explicit guidance. The major recommendation is that the instruments be revised or augmented to explicitly address the management of aging

Development of the NRC`s Human Performance Investigation Process (HPIP). Volume 3, Development documentation

Energy Technology Data Exchange (ETDEWEB)

Paradies, M.; Unger, L. [System Improvements, Inc., Knoxville, TN (United States); Haas, P.; Terranova, M. [Concord Associates, Inc., Knoxville, TN (United States)

1993-10-01

The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause events at nuclear power plants. This document, Volume III, is a detailed documentation of the development effort and the pilot training program.

NRC systematic evaluation program: seismic review

International Nuclear Information System (INIS)

Levin, H.A.

1980-01-01

The NRC Systematic Evaluation Program is currently making an assessment of the seismic design safety of 11 older nuclear power plant facilities. The general review philosophy and review criteria relative to seismic input, structural response, and equipment functionability are presented, including the rationale for the development of these guidelines considering the significant evolution of seismic design criteria since these plants were originally licensed. Technical approaches thought more realistic in light of current knowledge are utilized. Initial findings for plants designed to early seismic design procedures suggest that with minor exceptions, these plants possess adequate seismic design margins when evaluated against the intent of current criteria. However, seismic qualification of electrical equipment has been identified as a subject which requires more in-depth evaluation

NRC Seismic Design Margins Program Plan

International Nuclear Information System (INIS)

Cummings, G.E.; Johnson, J.J.; Budnitz, R.J.

1985-08-01

Recent studies estimate that seismically induced core melt comes mainly from earthquakes in the peak ground acceleration range from 2 to 4 times the safe shutdown earthquake (SSE) acceleration used in plant design. However, from the licensing perspective of the US Nuclear Regulatory Commission, there is a continuing need for consideration of the inherent quantitative seismic margins because of, among other things, the changing perceptions of the seismic hazard. This paper discusses a Seismic Design Margins Program Plan, developed under the auspices of the US NRC, that provides the technical basis for assessing the significance of design margins in terms of overall plant safety. The Plan will also identify potential weaknesses that might have to be addressed, and will recommend technical methods for assessing margins at existing plants. For the purposes of this program, a general definition of seismic design margin is expressed in terms of how much larger that the design basis earthquake an earthquake must be to compromise plant safety. In this context, margin needs to be determined at the plant, system/function, structure, and component levels. 14 refs., 1 fig

Fuel safety research 1999

Energy Technology Data Exchange (ETDEWEB)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-07-01

In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

NRC TLD Direct Radiation Monitoring Network

International Nuclear Information System (INIS)

Struckmeyer, R.; McNamara, N.

1991-12-01

This report provides the status and results of the NRC Thermoluminescent Dosimeter (TLD) Direct Radiation Monitoring Network. It presents the radiation levels measured in the vicinity of NRC licensed facilities throughout the country for the third quarter of 1991

Comparisons of ANSI standards cited in the NRC standard review plan, NUREG-0800 and related documents

International Nuclear Information System (INIS)

Ankrum, A.R.; Bohlander, K.L.; Gilbert, E.R.; Pawlowski, R.A.; Spiesman, J.B.

1995-11-01

This report provides the results of comparisons of the cited and latest versions of ANSI standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC's Standard Review Plan Update and Development Program. Significant changes to the standards, from the cited version to the latest version, are described and discussed in a tabular format for each standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the standard. The results and recommendations presented in this document have not been subjected to NRC staff review

Comparisons of ASTM standards cited in the NRC standard review plan, NUREG-0800 and related documents

International Nuclear Information System (INIS)

Ankrum, A.R.; Bohlander, K.L.; Gilbert, E.R.; Pawlowski, R.A.; Spiesman, J.B.

1995-10-01

This report provides the results of comparisons of the cited and latest versions of ASTM standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC's Standard Review Plan Update and Development Program. Significant changes to the standards, from the cited version to the latest version, are described and discussed in a tabular format for each standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the standard. The results and recommendations presented in this document have not been subjected to NRC staff review

Status of ECCS Acceptance Criteria Revision in U. S. NRC and Perspective

International Nuclear Information System (INIS)

Lee, Joosuk; Woo, Swengwoong

2013-01-01

Subsequently the NRC had conducted a fuel cladding research program to investigate the behavior of high burnup fuel cladding under LOCA conditions. This research program conducted at Argonne National Laboratory (ANL), as well as conducted as jointly-funded programs at the Kurchatov Institute and the Halden Reactor. From these programs, several important technical findings, listed in following section, for rule revision were obtained. On March 14, 2000, the Nuclear Energy Institute (NEI) submitted a petition for rulemaking (PRM) requesting that the NRC amend its regulations in 50.44 and 50.46 (PRM 50-71). The NEI petition stated that these regulations apply to only two specific zirconium-alloy fuel cladding materials (zircaloy and ZIRLO TM ). The NRC resolved PRM-50-71 by deciding that it should be considered in the following rulemaking process. Meanwhile, on March 15, 2007, Mark Leyse submitted a PRM to the NRC (PRM 50-84). The petitioner requests that the NRC conduct rulemaking in the following areas: Establish regulations that require licensees to operate light-water power reactors under conditions that are effective in limiting the thickness of crud and/or oxide layers. Amend Appendix K to Part 50 to explicitly require that steady-state temperature distribution and stored energy in the reactor fuel at the onset of a postulated LOCA be calculated by factoring in the role of the thermal resistance of crud and oxide layers. Amend 50.46 to specify a maximum allowable percentage of hydrogen content in the cladding. The NRC resolved PRM-50-84 by deciding that the petitioner's issues should be considered in the rulemaking process. On August 13, 2009, the NRC published an Advance Notice of Proposed Rulemaking (ANPR) to obtain stakeholder views on issues associated with amending 50.46. On March 1, 2012, and subsequently modified by the staff's June 1, 2012, SECY-12-0034 was submitted to the Commission to obtain approval to publish for public comment, and it was approved

Status of ECCS Acceptance Criteria Revision in U. S. NRC and Perspective

Energy Technology Data Exchange (ETDEWEB)

Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

Subsequently the NRC had conducted a fuel cladding research program to investigate the behavior of high burnup fuel cladding under LOCA conditions. This research program conducted at Argonne National Laboratory (ANL), as well as conducted as jointly-funded programs at the Kurchatov Institute and the Halden Reactor. From these programs, several important technical findings, listed in following section, for rule revision were obtained. On March 14, 2000, the Nuclear Energy Institute (NEI) submitted a petition for rulemaking (PRM) requesting that the NRC amend its regulations in 50.44 and 50.46 (PRM 50-71). The NEI petition stated that these regulations apply to only two specific zirconium-alloy fuel cladding materials (zircaloy and ZIRLO{sup TM}). The NRC resolved PRM-50-71 by deciding that it should be considered in the following rulemaking process. Meanwhile, on March 15, 2007, Mark Leyse submitted a PRM to the NRC (PRM 50-84). The petitioner requests that the NRC conduct rulemaking in the following areas: Establish regulations that require licensees to operate light-water power reactors under conditions that are effective in limiting the thickness of crud and/or oxide layers. Amend Appendix K to Part 50 to explicitly require that steady-state temperature distribution and stored energy in the reactor fuel at the onset of a postulated LOCA be calculated by factoring in the role of the thermal resistance of crud and oxide layers. Amend 50.46 to specify a maximum allowable percentage of hydrogen content in the cladding. The NRC resolved PRM-50-84 by deciding that the petitioner's issues should be considered in the rulemaking process. On August 13, 2009, the NRC published an Advance Notice of Proposed Rulemaking (ANPR) to obtain stakeholder views on issues associated with amending 50.46. On March 1, 2012, and subsequently modified by the staff's June 1, 2012, SECY-12-0034 was submitted to the Commission to obtain approval to publish for public comment, and it

Transactions of the twenty-fifth water reactor safety information meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session.

Transactions of the twenty-fifth water reactor safety information meeting

International Nuclear Information System (INIS)

Monteleone, S.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session

EPA Response to the National Research Council (NRC) Report - A Review of the Technical Basis of the Chemical and Pathogen Regulations for Biosolids

Science.gov (United States)

The NRC published two reports over a twenty year span. They concluded that there is no documented scientific evidence that sewage sludge regulations have failed to protect public health, but identified areas of research.

78 FR 5838 - NRC Enforcement Policy

Science.gov (United States)

2013-01-28

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0014] NRC Enforcement Policy AGENCY: Nuclear Regulatory Commission. ACTION: Policy revision; issuance and request for comments. SUMMARY: The U.S. Nuclear Regulatory... Nuclear Regulatory Commission Enforcement Policy,'' December 30, 2009 (ADAMS Accession No. ML093200520);(2...

Transactions of the Twenty-First Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Monteleone, S.

1993-10-01

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

Transactions of the Twenty-First Water Reactor Safety Information Meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1993-10-01

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

NRC overview: Repository QA

International Nuclear Information System (INIS)

Kennedy, J.E.

1988-01-01

The US Department of Energy (DOE) is on the threshold of an extensive program for characterizing Yucca Mountain in Nevada to determine if it is a suitable site for the permanent disposal of high-level nuclear waste. Earlier this year, the DOE published the Consultation Draft Site Characterization Plan for the Nevada site, which describes in some detail the studies that need to be performed to determine if the site is acceptable. In the near future, the final site characterization plan (SCP) is expected to be issued and large-scale site characterization activities to begin. The data and analyses that will result from the execution of that plan are expected to be the primary basis for the license application to the US Nuclear Regulatory Commission (NRC). Because of the importance of these data and analyses in the assessment of the suitability of the site and in the demonstration of that suitability in the NRC licensing process, the NRC requires in 10CFR60 that site characterization be performed under a quality assurance (QA) program. The QA program is designed to provide confidence that data are valid, retrievable, and reproducible. The documentation produced by the program will form an important part of the record on which the suitability of the site is judged in licensing. In addition, because the NRC staff can review only a selected portion of the data collected, the staff will need to rely on the system of controls in the DOE QA program

Trends in HFE Methods and Tools and Their Applicability to Safety Reviews

Energy Technology Data Exchange (ETDEWEB)

O' Hara, J.M.; Plott, C.; Milanski, J.; Ronan, A.; Scheff, S.; Laux, L.; and Bzostek, J.

2009-09-30

The U.S. Nuclear Regulatory Commission's (NRC) conducts human factors engineering (HFE) safety reviews of applicant submittals for new plants and for changes to existing plants. The reviews include the evaluation of the methods and tools (M&T) used by applicants as part of their HFE program. The technology used to perform HFE activities has been rapidly evolving, resulting in a whole new generation of HFE M&Ts. The objectives of this research were to identify the current trends in HFE methods and tools, determine their applicability to NRC safety reviews, and identify topics for which the NRC may need additional guidance to support the NRC's safety reviews. We conducted a survey that identified over 100 new HFE M&Ts. The M&Ts were assessed to identify general trends. Seven trends were identified: Computer Applications for Performing Traditional Analyses, Computer-Aided Design, Integration of HFE Methods and Tools, Rapid Development Engineering, Analysis of Cognitive Tasks, Use of Virtual Environments and Visualizations, and Application of Human Performance Models. We assessed each trend to determine its applicability to the NRC's review by considering (1) whether the nuclear industry is making use of M&Ts for each trend, and (2) whether M&Ts reflecting the trend can be reviewed using the current design review guidance. We concluded that M&T trends that are applicable to the commercial nuclear industry and are expected to impact safety reviews may be considered for review guidance development. Three trends fell into this category: Analysis of Cognitive Tasks, Use of Virtual Environments and Visualizations, and Application of Human Performance Models. The other trends do not need to be addressed at this time.

Trends in HFE Methods and Tools and Their Applicability to Safety Reviews

International Nuclear Information System (INIS)

O'Hara, J.M.; Plott, C.; Milanski, J.; Ronan, A.; Scheff, S.; Laux, L.; Bzostek, J.

2009-01-01

The U.S. Nuclear Regulatory Commission's (NRC) conducts human factors engineering (HFE) safety reviews of applicant submittals for new plants and for changes to existing plants. The reviews include the evaluation of the methods and tools (M and T) used by applicants as part of their HFE program. The technology used to perform HFE activities has been rapidly evolving, resulting in a whole new generation of HFE M and Ts. The objectives of this research were to identify the current trends in HFE methods and tools, determine their applicability to NRC safety reviews, and identify topics for which the NRC may need additional guidance to support the NRC's safety reviews. We conducted a survey that identified over 100 new HFE M and Ts. The M and Ts were assessed to identify general trends. Seven trends were identified: Computer Applications for Performing Traditional Analyses, Computer-Aided Design, Integration of HFE Methods and Tools, Rapid Development Engineering, Analysis of Cognitive Tasks, Use of Virtual Environments and Visualizations, and Application of Human Performance Models. We assessed each trend to determine its applicability to the NRC's review by considering (1) whether the nuclear industry is making use of M and Ts for each trend, and (2) whether M and Ts reflecting the trend can be reviewed using the current design review guidance. We concluded that M and T trends that are applicable to the commercial nuclear industry and are expected to impact safety reviews may be considered for review guidance development. Three trends fell into this category: Analysis of Cognitive Tasks, Use of Virtual Environments and Visualizations, and Application of Human Performance Models. The other trends do not need to be addressed at this time.

10 CFR 51.40 - Consultation with NRC staff.

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Consultation with NRC staff. 51.40 Section 51.40 Energy....40 Consultation with NRC staff. (a) A prospective applicant or petitioner for rulemaking is encouraged to confer with NRC staff as early as possible in its planning process before submitting...

Implementation of safety goals in NRC's regulatory process

International Nuclear Information System (INIS)

Murley, T.E.

1985-01-01

In May 1983 the Nuclear Regulatory Commission issued a policy statement on Safety Goals For Nuclear Power Plant Operation. The Commission at the same time judged that a two-year evaluation period was necessary to judge the effectiveness of the goals and design objectives, and directed the staff to develop information and understanding as to how to further define and use the design objectives and the cost-benefit guidelines. In carrying out the Commission's mandate, the staff framed three major questions to be addressed during the safety goal evaluation period. These three questions are: 1) to what extent is it practical to use safety goals in the regulatory process. 2) Should the quantitative design objectives be modified or supplemented. If so, how. 3) How should the safety goals be implemented at the end of the evaluation period. The staff's conclusions are discussed

ITAAC Development for APR1400 NRC Design Certification

International Nuclear Information System (INIS)

Kim, Jin Kwon; Kim, Myoung Ki

2013-01-01

ITAAC (Inspections, Tests, Analyses and Acceptance Criteria) is essential document for Design Certification, which is certified by NRC and ruled as the Appendix of the 10 CFR Part 52. Approximately 870 ITAAC items were selected for APR1400 Design Certification to be submitted to NRC. In this paper, the code and standard related to ITAAC and process of ITACC development are discussed to seek the way to complete the best ITAAC to get the Design Certification of APR1400 from NRC through the lessons learned from other competitive applicants. For saving the time and manpower to complete ITAAC of APR1400 NRC DC and reduction of the RAIs (Request Additional Information) from NRC, we took advantage of the lessons learned from the competitive designs like US-APWR and AP1000 based on KNGR ITAAC. Finally new and different ITAAC from ITAAC of KNGR was developed for APR1400 DC from NRC. RG 1.206 and SRPs were applied first in Korea to the APR1400 DCD preparation. For testability and acceptability check, ITAAC V and V table was completed by the system designers. APR1400 DCD will be submitted by the end of September this year and we hope that NRC issues lees RAIs on ITAAC document

Fuel safety research 2000

Energy Technology Data Exchange (ETDEWEB)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

In April 1999, the Fuel Safety Research Laboratory was newly established as a part of reorganization of the Nuclear Safety Research Center, JAERI. The new laboratory was organized by combining three pre-existing laboratories, Reactivity Accident Laboratory, Fuel Reliability Laboratory, and a part of Severe Accident Research Laboratory. The Fuel Safety Research Laboratory becomes to be in charge of all fuel safety research in JAERI. Various experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of following five research groups corresponding to each research fields; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). The research activities in year 2000 produced many important data and information. They are, for example, failure of high burnup BWR fuel rod under RIA conditions, data on the behavior of hydrided Zircaloy cladding under LOCA conditions and FP release data from VEGA experiments at very high temperature/pressure condition. This report summarizes the outline of research activities and major outcomes of the research executed in 2000 in the Fuel Safety Research Laboratory. (author)

Advanced reactor safety research quarterly report, October-December 1982. Volume 24

Energy Technology Data Exchange (ETDEWEB)

None

1984-04-01

This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

Proceedings of the 21st DOE/NRC nuclear air cleaning conference

International Nuclear Information System (INIS)

First, M.W.

1991-02-01

The 21st meeting of the Department of Energy/Nuclear Regulatory Commission (DOE/NRC) Nuclear Air Cleaning Conference was held in San Diego, CA on August 13--16, 1990. The proceedings have been published as a two volume set. Volume 2 contains sessions covering adsorbents, nuclear codes and standards, modelling, filters, safety, containment venting and a review of nuclear air cleaning programs around the world. Also included is the list of attendees and an index of authors and speakers

Draft pre-application safety evaluation report for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Williams, P.M.; King, T.L.; Wilson, J.N.

1989-03-01

This draft safety evaluation report (SER) presents the preliminary results of a pre-application design review for the standard modular high-temperature gas-cooled reactor (MHTGR) (Project 672). The MHTGR conceptual design was submitted by the U.S. Department of Energy (DOE) in accordance with the U.S. Nuclear Regulatory Commission(NRC) 'Statement of Policy for the Regulation of Advanced Nuclear Power Plants' (51 FR 24643), which provides for early Commission review and interaction. The standard MHTGR consists of four identical reactor modules, each with a thermal output of 350 MWt, coupled with two steam turbine-generator sets to produce a total plant electrical output of 540 MWe. The reactors are helium cooled and graphite moderated and utilize ceramically coated particle-type nuclear fuel. The design includes passive reactor-shutdown and decay-heat-removal features. The staff and its contractors at the Oak Ridge National Laboratory and the Brookhaven National Laboratory have reviewed this design with emphasis on those unique provisions in the design that accomplish the key safety functions of reactor shutdown, decay-heat removal, and containment of radioactive material. This report presents the NRC staff's technical evaluation of those features in the MHTGR design important to safety, including their proposed research and testing needs. In addition this report presents the criteria proposed by the NRC staff to judge the acceptability of the MHTGR design and, where possible, includes statements on the potential of the MHTGR to meet these criteria. However, it should be recognized that final conclusions in all matters discussed in this report require approval by the Commission. Final determination on the acceptability of the MHTGR standard design is contingent on receipt and evaluation of additional information requested from DOE pertaining to the adequacy of the containment design and on the following: (1) satisfactory resolution of open safety issues identified

Transactions of the twenty-second water reactor safety information meeting

International Nuclear Information System (INIS)

1994-10-01

This report contains summaries of papers on reactor safety research to be presented at the 22nd Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 24--26, 1994. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session. Individual papers have been cataloged separately

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

International Nuclear Information System (INIS)

2017-01-01

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

Methods and strategies for future reactor safety goals

Science.gov (United States)

Arndt, Steven Andrew

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

Aging research

International Nuclear Information System (INIS)

Ross, D.F. Jr.

1988-01-01

The USNRC Office of Nuclear Regulatory Research has developed a program for nuclear plant aging research (NPAR) to achieve an understanding of nuclear plant aging, its potential effects on safety, and methods for its detection and mitigation, sufficient for addressing safety and regulatory issues and supporting regulatory decisions on issues. Specifically, the NRC has aggressive research and regulatory programs associated with aging effects on piping, steam generators, containments, structures, and electrical and mechanical systems and components. In addition to safety assessment for the original license period for nuclear power plants, this aging information will be extremely useful in providing technical bases for efficient and effective regulation associated with possible license extension. This paper discusses the major activities of USNRC sponsored aging research program and recommends an approach to manage and handle aging at nuclear power plants

SU-A-210-02: Medical Physics Opportunities at the NRC

International Nuclear Information System (INIS)

Abogunde, M.

2015-01-01

The purpose of this student annual meeting is to address topics that are becoming more relevant to medical physicists, but are not frequently addressed, especially for students and trainees just entering the field. The talk is divided into two parts: medical billing and regulations. Hsinshun Wu – Why should we learn radiation oncology billing? Many medical physicists do not like to be involved with medical billing or coding during their career. They believe billing is not their responsibility and sometimes they even refuse to participate in the billing process if given the chance. This presentation will talk about a physicist’s long career and share his own experience that knowing medical billing is not only important and necessary for every young medical physicist, but that good billing knowledge could provide a valuable contribution to his/her medical physics development. Learning Objectives: The audience will learn the basic definition of Current Procedural Terminology (CPT) codes performed in a Radiation Oncology Department. Understand the differences between hospital coding and physician-based or freestanding coding. Apply proper CPT coding for each Radiation Oncology procedure. Each procedure with its specific CPT code will be discussed in detail. The talk will focus on the process of care and use of actual workflow to understand each CPT code. Example coding of a typical Radiation Oncology procedure. Special procedure coding such as brachytherapy, proton therapy, radiosurgery, and SBRT. Maryann Abogunde – Medical physics opportunities at the Nuclear Regulatory Commission (NRC) The NRC’s responsibilities include the regulation of medical uses of byproduct (radioactive) materials and oversight of medical use end-users (licensees) through a combination of regulatory requirements, licensing, safety oversight including inspection and enforcement, operational experience evaluation, and regulatory support activities. This presentation will explore the

SU-A-210-02: Medical Physics Opportunities at the NRC

Energy Technology Data Exchange (ETDEWEB)

Abogunde, M. [U.S. Nuclear Regulatory Commission (United States)

2015-06-15

The purpose of this student annual meeting is to address topics that are becoming more relevant to medical physicists, but are not frequently addressed, especially for students and trainees just entering the field. The talk is divided into two parts: medical billing and regulations. Hsinshun Wu – Why should we learn radiation oncology billing? Many medical physicists do not like to be involved with medical billing or coding during their career. They believe billing is not their responsibility and sometimes they even refuse to participate in the billing process if given the chance. This presentation will talk about a physicist’s long career and share his own experience that knowing medical billing is not only important and necessary for every young medical physicist, but that good billing knowledge could provide a valuable contribution to his/her medical physics development. Learning Objectives: The audience will learn the basic definition of Current Procedural Terminology (CPT) codes performed in a Radiation Oncology Department. Understand the differences between hospital coding and physician-based or freestanding coding. Apply proper CPT coding for each Radiation Oncology procedure. Each procedure with its specific CPT code will be discussed in detail. The talk will focus on the process of care and use of actual workflow to understand each CPT code. Example coding of a typical Radiation Oncology procedure. Special procedure coding such as brachytherapy, proton therapy, radiosurgery, and SBRT. Maryann Abogunde – Medical physics opportunities at the Nuclear Regulatory Commission (NRC) The NRC’s responsibilities include the regulation of medical uses of byproduct (radioactive) materials and oversight of medical use end-users (licensees) through a combination of regulatory requirements, licensing, safety oversight including inspection and enforcement, operational experience evaluation, and regulatory support activities. This presentation will explore the

Comparison of the Safety Critical Software V and V Requirements for the Research Reactor Instrumentation and Control System

Energy Technology Data Exchange (ETDEWEB)

Joo, Sungmoon; Suh, Yong-Suk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

This study was motivated by a research reactor project where the owner of the project and the equipment vendors are from two different standards frameworks. This paper reviews two major standards frameworks - NRC-IEEE and IAEA-IEC - and the software classification schemes as a background, then discuss the V and V issue. The purpose of this paper is by no means to solve the cross-standards-framework qualification issue, but, rather, is to remind the stakeholders of research reactor projects. V and V are also essential for the approval from regulatory bodies. As standards define or recommend consolidated engineering practices, methods, or criteria, V and V activities for software qualification are not exceptional. Within a standards framework, usually, the processes for the qualification of safety-critical software are well-established such that the safety is maximized while minimizing the compromises in software quality, safety, and reliability. When, however, multiple standards frameworks are involved in a research reactor project, it is difficult for equipment vendors to implement appropriate V and V activities as there is no unified view on this cross-standards-framework qualification issue yet. There are two major standards frameworks for safety-critical software development in nuclear industry. Unfortunately different safety classifications for software and thus different requirements for qualification are in place. What makes things worse is that (i) there are ambiguities in the standards and rooms for each stakeholders’ interpretation, and (ii) there is no one-to-one mapping between the associated V and V methods and activities. These may put the stakeholders of research reactor projects in trouble.

Comparison of the Safety Critical Software V and V Requirements for the Research Reactor Instrumentation and Control System

International Nuclear Information System (INIS)

Joo, Sungmoon; Suh, Yong-Suk; Park, Cheol

2016-01-01

This study was motivated by a research reactor project where the owner of the project and the equipment vendors are from two different standards frameworks. This paper reviews two major standards frameworks - NRC-IEEE and IAEA-IEC - and the software classification schemes as a background, then discuss the V and V issue. The purpose of this paper is by no means to solve the cross-standards-framework qualification issue, but, rather, is to remind the stakeholders of research reactor projects. V and V are also essential for the approval from regulatory bodies. As standards define or recommend consolidated engineering practices, methods, or criteria, V and V activities for software qualification are not exceptional. Within a standards framework, usually, the processes for the qualification of safety-critical software are well-established such that the safety is maximized while minimizing the compromises in software quality, safety, and reliability. When, however, multiple standards frameworks are involved in a research reactor project, it is difficult for equipment vendors to implement appropriate V and V activities as there is no unified view on this cross-standards-framework qualification issue yet. There are two major standards frameworks for safety-critical software development in nuclear industry. Unfortunately different safety classifications for software and thus different requirements for qualification are in place. What makes things worse is that (i) there are ambiguities in the standards and rooms for each stakeholders’ interpretation, and (ii) there is no one-to-one mapping between the associated V and V methods and activities. These may put the stakeholders of research reactor projects in trouble

Independent peer review of nuclear safety computer codes

International Nuclear Information System (INIS)

Boyack, B.E.; Jenks, R.P.

1993-01-01

A structured, independent computer code peer-review process has been developed to assist the US Nuclear Regulatory Commission (NRC) and the US Department of Energy in their nuclear safety missions. This paper describes a structured process of independent code peer review, benefits associated with a code-independent peer review, as well as the authors' recent peer-review experience. The NRC adheres to the principle that safety of plant design, construction, and operation are the responsibility of the licensee. Nevertheless, NRC staff must have the ability to independently assess plant designs and safety analyses submitted by license applicants. According to Ref. 1, open-quotes this requires that a sound understanding be obtained of the important physical phenomena that may occur during transients in operating power plants.close quotes The NRC concluded that computer codes are the principal products to open-quotes understand and predict plant response to deviations from normal operating conditionsclose quotes and has developed several codes for that purpose. However, codes cannot be used blindly; they must be assessed and found adequate for the purposes they are intended. A key part of the qualification process can be accomplished through code peer reviews; this approach has been adopted by the NRC

Cyber Security in Nuclear Power Plants - U.S. NRC Regulatory Guide 5.71

International Nuclear Information System (INIS)

Pogacic, Goran

2014-01-01

We have already made a big step into new millennia and with it there is no more dilemma about presence of computers and internet in our lives. Almost all modern facilities struggle with this new dimension of information flow and how to use it to their best interest. But there is also the other side of the coin- the security threat. For nuclear power plants this threat poses even greater risk. In addition to protecting their trade secrets, personal data or other common targets of cyber attacks, nuclear power plants need to protect their digital computers, communication systems and networks up to and including the design basis threat (DBT). As stated in U.S. Nuclear Regulatory Commission (NRC) Regulatory Commission Regulations, Title 10, Code of Federal Regulations (CFR), section 73.1, 'Purpose and Scope' this includes protection against acts of radiological sabotage and prevention of the theft or diversion of special nuclear material. The main purpose of this paper is to explore the NRC Regulatory Guide (RG) 5.71 and its guidance in implementing cyber security requirements stated in NRC 10 CFR, section 73.54, 'Protection of Digital Computer and Communication Systems and Networks'. In particular, this section requires protection of digital computers, communication systems and networks associated with the following categories of functions: · safety-related and important-to-safety functions, · security functions, · emergency preparedness functions, including offsite communication, and · support systems and equipment which, if compromised, would adversely impact safety, security, or emergency preparedness functions. This section requires protection of such systems and networks from those cyber attacks that would act to modify, destroy, or compromise the integrity or confidentiality of data or software; deny access to systems, services or data; and impact the operation of systems, networks, and equipment. This paper will also present some of

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

International Nuclear Information System (INIS)

2001-07-01

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

Resolution of thermal-hydraulic safety and licensing issues for the system 80+trademark design

International Nuclear Information System (INIS)

Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E.

1995-01-01

The System 80+ trademark Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC's new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs

Public citizen slams NRC on nuclear inspections

International Nuclear Information System (INIS)

Newman, P.

1993-01-01

Charging the Nuclear Regulatory Commission with open-quotes abandoning tough regulation of the nuclear power industry,close quotes Public Citizen's Critical Mass Energy Project on Wednesday released a report asserting that NRC is shielding sensitive internal nuclear industry self-evaluations from public scrutiny. Based on their review of 56 Institute of Nuclear Power Operations reports and evaluations and comparing these to the NRC's Systematic Assessment of Licensee Performance reports for the same plants, it was concluded that the NRC failed to address issues raised in all eight areas evaluated by the INPO reports

Recommendations for NEAMS Engagement with the NRC: Preliminary Report

International Nuclear Information System (INIS)

Bernholdt, David E.

2012-01-01

The vision of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is to bring a new generation of analytic tools to the nuclear engineering community in order to facilitate students, faculty, industry and laboratory researchers in investigating advanced reactor and fuel cycle designs. Although primarily targeting at advance nuclear technologies, it is anticipated that these new capabilities will also become interesting and useful to the nuclear regulator Consequently, the NEAMS program needs to engage with the Nuclear Regulatory Commission as the software is being developed to ensure that they are familiar with and ready to respond to this novel approach when the need arises. Through discussions between key NEAMS and NRC staff members, we tentatively recommend annual briefings to the Division of Systems Analysis in the NRC's Office of Nuclear Regulatory Research. However the NEAC subcommittee review of the NEAMS program may yield recommendations that would need to be considered before finalizing this plan.

Recommendations for NEAMS Engagement with the NRC: Preliminary Report

Energy Technology Data Exchange (ETDEWEB)

Bernholdt, David E [ORNL

2012-06-01

The vision of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is to bring a new generation of analytic tools to the nuclear engineering community in order to facilitate students, faculty, industry and laboratory researchers in investigating advanced reactor and fuel cycle designs. Although primarily targeting at advance nuclear technologies, it is anticipated that these new capabilities will also become interesting and useful to the nuclear regulator Consequently, the NEAMS program needs to engage with the Nuclear Regulatory Commission as the software is being developed to ensure that they are familiar with and ready to respond to this novel approach when the need arises. Through discussions between key NEAMS and NRC staff members, we tentatively recommend annual briefings to the Division of Systems Analysis in the NRC's Office of Nuclear Regulatory Research. However the NEAC subcommittee review of the NEAMS program may yield recommendations that would need to be considered before finalizing this plan.

Creation of nuclear research centres in new entrant countries

International Nuclear Information System (INIS)

Caoui, A.; Monnet, F.

2010-01-01

A number of prerequisites must be fulfilled before setting up a nuclear program in a new entrant country. This includes the training of future operating teams in the safety culture and the acquisition of the necessary technological knowledge. A NRC brings together the facilities and structures to meet this need. Similarly, prerequisites for the construction of an NRC in Maamora (Morocco) have been completed by AREVA. These include establishing the contract, constructing the nuclear research centre and above all, commissioning the facilities. This paper gives the feedback from the team formed by AREVA and CNESTEN on the adventure the construction of the NRC Maamora has been. (author)

Assessment of current NRC/IE professional training program and recommendations for improvement

International Nuclear Information System (INIS)

Bartley, H.J.; Hagerup, J.E.; Harrison, O.J.; Heyer, F.H.K.; Kaas, I.W.; Schwartz, E.G.

1978-05-01

This document is the General Research Corporation (GRC) report on Task III: to assess the current NRC/IE professional training program and to provide recommendations for improvement. The major objectives of this task were to determine the overall effectiveness of the NRC/IE training program and to provide recommendations for improvements where appropriate. The research involved a review of course manuals and of student critiques, observation in the classroom and person to person interviews; it also included an evaluation of the assignment of instructors to the Career Management Branch. Findings addressed refresher training, retread training and initial training--with emphasis on the last of these. Conclusions are that: (1) The curriculum provides, in general, types and levels of training needed; (2) the mix of training methods used is correct; and (3) the training management is effective. However, the training facilities do not reflect a commitment to quality instruction nor is assignment as instructor to the Career Management Branch attractive to inspectors. Recommendations presented in the report are based upon the findings; all lie within the implementing authority of Headquarters NRC/IE

The NRC perspective on low-level radioactive waste disposal

International Nuclear Information System (INIS)

Thompson, H.L. Jr.; Knapp, M.R.

1987-01-01

This paper describes the Nuclear Regulatory Commission's (NRC) actions in response to the Low-Level Radioactive Waste Policy Amendments Act (the Act) and NRC's assistance to States and Compacts working to discharge their responsibilities under the Act. Three of NRC's accomplishments which respond explicitly to direction in the Act are highlighted. These are: development of the capability of expedited handling of petitions addressing wastes below regulatory concern (BRC); development of capability to review and process an application within fifteen months; and development of guidance on alternatives to shallow land burial. Certain NRC efforts concerning special topics related to the Act as well as NRC efforts to assist States and Compacts are summarized

Development of the NRC`s Human Performance Investigation Process (HPIP). Volume 2, Investigators`s Manual

Energy Technology Data Exchange (ETDEWEB)

Paradies, M.; Unger, L. [System Improvements, Inc., Knoxville, TN (United States); Haas, P.; Terranova, M. [Concord Associates, Inc., Knoxville, TN (United States)

1993-10-01

The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause event at nuclear power plants. This document, Volume II, is a field manual for use by investigators when performing event investigations. Volume II includes the HPIP Procedure, the HPIP Modules, and Appendices that provide extensive documentation of each investigation technique.

Progress of nuclear safety research, (1)

International Nuclear Information System (INIS)

Amano, Hiroshi; Nakamura, Hiroei; Nozawa, Masao

1981-01-01

The Japan Atomic Energy Research Institute was established in 1956 in conformity with the national policy to extensively conduct the research associated with nuclear energy. Since then, the research on nuclear energy safety has been conducted. In 1978, the Division of Reactor Safety was organized to conduct the large research programs with large scale test facilities. Thereafter, the Divisions of Reactor Safety Evaluation, Environmental Safety Research and Reactor Fuel Examination were organized successively in the Reactor Safety Research Center. The subjects of research have ranged from the safety of nuclear reactors to that in the recycling of nuclear fuel. In this pamphlet, the activities in JAERI associated with the safety research are reported, which have been carried out in the past two years. Also, the international cooperation research program in which JAERI participated is included. This pamphlet consists of two parts, and in this Part 1, the reactor safety research is described. The safety of nuclear fuel, the integrity and safety of pressure boundary components, the engineered safety in LOCA, fuel behavior in accident and others are reported. (Kako, I.)

HSE Nuclear Safety Research Program

Energy Technology Data Exchange (ETDEWEB)

Bagley, M.J. [Health and Safety Executive, Sheffield (United Kingdom)

1995-12-31

HSE funds two programmes of nuclear safety research: a programme of {approx} 2.2M of extramural research to support the Nuclear Safety Division`s regulatory activities and a programme of {approx} 11M of generic safety research managed by the Nuclear Safety Research Management Unit (NSRMU) in Sheffield, UK. This paper is concerned only with the latter programme; it describes how it is planned and procured and outlines some of the work on structural integrity problems. It also describes the changes that are taking place in the way nuclear safety research is procured in the UK. (author).

HSE Nuclear Safety Research Program

International Nuclear Information System (INIS)

Bagley, M.J.

1995-01-01

HSE funds two programmes of nuclear safety research: a programme of ∼ 2.2M of extramural research to support the Nuclear Safety Division's regulatory activities and a programme of ∼ 11M of generic safety research managed by the Nuclear Safety Research Management Unit (NSRMU) in Sheffield, UK. This paper is concerned only with the latter programme; it describes how it is planned and procured and outlines some of the work on structural integrity problems. It also describes the changes that are taking place in the way nuclear safety research is procured in the UK. (author)

10 CFR 2.1202 - Authority and role of NRC staff.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Authority and role of NRC staff. 2.1202 Section 2.1202... ORDERS Informal Hearing Procedures for NRC Adjudications § 2.1202 Authority and role of NRC staff. (a) During the pendency of any hearing under this subpart, consistent with the NRC staff's findings in its...

Annual safety research report, JFY 2010

International Nuclear Information System (INIS)

2011-09-01

In the safety infrastructure research working group report, 'the effective conducting of nuclear safety infrastructure research', published by METI in March 2010, the roles of regulatory agencies and JNES and their cooperation, and the research road map for nuclear safety regulation researches were summarized. As for the regulatory issues the governments or JNES considered necessary, JNES had compiled' safety research plan' in respective research areas necessary for solving the regulatory issues (safety research needs) and was conducting safety research to obtain the results, etc. Safety research areas, subjects and research projects were as follows: design review of nuclear power plant (4 subjects and each subject having several research projects totaled 19), control management of nuclear power plant (3 subjects and each subject having several research projects totaled 11), nuclear fuel cycle (2 subjects and each subject having several research projects totaled 5), nuclear fuel cycle backend (2 subjects and each subject having several research projects totaled 6), nuclear emergency preparedness and response (3 subjects and each subject having several research projects totaled 5) and bases of nuclear safety technology (3 subjects and each subject having several research projects totaled 7). In JFY 2010, JNES worked on the 53 research projects of 17 subjects in 6 areas as safety researches. This annual safety research report summarized respective achievements and stage of regulatory tools necessary for solving regulatory issues according to the safety research plan, JFY 2010 Edition as well as the situation of the reflection for the safety regulations. (T. Tanaka)

Fuel safety research 2001

Energy Technology Data Exchange (ETDEWEB)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-11-01

The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

NRC plan for cleanup operations at Three Mile Island Unit 2

International Nuclear Information System (INIS)

Lo, R.; Snyder, B.J.

1980-07-01

The NRC plan defines the functional role of the NRC in cleanup operations at Three Mile Island Unit 2 to assure that agency regulatory responsibilities and objectives will be fulfilled. The plan outlines NRC functions in TMI-2 cleanup operations in the following areas: (1) the functional relationship of NRC to other government agencies, the public, and the licensee to coordinate activities, (2) the functional roles of these organizations in cleanup operations, (3) the NRC review and decision-making procedure for the licensee's proposed cleanup operation, (4) the NRC/licensee estimated schedule of major actions, and (5) NRC's functional role in overseeing implementation of approved licensee activities

Chemical process safety at fuel cycle facilities

International Nuclear Information System (INIS)

Ayres, D.A.

1997-08-01

This NUREG provides broad guidance on chemical safety issues relevant to fuel cycle facilities. It describes an approach acceptable to the NRC staff, with examples that are not exhaustive, for addressing chemical process safety in the safe storage, handling, and processing of licensed nuclear material. It expounds to license holders and applicants a general philosophy of the role of chemical process safety with respect to NRC-licensed materials; sets forth the basic information needed to properly evaluate chemical process safety; and describes plausible methods of identifying and evaluating chemical hazards and assessing the adequacy of the chemical safety of the proposed equipment and facilities. Examples of equipment and methods commonly used to prevent and/or mitigate the consequences of chemical incidents are discussed in this document

NRC perspectives on fuel cycle and safeguards

International Nuclear Information System (INIS)

Chapman, K.R.

1976-01-01

This paper discusses NRC's mandate in the field of safeguards and the thoughts of NRC on other newly emerging policy considerations. The status of some of the current issues facing the nuclear community and the regulatory staff in particular is touched on

The importance of independent research and evaluation in assessing nuclear fuel cycle and waste management facility safety

International Nuclear Information System (INIS)

Downing, Walter D.; Patrick, Wesley C.; Sagar, Budhi

2009-01-01

In 1987, the United States Nuclear Regulatory Commission (NRC) established at Southwest Research Institute (SwRI) a federally funded research and development center. Known as the Center for Nuclear Waste Regulatory Analyses (CNWRA), its overall mission is to provide NRC with an independent assessment capability on technical and regulatory issues related to a potential geologic repository for spent nuclear fuel and high-level radioactive waste, as well as interim storage and other nuclear fuel-cycle facilities. For more than 20 years, the CNWRA has supported NRC through an extensive pre-licensing period of establishing the framework of regulations and guidance documents, developing computer codes and other review tools, and conducting independent laboratory, field, and numerical analyses. In June 2008, the United States Department of Energy (DOE) submitted a license application and final environmental impact statement to NRC seeking authorization to construct the nation's first geologic repository at Yucca Mountain, Nevada. The CNWRA will assist NRC in conducting a detailed technical review to critically evaluate the DOE license application to assess whether the potential repository has been designed and can be constructed and operated to safely dispose spent nuclear fuel and high-level radioactive waste. NRC access to independent, unbiased, technical advice from the CNWRA is an important aspect of the evaluation process. This paper discusses why an independent perspective is important when dealing with nuclear fuel cycle and waste management issues. It addresses practical considerations such as avoiding conflicts of interest while at the same time maintaining a world-class research program in technical areas related to the nuclear fuel cycle. It also describes an innovative approach for providing CNWRA scientists and engineers a creative outlet for professional development through an internally funded research program that is focused on future nuclear waste

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

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

NRC methods for evaluation of industry training

International Nuclear Information System (INIS)

Morisseau, D.S.; Koontz, J.L.; Persensky, J.J.

1987-01-01

On March 20, 1985, the Nuclear Regulatory Commission published the Policy Statement on Training and Qualification. The Policy Statement endorsed the INPO-managed Training Accreditation Program because it encompasses the five elements of performance-based training. This paper described the multiple methods that the NRC is using to monitor industry efforts to improve training and implement the NRC Policy Statement on Training and Qualification. The results of the evaluation of industry training improvement programs will be reviewed by the Commissioners in April 1987 to determine the nature of continuing NRC policy and programs for ensuring effective training for the US nuclear industry

Research program on regulatory safety research

International Nuclear Information System (INIS)

Mailaender, R.

2010-02-01

This paper elaborated for the Swiss Federal Office of Energy (SFOE) presents the synthesis report for 2009 made by the SFOE's program leader on the research program concerning regulatory nuclear safety research, as co-ordinated by the Swiss Nuclear Safety Inspectorate ENSI. Work carried out in various areas is reviewed, including that done on reactor safety, radiation protection and waste disposal as well as human aspects, organisation and safety culture. Work done concerning materials, pressure vessel integrity, transient analysis, the analysis of serious accidents in light-water reactors, fuel and material behaviour, melt cooling and concrete interaction is presented. OECD data bank topics are discussed. Transport and waste disposal research at the Mont Terri rock laboratory is looked at. Requirements placed on the personnel employed in nuclear power stations are examined and national and international co-operation is reviewed

Resolution of thermal-hydraulic safety and licensing issues for the system 80+{sup {trademark}} design

Energy Technology Data Exchange (ETDEWEB)

Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E. [ABB-Combustion Engineering, Windsor, CT (United States)] [and others

1995-09-01

The System 80+{sup {trademark}} Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC`s new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs.

Nuclear safety: risks and regulation

International Nuclear Information System (INIS)

Wood, W.C.

1983-01-01

Taking a fresh look at nuclear safety regulations, this study finds that the mandate and organization of the Nuclear Regulatory Commission (NRC) militate against its making sound decisions. The author criticizes failures to make hard decisions on societal risk, to clarify responsibility, and to implement cost-effective safety measures. Among his recommendations are reorganization of the NRC under a single authoritative administrator, separation of technical issues from social ones, and reform of the Price-Anderson Act. The author concludes that the worst eventuality would be to continue the current state of indecision. 161 references, 6 figures, 4 tables

Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

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

Significant NRC Enforcement Actions

Data.gov (United States)

Nuclear Regulatory Commission — This dataset provides a list of Nuclear Regulartory Commission (NRC) issued significant enforcement actions. These actions, referred to as "escalated", are issued by...

Expression and purification of toxic anti-breast cancer p28-NRC chimeric protein

OpenAIRE

Soleimani, Meysam; Mirmohammad-Sadeghi, Hamid; Sadeghi-Aliabadi, Hojjat; Jahanian-Najafabadi, Ali

2016-01-01

Background: Chimeric proteins consisting of a targeting moiety and a cytotoxic moiety are now under intense research focus for targeted therapy of cancer. Here, we report cloning, expression, and purification of such a targeted chimeric protein made up of p28 peptide as both targeting and anticancer moiety fused to NRC peptide as a cytotoxic moiety. However, since the antimicrobial activity of the NRC peptide would intervene expression of the chimeric protein in Escherichia coli, we evaluated...

What it took to get an NRC license for centralized incineration

International Nuclear Information System (INIS)

DiSalvo, R.; Zielenbach, W.

1987-01-01

In 1982, Battelle joined five other commercial generators of low level radioactive waste in conducting a study of the technical and economic feasibility and the licensability of a central facility for incinerating LLW. The project generated a license application to the USNRC and supporting documentation related to the safety and environmental impacts of the facility. After thorough review, the NRC has issued a Finding of No Significant Impact and the associated license authorization, which is the first of its kind for an incineration facility

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

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

Perspectives on reactor safety. Revision 1

International Nuclear Information System (INIS)

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

1997-11-01

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

Perspectives on reactor safety. Revision 1

Energy Technology Data Exchange (ETDEWEB)

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

1997-11-01

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

Human Factors Regulatory Research Program Plan, FY 1989--FY 1992

International Nuclear Information System (INIS)

Coffman, F.; Persensky, J.; Ryan, T.; Ramey-Smith, A.; Goodman, C.; Serig, D.; Trager, E; Nuclear Regulatory Commission, Washington, DC; Nuclear Regulatory Commission, Washington, DC; Nuclear Regulatory Commission, Washington, DC

1989-10-01

This report describes the currently ongoing (FY 1989) and planned (FY 1989-1992) Human Factors Regulatory Research Program in the NRC Office of Nuclear Regulatory Research (RES). Examples of the influence of human factors on nuclear safety are presented, and the role of personnel is discussed. Current regulatory issues associated with human factors in the nuclear system and the purpose of the research plan are provided. The report describes the research process applied to the human factors research issues and the program activities: Personnel Performance Measurement, Personnel Subsystem, Human-System Interface. Organization and Management, and Reliability Assessment. The research being conducted within each activity is summarized along with the objectives, background information, and expected regulatory products. Budget and personnel forecasts are provided along with a summary of contractors performing some of the ongoing research. Appendices contain a chronology of human factors research at NRC, a description of the research approach, an update on human factors programs and initiatives in RES and other NRC offices, and the integration among these programs. 46 refs., 5 tabs

10 CFR 2.1505 - Role of the NRC staff.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Role of the NRC staff. 2.1505 Section 2.1505 Energy... Legislative Hearings § 2.1505 Role of the NRC staff. The NRC staff shall be available to answer any Commission or presiding officer's questions on staff-prepared documents, provide additional information or...

Progress of nuclear safety research, (2)

International Nuclear Information System (INIS)

Amano, Hiroshi; Nakamura, Hiroei; Nozawa, Masao

1981-01-01

The Japan Atomic Energy Research Institute was established in 1956 in conformity with the national policy to extensively conduct the research associated with nuclear energy. Since then, the research on nuclear energy safety has been conducted. In 1978, the Division of Reactor Safety was organized to conduct the large research programs with large scale test facilities. Thereafter, the Divisions of Reactor Safety Evaluation, Environmental Safety Research and Reactor Fuel Examination were organized successevely in the Reactor Safety Research Center. The subjects of research have ranged from the safety of nuclear reactors to that in the recycling of nuclear fuel. In this pamphlet, the activities in JAERI associated with the safety research are reported, which have been carried out in the past two years. Also the international cooperation research program in which JAERI participated is included. This pamphlet consists of two parts and in this Part 2, the environmental safety research is described. The evaluation and analysis of environmental radioactivity, the study on radioactive waste management and the studies on various subjects related to environmental safety are reported. (Kako, I.)

Progress of nuclear safety research. 2001

Energy Technology Data Exchange (ETDEWEB)

Anoda, Yoshinari; Sasajima, Hideo; Nishiyama, Yutaka (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

2001-10-01

JAERI is conducting nuclear safety research primarily at the Nuclear Safety Research Center in close cooperation with the related departments in accordance with the Long Term Plan for Development and Utilization of Nuclear Energy or the Safety Research Annual Plan issued by the Japanese government. The safety research at JAERI concerns the engineering safety of nuclear power plants and nuclear fuel cycle facilities, and radioactive waste management as well as advanced technology for safety improvement or assessment. Also, JAERI has conducted international collaboration to share the information on common global issues of nuclear safety. This report summarizes the nuclear safety research activities of JAERI from April 1999 through March 2001. (author)

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

International Nuclear Information System (INIS)

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

1994-02-01

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

An Empirical Analysis of Human Performance and Nuclear Safety Culture

International Nuclear Information System (INIS)

Jeffrey Joe; Larry G. Blackwood

2006-01-01

The purpose of this analysis, which was conducted for the US Nuclear Regulatory Commission (NRC), was to test whether an empirical connection exists between human performance and nuclear power plant safety culture. This was accomplished through analyzing the relationship between a measure of human performance and a plant's Safety Conscious Work Environment (SCWE). SCWE is an important component of safety culture the NRC has developed, but it is not synonymous with it. SCWE is an environment in which employees are encouraged to raise safety concerns both to their own management and to the NRC without fear of harassment, intimidation, retaliation, or discrimination. Because the relationship between human performance and allegations is intuitively reciprocal and both relationship directions need exploration, two series of analyses were performed. First, human performance data could be indicative of safety culture, so regression analyses were performed using human performance data to predict SCWE. It also is likely that safety culture contributes to human performance issues at a plant, so a second set of regressions were performed using allegations to predict HFIS results

Regulation of chemical safety at fuel cycle facilities by the United States Nuclear Regulatory Commission

International Nuclear Information System (INIS)

Ramsey, Kevin M.

2013-01-01

When the U.S. Nuclear Regulatory Commission (NRC) was established in 1975, its regulations were based on radiation dose limits. Chemical hazards rarely influenced NRC regulations. After the Three Mile Island reactor accident in 1979, the NRC staff was directed to address emergency planning at non-reactor facilities. Several fuel cycle facilities were ordered to submit emergency plans consistent with reactor emergency plans because no other guidance was available. NRC published a notice that it was writing regulations to codify the requirements in the Orders and upgrade the emergency plans to address all hazards, including chemical hazards. The legal authority of NRC to regulate chemical safety was questioned. In 1986, an overfilled uranium hexafluoride cylinder ruptured and killed a worker. The NRC staff was directed to address emergency planning for hazardous chemicals in its regulations. The final rule included a requirement for fuel cycle facilities to certify compliance with legislation requiring local authorities to establish emergency plans for hazardous chemicals. As with emergency planning, NRC's authority to regulate chemical safety during routine operations was limited. NRC established memoranda of understanding (MOUs) with other regulatory agencies to encourage exchange of information between the agencies regarding occupational hazards. In 2000, NRC published new, performance-based, regulations for fuel cycle facilities. The new regulations required an integrated safety analysis (ISA) which used quantitative standards to assess chemical exposures. Some unique chemical exposure cases were addressed while implementing the new regulations. In addition, some gaps remain in the regulation of hazardous chemicals at fuel cycle facilities. The status of ongoing efforts to improve regulation of chemical safety at fuel cycle facilities is discussed. (authors)

Nuclear safety research master plan

Energy Technology Data Exchange (ETDEWEB)

Ha, Jae Joo; Yang, J. U.; Jun, Y. S. and others

2001-06-01

The SRMP (Safety Research Master Plan) is established to cope with the changes of nuclear industry environments. The tech. tree is developed according to the accident progress of the nuclear reactor. The 11 research fields are derived to cover the necessary technologies to ensure the safety of nuclear reactors. Based on the developed tech. tree, the following four main research fields are derived as the main safety research areas: 1. Integrated nuclear safety enhancement, 2. Thermal hydraulic experiment and assessment, 3. Severe accident management and experiment, and 4. The integrity of equipment and structure. The research frame and strategies are also recommended to enhance the efficiency of research activity, and to extend the applicability of research output.

Compilation of contract research for the Materials Branch, Division of Engineering Safety: Annual report for FY 1986

International Nuclear Information System (INIS)

1987-03-01

This report presents summaries of the research work performed during Fiscal Year 1986 by laboratories and organizations under contracts administered by the NRC's Materials Branch, Office of Nuclear Regulatory Research. Each contractor has written a more complete and detailed annual report of their work which can be obtained by writing to NRC; however, we believe it is useful to have a summary of each contractor's efforts for the year combined into one volume

NRC Regulatory Agenda

International Nuclear Information System (INIS)

1989-07-01

This document is a compilation of all rules on which the NRC has proposed or is considering action and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

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

Safety research plan, JFY 2013 edition

International Nuclear Information System (INIS)

2013-09-01

As for the regulatory issues the governments or JNES considered necessary, JNES had updated every year 'safety research plan' in respective research areas necessary for solving the regulatory issues (safety research needs) and was conducting safety research to obtain the results, etc. 'Safety research plan, JFY 2013 Edition' was compiled aiming at promotion of appropriate reflection and flexible application of research achievements for tacking the regulatory issues taking account of importance and urgency dependent on trend of nuclear safety regulations as well as collective management of safety research and safety survey. 5 new research projects were established with 4 unified research projects and 6 terminated research projects. Finally modified safety research areas, subjects and research projects, JFY 2013 Edition were as follows: design review of nuclear power plant (7 subjects and each subject having several research projects totaled 19), control management of nuclear power plant (one subject having 4 research projects), nuclear fuel cycle (2 subjects and each subject having several research projects totaled 4), nuclear fuel cycle backend (2 subjects and each subject having several research projects totaled 5), nuclear emergency preparedness and response (3 subjects and each subject having several research projects totaled 7) and bases of nuclear safety technology (3 subjects and each subject having several research projects totaled 6). Safety reviews consisted of 6 projects in 3 areas extracting the regulatory issues. As for urgent research projects on the basis of the disaster at Fukushima Daiichi NPP accident, 7 research projects in 4 urgent subjects were as follows: examination for new safety regulation (4 research projects generalized in the above research projects), development of newly necessary evaluation methods (one research project generalized in the above research project), evaluation of the validity for the work for convergence at Fukushima

NRC integrated program for the resolution of Unresolved Safety Issues A-3, A-4 and A-5 regarding steam generator tube integrity: Final report

International Nuclear Information System (INIS)

1988-09-01

This report presents the results of the NRC integrated program for the resolution of Unresolved Safety Issues (USIs) A-3, A-4, and A-5 regarding steam generator tube integrity. A generic risk assessment is provided and indicates that risk from steam generator tube rupture (SGTR) events is not a significant contributor to total risk at a given site, nor to the total risk to which the general public is routinely exposed. This finding is considered to be indicative of the effectiveness of licensee programs and regulatory requirements for ensuring steam generator tube integrity in accordance with 10 CFR 50, Appendices A and B. This report also identifies a number of staff-recommended actions that the staff finds can further improve the effectiveness of licensee programs in ensuring the integrity of steam generator tubes and in mitigating the consequences of an SGTR. As part of the integrated program, the staff issued Generic Letter 85-02 encouraging licensees of pressurized water reactors (PWRs) to upgrade their programs, as necessary, to meet the intent of the staff-recommended actions; however, such actions do not constitute NRC requirements. In addition, this report describes a number of ongoing staff actions and studies involving steam generator issues which are being pursued to provide added assurance that risk from SGTR events will continue to be small. 146 refs., 5 figs., 11 tabs

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

International Nuclear Information System (INIS)

1975-06-01

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

U.S. regulatory initiatives related to nuclear research and development

International Nuclear Information System (INIS)

Carr, K.M.

1987-01-01

U.S. Nuclear Regulatory Commission is working to improve its 'early warning system' in order to recognize trends and identify potential safety problems whether the source of the problems relates to a reactor safety system or to a utility management or organizational issue. The Commission has approved the use of performance indicators to provide a more effective means for top NRC management and the Commission to recognize changes in the safety performance of operating reactors. Research also has a leading role in assuring the safety of the evolving nuclear technology. The primary role of NRC's program is to provide a sound technical basis for decisions on the resolution of safety issues for the current generation of plants and safety issues that may arise for future designs. Topical areas essential to resolve the remaining safety issues or uncertainties related to LWR operation are human factors and accident management, life extension and plant aging, studies on containment integrity, severe accident risk analysis and uncertainty resolution, reliability engineering and the analysis of thermal-hydraulic transients. (Nogami, K.)

Status of nuclear safety research - 2000

International Nuclear Information System (INIS)

Sobajima, Makoto; Sasajima, Hideo; Umemoto, Michitaka; Yamamoto, Toshihiro; Tanaka, Tadao; Togashi, Yoshihiro; Nakata, Masahito

2000-11-01

The nuclear safety research at JAERI is performed in accordance with the long term plan on nuclear research, development and use and the safety research yearly plan determined by the government and under close relationship to the related departments in and around the Nuclear Safety Research Center. The criticality accident having occurred in Tokai-mura in 1999 has been the highest level nuclear accident in Japan and ensuring safety in whole nuclear cycle is severely questioned. The causes of such an accident have to be clarified not only technical points but also organizational points, and it is extremely important to make efforts in preventing recurrence, to fulfill emergency plan and to improve the safety of whole nuclear fuel cycle for restoring the reliability by the people to nuclear energy system. The fields of conducting safety research are engineering safety research on reactor facilities and nuclear fuel cycle facilities including research on radioactive waste processing and disposal and research and development on future technology for safety improvement. Also, multinational cooperation and bilateral cooperation are promoted in international research organizations in the center to internationally share the recognition of world-common issues of nuclear safety and to attain efficient promotion of research and effective utilization of research resources. (author)

Report to Congress on NRC emergency communications

Energy Technology Data Exchange (ETDEWEB)

NONE

1980-09-01

The accident at Three Mile Island highlighted the need for improved communications among the NRC and other organizations which respond to such emergencies. This report summarizes the communication problems identified by several major review groups after the accident, the status of corrective actions, and NRC plans to improve communications still further. (author)

Nuclear safety research

International Nuclear Information System (INIS)

1996-01-01

The topics 'Large-sized PWR-NPP Safety Techniques Research',and 'The Key Techniques Research on the Safety Supervision and Control for Operation of Nuclear Installations' have been adopted as an apart of 'the National 9th five Year Programs for Tacking the Key Scientific and Technical Topics' which are organized by the State Planning Commission (SPC) and State Science and Technology Commission (SSTC) respectively, and have obtained a financial support from them. To play a better role with the limited fund, the NNSA laid special stress on selecting key sub-topics on nuclear safety, and carefully choosing units which would undertake sub-topics and signing technical contracts with them

NRC comprehensive records disposition schedule

International Nuclear Information System (INIS)

1982-07-01

Effective January 1, 1982, NRC will institute records retention and disposal practices in accordance with the approved Comprehensive Records Disposition Schedule (CRDS). CRDS is comprised of NRC Schedules (NRCS) 1 to 4 which apply to the agency's program or substantive records and General Records Schedules (GRS) 1 to 22 which apply to housekeeping or facilitative records. The schedules are assembled functionally/organizationally to facilitate their use. Preceding the records descriptions and disposition instructions for both NRCS and GRS, there are brief statements on the organizational units which accumulate the records in each functional area, and other information regarding the schedules' applicability

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

Reassessment of the NRC's program for protecting allegers against retaliation

International Nuclear Information System (INIS)

1994-01-01

On July 6, 1993, the Nuclear Regulatory Commission's (NRC's) Executive Director for Operations established a review team to reassess the NRC's program for protecting allegers against retaliation. The team evaluated the current system, and solicited comments from various NRC offices, other Federal agencies, licensees, former allegers, and the public. This report is subject to agency review. The report summarizes current processes and gives an overview of current problems. It discusses: (1) ways in which licensees can promote a quality-conscious work environment, in which all employees feel free to raise concerns without fear of retaliation; (2) ways to improve the NRC's overall handling of allegations; (3) the NRC's involvement in the Department of Labor process; (4) related NRC enforcement practices; and (5) methods other than investigation and enforcement that may be useful in treating allegations of potential or actual discrimination. Recommendations are given in each area

NRC comprehensive records disposition schedule. Revision 3

International Nuclear Information System (INIS)

1998-02-01

Title 44 US Code, ''Public Printing and Documents,'' regulations issued by the General Service Administration (GSA) in 41 CFR Chapter 101, Subchapter B, ''Management and Use of Information and Records,'' and regulations issued by the National Archives and Records Administration (NARA) in 36 CFR Chapter 12, Subchapter B, ''Records Management,'' require each agency to prepare and issue a comprehensive records disposition schedule that contains the NARA approved records disposition schedules for records unique to the agency and contains the NARA's General Records Schedules for records common to several or all agencies. The approved records disposition schedules specify the appropriate duration of retention and the final disposition for records created or maintained by the NRC. NUREG-0910, Rev. 3, contains ''NRC's Comprehensive Records Disposition Schedule,'' and the original authorized approved citation numbers issued by NARA. Rev. 3 incorporates NARA approved changes and additions to the NRC schedules that have been implemented since the last revision dated March, 1992, reflects recent organizational changes implemented at the NRC, and includes the latest version of NARA's General Records Schedule (dated August 1995)

NRC comprehensive records disposition schedule. Revision 3

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

Title 44 US Code, ``Public Printing and Documents,`` regulations issued by the General Service Administration (GSA) in 41 CFR Chapter 101, Subchapter B, ``Management and Use of Information and Records,`` and regulations issued by the National Archives and Records Administration (NARA) in 36 CFR Chapter 12, Subchapter B, ``Records Management,`` require each agency to prepare and issue a comprehensive records disposition schedule that contains the NARA approved records disposition schedules for records unique to the agency and contains the NARA`s General Records Schedules for records common to several or all agencies. The approved records disposition schedules specify the appropriate duration of retention and the final disposition for records created or maintained by the NRC. NUREG-0910, Rev. 3, contains ``NRC`s Comprehensive Records Disposition Schedule,`` and the original authorized approved citation numbers issued by NARA. Rev. 3 incorporates NARA approved changes and additions to the NRC schedules that have been implemented since the last revision dated March, 1992, reflects recent organizational changes implemented at the NRC, and includes the latest version of NARA`s General Records Schedule (dated August 1995).

Safety research program of NUCEF

International Nuclear Information System (INIS)

Naito, Y.

1996-01-01

To contribute the safety and establishment of advanced technologies in the area of nuclear fuel cycle, Japan Atomic Energy Research Institute (JAERI) has constructed a new research facility NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) as the center for the research and development, particularly on the reprocessing technology and transuranium (TRU) waste management. NUCEF consist of three buildings, administration building, experiment building A and B. Building A has two experiment facilities STACY (Static Experiment Critical Facility) and TRACY (Transient Experiment Critical Facility). The experiment building B is referred to as BECKY (Back-end Fuel Cycle Key Elements Research Facility). Researches on the reprocessing and the waste management are carried out with spent fuels, high-level liquid waste, TRU etc. in the α γ cell and glove boxes. NUCEF was constructed with the following aims. Using STACY and TRACY, are aimed, (1) research on advanced technology for criticality safety control, (2) reconfirmation of criticality safety margin of the Rokkasho reprocessing plant. Using BECKY, are aimed, (1) research on advanced technology of reprocessing process, (2) contribution to develop the scenario for TRU waste disposal, (3) development of new technology for TRU partitioning and volume reduction of radioactive waste. To realize the above aims, following 5 research subjects are settled in NUCEF, (1) Criticality safety research, (2) Research on safety and advanced technology of fuel reprocessing, (3) Research on TRU waste management, (4) Fundamental research on TRU chemistry, (5) Key technology development for TRU processing. (author)

Progress of nuclear safety research - 2005

International Nuclear Information System (INIS)

Anoda, Yoshinari; Amaya, Masaki; Saito, Junichi; Sato, Atsushi; Sono, Hiroki; Tamaki, Hitoshi; Tonoike, Kotaro; Nemoto, Yoshiyuki; Motoki, Yasuo; Moriyama, Kiyofumi; Yamaguchi, Tetsuji; Araya, Fumimasa

2006-03-01

The Japan Atomic Energy Research Institute (JAERI), one of the predecessors of the Japan Atomic Energy Agency (JAEA), had conducted nuclear safety research primarily at the Nuclear Safety Research Center in close cooperation with the related departments in accordance with the Long Term Plan for Development and Utilization of Nuclear Energy and Five-Years Program for Safety Research issued by the Japanese government. The fields of conducting safety research at JAERI were the engineering safety of nuclear power plants and nuclear fuel cycle facilities, and radioactive waste management as well as advanced technology for safety improvement or assessment. Also, JAERI had conducted international collaboration to share the information on common global issues of nuclear safety and to supplement own research. Moreover, when accidents occurred at nuclear facilities, JAERI had taken a responsible role by providing experts in assistance to conducting accident investigations or emergency responses by the government or local government. These nuclear safety research and technical assistance to the government have been taken over as an important role by JAEA. This report summarizes the nuclear safety research activities of JAERI from April 2003 through September 2005 and utilized facilities. (author)

Progress of nuclear safety research-2004

International Nuclear Information System (INIS)

Anoda, Yoshinari; Ebine, Noriya; Chuto, Toshinori; Sato, Satoshi; Ishikawa, Jun; Yamamoto, Toshihiro; Munakata, Masahiro; Asakura, Toshihide; Yamaguchi, Tetsuji; Kida, Takashi; Matsui, Hiroki; Haneishi, Akihiro; Araya, Fumimasa

2005-03-01

JAERI is conducting nuclear safety research primarily at the Nuclear Safety Research Center in close cooperation with the related departments in accordance with the Long Term Plan for Development and Utilization of Nuclear Energy and Annual Plan for Safety Research issued by the Japanese government. The fields of conducting safety research at JAERI are the engineering safety of nuclear power plants and nuclear fuel cycle facilities, and radioactive waste management as well as advanced technology for safety improvement or assessment. Also, JAERI has conducted international collaboration to share the information on common global issues of nuclear safety and to supplement own research. Moreover, when accidents occurred at nuclear facilities, JAERI has taken a responsible role by providing technical experts and investigation for assistance to the government or local public body. This report summarizes the nuclear safety research activities of JAERI from April 2002 through March 2004 and utilized facilities. (author)

NRC Regulatory Agenda

International Nuclear Information System (INIS)

1992-07-01

This document compilation of all rules on which the NRC has recently completed action, or has proposed action, or is considering action, and all petitions for rule making which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

Annual safety research report, JFY 2012

International Nuclear Information System (INIS)

2013-08-01

As for the regulatory issues the governments or JNES considered necessary, JNES had compiled 'safety research plan' in respective research areas necessary for solving the regulatory issues (safety research needs) and was conducting safety research to obtain the results, etc. Safety research areas, subjects and research projects were as follows: design review of nuclear power plant (5 subjects and each subject having several research projects totaled 20), control management of nuclear power plant (3 subjects and each subject having several research projects totaled 6), nuclear fuel cycle (2 subjects and each subject having several research projects totaled 4), nuclear fuel cycle backend (2 subjects and each subject having several research projects totaled 6), nuclear emergency preparedness and response (3 subjects and each subject having several research projects totaled 7) and bases of nuclear safety technology (3 subjects and each subject having several research projects totaled 6). In addition to these 49 research projects of 18 subjects in 6 areas, JNES worked on 19 research projects of 7 subjects in added areas (specific research projects on of the disaster at Fukushima Daiichi NPP accident and other challenges JNES considered necessary) in JFY 2012. This annual safety research report summarized respective achievements and state of regulatory tools necessary for solving regulatory issues according to the safety research plan, JFY 2012 Edition as well as the situation of the reflection for the safety regulations, and also described 16 research projects of 4 subjects: examination for new safety regulation (8 research projects), development of newly necessary evaluation methods (one research project), evaluation of the validity for the work for convergence at Fukushima Daiichi NPP accident (4 research project) and horizontal development to other nuclear power plants (3 research projects), and 3 research projects of 3 subjects as other challenges. A list of JNES

NRC's object-oriented simulator instructor station

International Nuclear Information System (INIS)

Griffin, J.I.; Griffin, J.P.

1995-06-01

As part of a comprehensive simulator upgrade program, the simulator computer systems associated with the Nuclear Regulatory Commission's (NRC) nuclear power plant simulators were replaced. Because the original instructor stations for two of the simulators were dependent on the original computer equipment, it was necessary to develop and implement new instructor stations. This report describes the Macintosh-based Instructor Stations developed by NRC engineers for the General Electric (GE) and Babcock and Wilcox (B and W) simulators

NRC Regulatory Agenda: Quarterly report, October--December 1988

International Nuclear Information System (INIS)

1989-01-01

The NRC Regulatory Agenda is a compilation of all rules which the NRC has proposed or is considering action on, and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission

Trends in fuel reprocessing safety research

International Nuclear Information System (INIS)

Tsujino, Takeshi

1981-01-01

With the operation of a fuel reprocessing plant in the Power Reactor and Nuclear Fuel Development Corporation (PNC) and the plan for a second fuel reprocessing plant, the research on fuel reprocessing safety, along with the reprocessing technology itself, has become increasingly important. As compared with the case of LWR power plants, the safety research in this field still lags behind. In the safety of fuel reprocessing, there are the aspects of keeping radiation exposure as low as possible in both personnel and local people, the high reliability of the plant operation and the securing of public safety in accidents. Safety research is then required to establish the safety standards and to raise the rate of plant operation associated with safety. The following matters are described: basic ideas for the safety design, safety features in fuel reprocessing, safety guideline and standards, and safety research for fuel reprocessing. (J.P.N.)

Comparisons of ANS, ASME, AWS, and NFPA standards cited in the NRC standard review plan, NUREG-0800, and related documents

International Nuclear Information System (INIS)

Ankrum, A.R.; Bohlander, K.L.; Gilbert, E.R.; Spiesman, J.B.

1995-11-01

This report provides the results of comparisons of the cited and latest versions of ANS, ASME, AWS and NFPA standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC's Standard Review Plan Update and Development Program. Significant changes to the standards, from the cited version to the latest version, are described and discussed in a tabular format for each standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the standard. The results and recommendations presented in this document have not been subjected to NRC staff review

NRC regulatory agenda

International Nuclear Information System (INIS)

1993-02-01

This document is a compilation of all rules on which the NRC has recently completed action, or has proposed action, or is considered action, and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

NRC regulatory agenda

International Nuclear Information System (INIS)

1992-11-01

This document provides a compilation of all rules on which the NRC has recently completed action, or has proposed action, or is considering action, and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

NRC plan for cleanup operations at Three Mile Island Unit 2

International Nuclear Information System (INIS)

Lo, R.; Snyder, B.

1982-02-01

This NRC Plan, which defines NRC's functional role in cleanup operations at Three Mile Island Unit 2 and outlines NRC's regulatory responsibilities in fulfilling this role, is the first revision to the initial plan issued in July 1980 (NUREG-0698). Since 1980, a number of policy developments have occurred which will have an impact on the course of cleanup operations. This revision reflects these developments in the area of NRC's review and approval process with regard to cleanup operations as well as NRC's interface with the Department of Energy's involvement in the cleanup and waste disposal. This revision is also intended to update the cleanup schedule by presenting the cleanup progress that has taken place and NRC's role in ongoing and future cleanup activities

NRC comprehensive records disposition schedule

International Nuclear Information System (INIS)

1992-03-01

Title 44 United States Code, ''Public Printing and Documents,'' regulations cited in the General Services Administration's (GSA) ''Federal Information Resources Management Regulations'' (FIRMR), Part 201-9, ''Creation, Maintenance, and Use of Records,'' and regulation issued by the National Archives and Records Administration (NARA) in 36 CFR Chapter XII, Subchapter B, ''Records Management,'' require each agency to prepare and issue a comprehensive records disposition schedule that contains the NARA approved records disposition schedules for records unique to the agency and contains the NARA's General Records Schedules for records common to several or all agencies. The approved records disposition schedules specify the appropriate duration of retention and the final disposition for records created or maintained by the NRC. NUREG-0910, Rev. 2, contains ''NRC's Comprehensive Records Disposition Schedule,'' and the original authorized approved citation numbers issued by NARA. Rev. 2 totally reorganizes the records schedules from a functional arrangement to an arrangement by the host office. A subject index and a conversion table have also been developed for the NRC schedules to allow staff to identify the new schedule numbers easily and to improve their ability to locate applicable schedules

Memorandum of Understanding Between U.S. EPA Superfund and U.S. NRC

International Nuclear Information System (INIS)

Walker, Stuart

2008-01-01

The Environmental Protection Agency (EPA) Office of Superfund Remediation and Technology Innovation (OSRTI) and the Nuclear Regulatory Commission (NRC) are responsible for implementing the 'Memorandum of Understanding Between the Environmental Protection Agency and the Nuclear Regulatory Commission: Consultation and Finality on Decommissioning and Decontamination of Contaminated Sites'. This paper provides a brief overview of the origin of the Memorandum of Understanding (MOU), the major features of the MOU, and how the MOU has been implemented site specifically. EPA and NRC developed the MOU in response to direction from the House Committee on Appropriations to EPA and NRC to work together to address the potential for dual regulation. The MOU was signed by EPA on September 30, 2002 and NRC on October 9, 2002. The two agencies had worked on the MOU since March 2000. While both EPA and NRC have statutory authority to clean up these sites, the MOU provides consultation procedures between EPA and NRC to eliminate dual regulation. Under the MOU, EPA and NRC identified the interactions of the two agencies for the decommissioning and decontamination of NRC-licensed sites and the ways in which those responsibilities will be exercised. Except for Section VI, which addresses corrective action under the Resource Conservation and Recovery Act (RCRA), this MOU is limited to the coordination between EPA, when acting under its CERCLA authority, and NRC, when a facility licensed by the NRC is undergoing decommissioning, or when a facility has completed decommissioning, and the NRC has terminated its license. EPA believes that implementation of the MOU between the two agencies will ensure that future confusion about dual regulation does not occur regarding the cleanup and reuse of NRC-licensed sites. NRC and EPA have so far exchanged MOU consultation letters on eight NRC-licensed sites. EPA has responded to each consultation request with a letter expressing its views on actions

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

Energy Technology Data Exchange (ETDEWEB)

Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

1994-02-01

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

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

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

Review of NRC Commission Papers on Regulatory Basis for Licensing and Regulating Reprocessing Facilities

Energy Technology Data Exchange (ETDEWEB)

Park, Jae Yeong; Shin, Hyeong Ki [KINS, Daejeon (Korea, Republic of)

2016-05-15

Spent nuclear fuel (SNF) accumulated in nuclear power plant has been a serious issue in most countries with operating nuclear power plants. Direct disposal of SNF could be a solution of the problem but many countries including the Republic of Korea have had a hard time selecting a site for high level waste repository because of low public acceptance. SNF recycling technologies consisting of reprocessing and transmutation have been developed so as to reduce the final volume of the disposed radioactive waste and to diminish the radiotoxicity of the waste. The Republic of Korea is now developing pyroprocessing and sodium-cooled fast reactor (SFR) technology to be used for the recycling of the wastes. KAERI has a plan to construct a pyroprocessing facility with a capacity of 30 tHM/y and a facility manufacturing TRU fuel for SFR by 2025. However, to license these facility and secure the safety, the current regulatory system related to SNF treatment needs to be improved and amended since the system has been developed focusing on facilities to examine irradiated nuclear materials. Status of reprocessing facility regulations developed by U.S.NRC was reviewed based on SECY papers. U.S.NRC has approved the development of a new rule referred to nationally as '10CFR Part 7x'. Existing 10CFR 50 and 70 has been evolved mainly for nuclear power plants and fuel cycle facilities whose radiological hazard is much lower than reprocessing plants respectively. U.S.NRC also derived many regulatory gaps including safety assessment methods, technical specification, general design criteria and waste classification and continue to develop the regulatory framework limited in scope to the resolution of Gap 5.

Researchers' Roles in Patient Safety Improvement.

Science.gov (United States)

Pietikäinen, Elina; Reiman, Teemu; Heikkilä, Jouko; Macchi, Luigi

2016-03-01

In this article, we explore how researchers can contribute to patient safety improvement. We aim to expand the instrumental role researchers have often occupied in relation to patient safety improvement. We reflect on our own improvement model and experiences as patient safety researchers in an ongoing Finnish multi-actor innovation project through self-reflective narration. Our own patient safety improvement model can be described as systemic. Based on the purpose of the innovation project, our improvement model, and the improvement models of the other actors in the project, we have carried out a wide range of activities. Our activities can be summarized in 8 overlapping patient safety improvement roles: modeler, influencer, supplier, producer, ideator, reflector, facilitator, and negotiator. When working side by side with "practice," researchers are offered and engage in several different activities. The way researchers contribute to patient safety improvement and balance between different roles depends on the purpose of the study, as well as on the underlying patient safety improvement models. Different patient safety research paradigms seem to emphasize different improvement roles, and thus, they also face different challenges. Open reflection on the underlying improvement models and roles can help researchers with different backgrounds-as well as other actors involved in patient safety improvement-in structuring their work and collaborating productively.

USNRC licensing process as related to nuclear criticality safety

International Nuclear Information System (INIS)

Ketzlach, N.

1987-01-01

The U.S. Code of Federal Regulations establishes procedures and criteria for the issuance of licenses to receive title to, own, acquire, deliver, receive, possess, use, and initially transfer special nuclear material; and establishes and provides for the terms and conditions upon which the Nuclear Regulatory Commission (NRC) will issue such licenses. Section 70.22 of the regulations, ''Contents of Applications'', requires that applications for licenses contain proposed procedures to avoid accidental conditions of criticality. These procedures are elements of a nuclear criticality safety program for operations with fissionable materials at fuels and materials facilities (i.e., fuel cycle facilities other than nuclear reactors) in which there exists a potential for criticality accidents. To assist the applicant in providing specific information needed for a nuclear criticality safety program in a license application, the NRC has issued regulatory guides. The NRC requirements for nuclear criticality safety include organizational, administrative, and technical requirements. For purely technical matters on nuclear criticality safety these guides endorse national standards. Others provide guidance on the standard format and content of license applications, guidance on evaluating radiological consequences of criticality accidents, or guidance for dealing with other radiation safety issues. (author)

Progress of nuclear safety research. 2002

Energy Technology Data Exchange (ETDEWEB)

Anoda, Yoshinari; Kudo, Tamotsu; Tobita, Tohru (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] (and others)

2002-11-01

JAERI is conducting nuclear safety research primarily at the Nuclear Safety Research Center in close cooperation with the related departments in accordance with the Long Term Plan for Development and Utilization of Nuclear Energy and Annual Plan for Safety Research issued by the Japanese government. The fields of conducting safety research at JAERI are the engineering safety of nuclear power plants and nuclear fuel cycle facilities, and radioactive waste management as well as advanced technology for safety improvement or assessment. Also, JAERI has conducted international collaboration to share the information on common global issues of nuclear safety and to supplement own research. Moreover, when accidents occurred at nuclear facilities, JAERI has taken a responsible role by providing technical experts and investigation for assistance to the government or local public body. This report summarizes the nuclear safety research activities of JAERI from April 2000 through April 2002 and utilized facilities. This report also summarizes the examination of the ruptured pipe performed for assistance to the Nuclear and Industrial Safety Agency (NISA) for investigation of the accident at the Hamaoka Nuclear Power Station Unit-1 on November, 2001. (author)

10 CFR 2.1316 - Authority and role of NRC staff.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Authority and role of NRC staff. 2.1316 Section 2.1316... ORDERS Procedures for Hearings on License Transfer Applications § 2.1316 Authority and role of NRC staff. (a) During the pendency of any hearing under this subpart, consistent with the NRC staff's findings...

NRC performance assessment program

International Nuclear Information System (INIS)

Coplan, S.M.

1986-01-01

The U.S. Nuclear Regulatory Commission's (NRC) performance assessment program includes the development of guidance to the U.S. Department of Energy (DOE) on preparation of a license application and on conducting the studies to support a license application. The nature of the licensing requirements of 10 CFR Part 60 create a need for performance assessments by the DOE. The NRC and DOE staffs each have specific roles in assuring the adequacy of those assessments. Performance allocation is an approach for determining what testing and analysis will be needed during site characterization to assure that an adequate data base is available to support the necessary performance assessments. From the standpoint of establishing is implementable methodology, the most challenging performance assessment needed for licensing is the one that will be used to determine compliance with the U.S. Environmental Protection Agency's (EPA) containment requirement

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

International Nuclear Information System (INIS)

1997-01-01

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

IAEA programme on research reactor safety

International Nuclear Information System (INIS)

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

1995-01-01

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

Results from the Nuclear Plant Aging Research Program: Their use in inspection activities

International Nuclear Information System (INIS)

Gunther, W.; Taylor, J.

1990-09-01

The US NCR's Nuclear Plant Aging Research (NPAR) Program has determined the susceptibility to aging of components and systems, and the potential for aging to impact plant safety and availability. The NPAR Program also identified methods for detecting and mitigating aging in components. This report describes the NPAR results which can enhance NRC inspection activities. Recommendations are provided for communicating pertinent information to NRC inspectors. These recommendations are based on a detailed assessment of the NRC's Inspection Program, and feedback from resident and regional inspectors as described within. Examples of NPAR report summaries and aging inspection guides for components and systems are included. 13 refs., 3 figs., 4 tabs

A qualified safety I and C for application in reactors of all kinds

International Nuclear Information System (INIS)

Stimler, M.

2001-01-01

for instrumentation and control systems in U.S. nuclear poer plants. Approval by the NRC reflects the results of a detailed technical and regulatory review, following which the NRC Staff concluded, '...the design of the TXS system is acceptable for all safety-related I and C applications and meets the relevant regulatory requirements.' By now more than 20 modernisation projects of Safety I and C have successfully been completed in most of the European countries that operate nuclear power plants and also in China and the USA projects are underway. It is remarkable, that in Germany, after first applications in safety related systems in nuclear power plants, the decision was made, to install TELEPERM XS in the research reactor FRM-II as protection system. The licensing process followed very much the guidelines of power reactors and was so far successful. The equipment has undergone a rigid test program in the test field and is ready for the startup tests at site. The presentation focuses on the special licensing process and explains why it was a good decision to utilize a system proven in commercial power reactors in a research reactor

Building beef cow nutritional programs with the 1996 NRC beef cattle requirements model.

Science.gov (United States)

Lardy, G P; Adams, D C; Klopfenstein, T J; Patterson, H H

2004-01-01

Designing a sound cow-calf nutritional program requires knowledge of nutrient requirements, diet quality, and intake. Effectively using the NRC (1996) beef cattle requirements model (1996NRC) also requires knowledge of dietary degradable intake protein (DIP) and microbial efficiency. Objectives of this paper are to 1) describe a framework in which 1996NRC-applicable data can be generated, 2) describe seasonal changes in nutrients on native range, 3) use the 1996NRC to predict nutrient balance for cattle grazing these forages, and 4) make recommendations for using the 1996NRC for forage-fed cattle. Extrusa samples were collected over 2 yr on native upland range and subirrigated meadow in the Nebraska Sandhills. Samples were analyzed for CP, in vitro OM digestibility (IVOMD), and DIP. Regression equations to predict nutrients were developed from these data. The 1996NRC was used to predict nutrient balances based on the dietary nutrient analyses. Recommendations for model users were also developed. On subirrigated meadow, CP and IVOMD increased rapidly during March and April. On native range, CP and IVOMD increased from April through June but decreased rapidly from August through September. Degradable intake protein (DM basis) followed trends similar to CP for both native range and subirrigated meadow. Predicted nutrient balances for spring- and summer-calving cows agreed with reported values in the literature, provided that IVOMD values were converted to DE before use in the model (1.07 x IVOMD - 8.13). When the IVOMD-to-DE conversion was not used, the model gave unrealistically high NE(m) balances. To effectively use the 1996NRC to estimate protein requirements, users should focus on three key estimates: DIP, microbial efficiency, and TDN intake. Consequently, efforts should be focused on adequately describing seasonal changes in forage nutrient content. In order to increase use of the 1996NRC, research is needed in the following areas: 1) cost-effective and

Utilizing the National Research Council's (NRC) Conceptual Framework for the Next Generation Science Standards (NGSS): A Self-Study in My Science, Engineering, and Mathematics Classroom

Science.gov (United States)

Corvo, Arthur Francis

Given the reality that active and competitive participation in the 21 st century requires American students to deepen their scientific and mathematical knowledge base, the National Research Council (NRC) proposed a new conceptual framework for K--12 science education. The framework consists of an integration of what the NRC report refers to as the three dimensions: scientific and engineering practices, crosscutting concepts, and core ideas in four disciplinary areas (physical, life and earth/spaces sciences, and engineering/technology). The Next Generation Science Standards (NGSS ), which are derived from this new framework, were released in April 2013 and have implications on teacher learning and development in Science, Technology, Engineering, and Mathematics (STEM). Given the NGSS's recent introduction, there is little research on how teachers can prepare for its release. To meet this research need, I implemented a self-study aimed at examining my teaching practices and classroom outcomes through the lens of the NRC's conceptual framework and the NGSS. The self-study employed design-based research (DBR) methods to investigate what happened in my secondary classroom when I designed, enacted, and reflected on units of study for my science, engineering, and mathematics classes. I utilized various best practices including Learning for Use (LfU) and Understanding by Design (UbD) models for instructional design, talk moves as a tool for promoting discourse, and modeling instruction for these designed units of study. The DBR strategy was chosen to promote reflective cycles, which are consistent with and in support of the self-study framework. A multiple case, mixed-methods approach was used for data collection and analysis. The findings in the study are reported by study phase in terms of unit planning, unit enactment, and unit reflection. The findings have implications for science teaching, teacher professional development, and teacher education.

NRC Regulatory Agenda: Quarterly report, July-September 1987

International Nuclear Information System (INIS)

1987-11-01

The NRC Regulatory agenda is a compilation of all rules on which the NRC has proposed or is considering action and all petitions for rulemaking which have been received by the commission and are pending disposition by the commission. The regulatory agenda is updated and issued each quarter

NRC regulatory agenda: Quarterly report, April--June 1988

International Nuclear Information System (INIS)

1988-08-01

The NRC Regulatory Agenda is a compilation of all rules on which the NRC has proposed or is considering action and all petitions for rulemaking which have been received by the Commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

Nuclear safety research in HGF 2011

International Nuclear Information System (INIS)

Tromm, Walter

2012-01-01

After the events at the Japanese nuclear power plant of Fukushima Daiichi, the German federal government decided that Germany will give up electricity generation from nuclear power within a decade. The last reactor will be disconnected from the power grid in 2022. Helping to make this opt-out as safe as possible is one of the duties of the Helmholtz Association with its Nuclear Safety Research Program within the Energy Research Area. Also the demolition of nuclear power plants and the repository problem will keep society, and thus also research, busy for a number of decades to come. Giving up electricity production from nuclear power thus must not mean giving up the required nuclear technology competences. In the fields of reactor safety, demolition, final storage, radiation protection, and crisis management, in critical support of international developments, and for competent evaluation of nuclear facilities around Germany, these competences will be in demand far beyond the German opt-out. This is the reason why the final report by the Ethics Committee on 'Safe Energy Supply' emphasizes the importance of nuclear technology research. Close cooperation on national, European and international levels is indispensable in this effort. Also nuclear safety research in the Helmholtz Association is aligned with the challenges posed by the opt-out of the use of nuclear power. It is important that the high competences in the areas of plant safety and demolition, handling of radioactive waste, and safe final storage as well as radiation protection be preserved. The Nuclear Safety Research Program within the Energy Research Area of the Helmholtz Association therefore will continue studying scientific and technical aspects of the safety of nuclear reactors and the safety of nuclear waste management. These research activities are provident research conducted for society and must be preserved for a long period of time. The work is closely harmonized with the activities of the

Progress of nuclear safety research. 2003

International Nuclear Information System (INIS)

Anoda, Yoshinari; Amagai, Masaki; Tobita, Tohru

2004-03-01

JAERI is conducting nuclear safety research primarily at the Nuclear Safety Research Center in close cooperation with the related departments in accordance with the Long Term Plan for Development and Utilization of Nuclear Energy and Annual Plan for Safety Research issued by the Japanese government. The fields of conducting safety research at JAERI are the engineering safety of nuclear power plants and nuclear fuel cycle facilities, and radioactive waste management as well as advanced technology for safety improvement or assessment. Also, JAERI has conducted international collaboration to share the information on common global issues of nuclear safety and to supplement own research. Moreover, when accidents occurred at nuclear facilities, JAERI has taken a responsible role by providing technical experts and investigation for assistance to the government or local public body. This report summarizes the nuclear safety research activities of JAERI from April 2001 through March 2003 and utilized facilities. This report also summarizes the examination of the ruptured pipe performed for assistance to the Nuclear and Industrial Safety Agency (NISA) for investigation of the accident at the Hamaoka Nuclear Power Station Unit-1 on November, 2001, and the integrity evaluation of cracked core shroud of BWRs of the Tokyo Electric Power Company performed for assistance to the Nuclear Safety Commission in reviewing the evaluation reports by the licensees. (author)

Grinding the Antitesting Ax: More Bias than Evidence behind NRC Panel's Conclusions

Science.gov (United States)

Hanushek, Eric A.

2012-01-01

In all the acrimonious discussion surrounding No Child Left Behind Act of 2001 (NCLB), surprisingly little attention has been given to the actual impact of that legislation and other accountability systems on student performance. Now a reputable body, a committee set up by the National Research Council (NRC), the research arm of the National…

NRC concerns about steam generator tube U-bend failures

International Nuclear Information System (INIS)

Dillon, R.L.

1981-01-01

This paper concerns itself with genralized NRC regulatory policy regarding SGT failures and staff reports and opinions which may tend to influence the developing policy specific to U-bend failures. The most significant analysis at hand in predicting NRC policy on SGT U-bend failures is Marsh's Evaluation of Steam Generator Tube Rupture Events. Marsh sets out to describe and analyze the five steam generator tube ruptures that are known to NRC. All have occurred in the period 1975 to 1980

Safety implications of diesel generator aging management

International Nuclear Information System (INIS)

Hoopingarner, K.R.

1989-01-01

Significant safety improvements can be achieved in diesel-generator management related to aging, testing, and other important regulatory concerns. This paper reports on the progress of aging research related to nuclear service diesel generators, which developed data and information supporting the recommended safety improvements. The key to diesel-generator safety improvements is the development of a new balanced approach where testing, inspections, monitoring and trending, training, and maintenance all have appropriate importance. Safety improvement is projected in a management program that concurrently achieves three goals: first, the reduction of the fast-start stressor by regulatory and utility actions; second, the establishment of more appropriate testing and trending procedures; third, the adoption and use of reliability-centered maintenance activities. This paper describes the recommended safety improvements and the positive role of utility management in the process and outlines a new recommended regulatory approach. Diesel generator aging and wear is the subject of research sponsored by the Nuclear Plant Aging Research (NPAR) Program under the US Nuclear Regulatory Commission (NRC). Office of Nuclear Regulatory Research. The research was conducted by Pacific Northwest Laboratory (PNL), which is operated for the US Department of Energy by Battelle Memorial Institute. 4 refs., 1 fig., 1 tab

Reassessment of the NRC`s program for protecting allegers against retaliation

Energy Technology Data Exchange (ETDEWEB)

1994-01-01

On July 6, 1993, the Nuclear Regulatory Commission`s (NRC`s) Executive Director for Operations established a review team to reassess the NRC`s program for protecting allegers against retaliation. The team evaluated the current system, and solicited comments from various NRC offices, other Federal agencies, licensees, former allegers, and the public. This report is subject to agency review. The report summarizes current processes and gives an overview of current problems. It discusses: (1) ways in which licensees can promote a quality-conscious work environment, in which all employees feel free to raise concerns without fear of retaliation; (2) ways to improve the NRC`s overall handling of allegations; (3) the NRC`s involvement in the Department of Labor process; (4) related NRC enforcement practices; and (5) methods other than investigation and enforcement that may be useful in treating allegations of potential or actual discrimination. Recommendations are given in each area.

Energy Technology Division research summary 1997

International Nuclear Information System (INIS)

1997-01-01

The Energy Technology Division provides materials and engineering technology support to a wide range of programs important to the US Department of Energy. As shown on the preceding page, the Division is organized into ten sections, five with concentrations in the materials area and five in engineering technology. Materials expertise includes fabrication, mechanical properties, corrosion, friction and lubrication, and irradiation effects. Our major engineering strengths are in heat and mass flow, sensors and instrumentation, nondestructive testing, transportation, and electromechanics and superconductivity applications. The Division Safety Coordinator, Environmental Compliance Officers, Quality Assurance Representative, Financial Administrator, and Communication Coordinator report directly to the Division Director. The Division Director is personally responsible for cultural diversity and is a member of the Laboratory-wide Cultural Diversity Advisory Committee. The Division's capabilities are generally applied to issues associated with energy production, transportation, utilization or conservation, or with environmental issues linked to energy. As shown in the organization chart on the next page, the Division reports administratively to the Associate Laboratory Director (ALD) for Energy and Environmental Science and Technology (EEST) through the General Manager for Environmental and Industrial Technologies. While most of our programs are under the purview of the EEST ALD, we also have had programs funded under every one of the ALDs. Some of our research in superconductivity is funded through the Physical Research Program ALD. We also continue to work on a number of nuclear-energy-related programs under the ALD for Engineering Research. Detailed descriptions of our programs on a section-by-section basis are provided in the remainder of this book. This Overview highlights some major trends. Research related to the operational safety of commercial light water nuclear

Energy Technology Division research summary 1997.

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-21

The Energy Technology Division provides materials and engineering technology support to a wide range of programs important to the US Department of Energy. As shown on the preceding page, the Division is organized into ten sections, five with concentrations in the materials area and five in engineering technology. Materials expertise includes fabrication, mechanical properties, corrosion, friction and lubrication, and irradiation effects. Our major engineering strengths are in heat and mass flow, sensors and instrumentation, nondestructive testing, transportation, and electromechanics and superconductivity applications. The Division Safety Coordinator, Environmental Compliance Officers, Quality Assurance Representative, Financial Administrator, and Communication Coordinator report directly to the Division Director. The Division Director is personally responsible for cultural diversity and is a member of the Laboratory-wide Cultural Diversity Advisory Committee. The Division's capabilities are generally applied to issues associated with energy production, transportation, utilization or conservation, or with environmental issues linked to energy. As shown in the organization chart on the next page, the Division reports administratively to the Associate Laboratory Director (ALD) for Energy and Environmental Science and Technology (EEST) through the General Manager for Environmental and Industrial Technologies. While most of our programs are under the purview of the EEST ALD, we also have had programs funded under every one of the ALDs. Some of our research in superconductivity is funded through the Physical Research Program ALD. We also continue to work on a number of nuclear-energy-related programs under the ALD for Engineering Research. Detailed descriptions of our programs on a section-by-section basis are provided in the remainder of this book. This Overview highlights some major trends. Research related to the operational safety of commercial light water

NRC Reviewer Aid for Evaluating the Human Factors Engineering Aspects of Small Modular Reactors

International Nuclear Information System (INIS)

OHara, J.M.; Higgins, J.C.

2012-01-01

Small modular reactors (SMRs) are a promising approach to meeting future energy needs. Although the electrical output of an individual SMR is relatively small compared to that of typical commercial nuclear plants, they can be grouped to produce as much energy as a utility demands. Furthermore, SMRs can be used for other purposes, such as producing hydrogen and generating process heat. The design characteristics of many SMRs differ from those of current conventional plants and may require a distinct concept of operations (ConOps). The U.S. Nuclear Regulatory Commission (NRC) conducted research to examine the human factors engineering (HFE) and the operational aspects of SMRs. The research identified thirty potential human-performance issues that should be considered in the NRC's reviews of SMR designs and in future research activities. The purpose of this report is to support NRC HFE reviewers of SMR applications by identifying some of the questions that can be asked of applicants whose designs have characteristics identified in the issues. The questions for each issue were identified and organized based on the review elements and guidance contained in Chapter 18 of the Standard Review Plan (NUREG-0800), and the Human Factors Engineering Program Review Model (NUREG-0711).

IRSN safety research carried out for reviewing safety cases

International Nuclear Information System (INIS)

Serres, Ch.

2010-01-01

Christophe Serres from IRSN (France) described the independent role of the IRSN regarding research related to nuclear safety in the context of the French Planning Act of 28 June 2006 foreseeing a licence application to be submitted in 2015 for the creation of a deep geological repository. IRSN research programme is organised along research activities devoted to addressing independently-identified k ey safety issues . These 'key issues' should also be of prime concern for the implementer since they relate to the demonstration of the overall safety of the repository, and the level of funding that the implementer should afford to research activities of concern for safety. He explained that the quality and independency of the research programme carried out by IRSN allow building and improving a set of scientific knowledge and technical skills that serves the public mission of delivering technical appraisal and advice, e.g., on behalf of the national safety authority. In particular they contribute to improving the decisional process by making possible scientific dialogue with stakeholders independently from regulator or implementer. The current IRSN R and D programme is developed along the following lines: - Test the adequacy of experimental methods for which feedback is not sufficient. - Develop basic scientific knowledge in the fields where there is a need for better understanding of complex phenomena and interactions. - Develop and use numerical modelling tools to support studies on complex phenomena and interactions. - Perform specific experimental tests aiming at assessing the key parameters that may warrant the performances of the different components of the repository. These studies are carried out by means of experiments performed either at IRSN surface laboratories, or in the Tournemire Experimental Station (TES), an underground facility operated by IRSN in the south-east of France. Targeted actions on research related to operational safety and reversibility

Environment and safety research status report: 1993

International Nuclear Information System (INIS)

1993-03-01

The 1993 status report discusses ongoing and planned research activities in the GRI Environment and Safety Program. The objectives and goals, accomplishments, and strategy along with the basis for each project area are presented for the supply, end use, and gas operations subprograms. Within the context of these subprograms, contract status summaries under their conceptual titles are given for the following project areas: Gas Supply Environmental and Safety Research, Air Quality Research, End Use Equipment Safety Research, Gas Operations Safety Research, Liquefied Natural Gas, Safety Research, and Gas Operations Environmental Research

Reactor safety research and safety technology. Pt. 2

International Nuclear Information System (INIS)

Theenhaus, R.; Wolters, J.

1987-01-01

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

Perspectives on reactor safety

Energy Technology Data Exchange (ETDEWEB)

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

1994-03-01

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

Perspectives on reactor safety

International Nuclear Information System (INIS)

Haskin, F.E.

1994-03-01

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

Reactor Safety Assessment System: a situation assessment aid for USNRC emergency response

International Nuclear Information System (INIS)

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

1985-04-01

The Reactor Safety Assessment System is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base which uses the parametric values, the known operator actions and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant. 5 figs

Applications of probabilistic risk analysis in nuclear criticality safety design

International Nuclear Information System (INIS)

Chang, J.K.

1992-01-01

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

Evaluating software for safety systems in nuclear power plants

International Nuclear Information System (INIS)

Lawrence, J.D.; Persons, W.L.; Preckshot, G.G.; Gallagher, J.

1994-01-01

In 1991, LLNL was asked by the NRC to provide technical assistance in various aspects of computer technology that apply to computer-based reactor protection systems. This has involved the review of safety aspects of new reactor designs and the provision of technical advice on the use of computer technology in systems important to reactor safety. The latter includes determining and documenting state-of-the-art subjects that require regulatory involvement by the NRC because of their importance in the development and implementation of digital computer safety systems. These subjects include data communications, formal methods, testing, software hazards analysis, verification and validation, computer security, performance, software complexity and others. One topic software reliability and safety is the subject of this paper

Fluorine--a current literature review. An NRC and ATSDR based review of safety standards for exposure to fluorine and fluorides.

Science.gov (United States)

Prystupa, Jeff

2011-02-01

A review of the literature of the element fluorine and its bonded-form, fluoride, was undertaken. Generally regarded as safe, an expanding body of literature reveals that fluoride's toxicity has been unappreciated, un-scrutinized, and hidden for over 70 years. The context for the literature search and review was an environmental climate-change study, which demonstrated widespread fluoride contamination by smokestack emissions from coal-fired electricity-generating plants. The objective of this review is to educate and inform regarding the ubiquitous presence and harmful nature of this now ever-present corrosive and reactive toxin. Methods include examination of national health agency reviews, primarily the National Research Council (NRC), Agency for Toxic Substances & Disease Registry (ATSDR), standard medical toxicology references, text books, as well as reports and documents from both private and public research as well as consumer-based NGOs. Study criteria were chosen for relevancy to the subject of the toxicity of fluoride. Fluoride is the extreme electron scavenger, the most corrosive of all elements, as well as the most-reactive. Fluoride appears to attack living tissues, via several mechanisms. Fluoride renders strong evidence that it is a non-biological chemical, demonstrating no observed beneficial function or role in organic chemistry, beyond use as a pesticide or insecticide. Fluorine has a strong role to play in industry, having been utilized extensively in metals, plastics, paints, aluminium, steel, and uranium production. Due to its insatiable appetite for calcium, fluorine and fluorides likely represent a form of chemistry that is incompatible with biological tissues and organ system functions. Based on an analysis of the affects of fluoride demonstrated consistently in the literature, safe levels have not been determined nor standardized. Mounting evidence presents conflicting value to its presence in biological settings and applications. Evidence

NRC Regulatory Agenda quarterly report, July--September 1993

International Nuclear Information System (INIS)

1993-10-01

The NRC Regulator Agenda is a compilation of all rules on which the NRC has recently completed action, or has proposed action, or is considering action, and all petitions for rulemaking which have been received by the commission and are pending disposition by the Commission. The Regulatory Agenda is updated and issued each quarter

NRC nuclear-plant-analyzer concept and status at INEL

International Nuclear Information System (INIS)

Aguilar, F.; Wagner, R.J.

1982-01-01

The Office of Research of the US NRC has proposed development of a software-hardware system called the Nuclear Plant Analyzer (NPA). This paper describes how we of the INEL envision the nuclear-plant analyzer. The paper also describes a pilot RELAP5 plant-analyzer project completed during the past year and current work. A great deal of analysis is underway to determine nuclear-steam-system response. System transient analysis being so complex, there is the need to present analytical results in a way that interconnections among phenomena and all the nuances of the transient are apparent. There is the need for the analyst to dynamically control system calculations to simulate plant operation in order to perform what if studies as well as the need to perform system analysis within hours of a plant emergency to diagnose the state of the stricken plant and formulate recovery actions. The NRC-proposed nuclear-plant analyzer can meet these needs

Outline of criticality safety research project

International Nuclear Information System (INIS)

Kobayashi, Iwao; Tachimori, Shoichi; Suzaki, Takenori; Takeshita, Isao; Miyoshi, Yoshinori; Nakajima, Ken; Sakurai, Satoshi; Yanagisawa, Hiroshi

1987-01-01

As the power generation capacity of LWRs in Japan increased, the establishment and development of nuclear fuel cycle have become the important subject. Conforming to the safety research project of the nation, the Japan Atomic Energy Research Institute has advanced the project of constructing a new research facility, that is, Nuclear Fuel Cycle Engineering Research Facility (NUCEF). In this facility, it is planned to carry out the research on criticality safety, upgraded reprocessing techniques, and the treatment and disposal of transuranium element wastes. In this paper, the subjects of criticality safety research and the research carried out with a criticality safety experiment facility which is expected to be installed in the NUCEF are briefly reported. The experimental data obtained from the criticality safety handbooks and published literatures in foreign countries are short of the data on the mixture of low enriched uranium and plutonium which is treated in the reprocessing of spent fuel from LWRs. The acquisition of the criticality data for various forms of fuel, the elucidation of the scenario of criticality accidents, and the soundness of the confinement system for gaseous fission products and plutonium are the main subjects. The Static Criticality Safety Facility, Transient Criticality Safety Facility and pulse column system are the main facilities. (Kako, I.)

78 FR 47805 - Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

Science.gov (United States)

2013-08-06

... Used in Safety Systems of Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Revision... Used in Safety Systems of Nuclear Power Plants.'' This RG endorses the Institute of Electrical and... nuclear power plants. ADDRESSES: Please refer to Docket ID NRC-2012-0195 when contacting the NRC about the...

NRC regualtory agenda. Semiannual report, July 1997--December 1997

International Nuclear Information System (INIS)

1998-02-01

The Regulatory Agenda is a semiannual compilation of all rules on which the NRC has recently completed action, or has proposed action, or is considering action, and of all petitions for rulemaking that the NRC has received that are pending disposition

Applying risk insights in US NRC reviews of integral pressurized water reactor designs

International Nuclear Information System (INIS)

Caruso, M.A.; Hilsmeier, T.; Kevern, T.A.

2012-01-01

In its Staff Requirements Memorandum (SRM) on COMGBJ-10-0004/COMGEA-10-0001, 'Use of Risk Insights to Enhance Safety Focus of Small Modular Reactor Reviews,' dated August 31, 2010 (ML102510405), the U.S. Nuclear Regulatory Commission (NRC) directed the NRC staff to more fully integrate the use of risk insights into pre-application activities and the review of small modular reactor (SMR) applications with near-term focus on integral pressurized water reactor (iPWR) designs. The Commission's objective is to align the review focus and resources with the risk-significant systems, structures, and components (SSCs) and other aspects of the design, that contribute most to safety in order to enhance the efficiency of the review process while still enabling a decision of reasonable assurance of the design's safety. The staff was directed to develop a design-specific, risk-informed review plan for each SMR to address pre-application and application review activities. The NRC staff submitted a response to the Commission which describes its approach for (1) using risk insights, consistent with current regulatory requirements, to assign SSCs to one of a limited set of graded categories, and (2) adjusting the scope and depth of current review plans--where possible--consistent with regulatory requirements and consistent with the applicable graded category. Because the staff's review constitutes an independent audit of the application, the staff may emphasize or de-emphasize particular aspects of its review guidance (i.e., Standard Review Plan), as appropriate and consistent with regulatory requirements, for the application being reviewed. The staff may propose justifications for not performing certain sections of the reviews called for by the applicable review plan. Examples of acceptable variations in the scope of a review can include reduced emphasis on SSC attributes such as reliability, availability, or functional performance when the SSC will be in the scope of a program

Progress of nuclear safety research, 1990

International Nuclear Information System (INIS)

1990-07-01

Since the Japan Atomic Energy Research Institute (JAERI) was founded as a nonprofit, general research and development organization for the peaceful use of nuclear energy, it has actively pursued the research and development of nuclear energy. Nuclear energy is the primary source of energy in Japan where energy resources are scarce. The safety research is recognized at JAERI as one of the important issues to be clarified, and the safety research on nuclear power generation, nuclear fuel cycle, waste management and environmental safety has been conducted systematically since 1973. As of the end of 1989, 38 reactors were in operation in Japan, and the nuclear electric power generated in 1988 reached 29 % of the total electric power generated. 50 years have passed since nuclear fission was discovered in 1939. The objective of the safety research at JAERI is to earn public support and trust for the use of nuclear energy. The overview of the safety research at JAERI, fuel behavior, reliability of reactor structures and components, reactor thermal-hydraulics during LOCA, safety assessment of nuclear power plants and nuclear fuel cycle facilities, radioactive waste management and environmental radioactivity are reported. (K.I.)

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-15

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

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

International Nuclear Information System (INIS)

2012-01-01

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

Regulatory considerations for computational requirements for nuclear criticality safety

International Nuclear Information System (INIS)

Bidinger, G.H.

1995-01-01

As part of its safety mission, the U.S. Nuclear Regulatory Commission (NRC) approves the use of computational methods as part of the demonstration of nuclear criticality safety. While each NRC office has different criteria for accepting computational methods for nuclear criticality safety results, the Office of Nuclear Materials Safety and Safeguards (NMSS) approves the use of specific computational methods and methodologies for nuclear criticality safety analyses by specific companies (licensees or consultants). By contrast, the Office of Nuclear Reactor Regulation approves codes for general use. Historically, computational methods progressed from empirical methods to one-dimensional diffusion and discrete ordinates transport calculations and then to three-dimensional Monte Carlo transport calculations. With the advent of faster computational ability, three-dimensional diffusion and discrete ordinates transport calculations are gaining favor. With the proper user controls, NMSS has accepted any and all of these methods for demonstrations of nuclear criticality safety

Results from the Nuclear Plant Aging Research Program: Their use in inspection activities

Energy Technology Data Exchange (ETDEWEB)

Gunther, W.; Taylor, J. (Brookhaven National Lab., Upton, NY (USA))

1990-09-01

The US NCR's Nuclear Plant Aging Research (NPAR) Program has determined the susceptibility to aging of components and systems, and the potential for aging to impact plant safety and availability. The NPAR Program also identified methods for detecting and mitigating aging in components. This report describes the NPAR results which can enhance NRC inspection activities. Recommendations are provided for communicating pertinent information to NRC inspectors. These recommendations are based on a detailed assessment of the NRC's Inspection Program, and feedback from resident and regional inspectors as described within. Examples of NPAR report summaries and aging inspection guides for components and systems are included. 13 refs., 3 figs., 4 tabs.

Research for enhancing reactor safety

International Nuclear Information System (INIS)

1989-05-01

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

Reactive inspection response of NRC Region III to potential technical deficiencies identified in recent Nuclear Air Cleaning Conference papers

International Nuclear Information System (INIS)

Gill, C.F.

1987-01-01

In order to effectively meet its responsibility to protect the public health and safety, the Nuclear Regulatory Commission (NRC) nuclear power plant licensing and inspection programs respond to potential technical deficiencies identified by conference and professional society meeting papers when deemed appropriate. The NRC staff's response mechanisms for such technical deficiencies include: generic letters, Bulletins, Information Notices, Standard Review Plan (NUREG-0800) revisions, docketed Final Safety Analysis Report (FSAR) questions, special studies, special (reactive) inspection, and inspection program revisions. This paper describes reactive inspection efforts by Region III in response to potential technical deficiencies identified in recent air cleaning conference papers, including: post-accident effluent sample line deposition losses; failure to implement good engineering practices in the design, construction, and testing of Nuclear Air Treatment Systems (NATS); filter bypass via filter housing drain lines; spinster carbon degradation; use of silicone sealants and other temporary patching material in NATS; filter housing fire protection deluge system problems; lack of charcoal batch traceability; Quality Assurance records problems involving equipment, vendor, filter, and personnel qualifications; inadequate ANSI/ASME N510 acceptance criteria and tests; and failure to adequately demonstrate control room habitability per 10 CFR 50, Appendix A, General Design Criterion-19. Region III inspections indicate that many of these deficiencies appear to be prevalent. Inspection findings and utility responses to the findings are discussed. NRC Region III and Headquarters programmatic reactions to the identified generic problem areas are also discussed

Safety Experts Complete IAEA Nuclear Regulatory Review of the United States

International Nuclear Information System (INIS)

2010-01-01

Full text: An international team of senior nuclear safety experts today completed a two-week International Atomic Energy Agency (IAEA) review of the governmental and regulatory framework for nuclear safety in the United States. The team identified good practices within the U.S. system and offered suggestions for ways the U.S. Nuclear Regulatory Commission (NRC) could improve. The IAEA has conveyed the team's main conclusions to the NRC, and a final report will be submitted to the NRC in about two months. At the request of the United States, the IAEA assembled a team of 19 international experts to conduct an Integrated Regulatory Review Service (IRRS) mission. This mission was a peer review based on the IAEA Safety Standards. It was not an inspection, nor an audit. The experts came from 14 different countries: Canada, China, the Czech Republic, Finland, France, Italy, Japan, Mexico, the Republic of Korea, Slovenia, Spain, Sweden, Switzerland, and the United Kingdom. Team leader Jukka Laaksonen of Finland said: ''We found a comprehensive, consistent, and mature regulatory system run by the NRC, which has a strong drive for continuous improvement.' The scope of the mission included the U.S. regulatory framework and the regulation of the nuclear plant operation. The mission was conducted from 18 to 29 October, mainly at NRC headquarters outside of Washington, D.C. To study U.S. regulatory activities, the mission conducted a series of interviews and discussions with NRC staff and other organizations to help assess the effectiveness of the regulatory system. In addition, the team observed regulatory activities at two operating nuclear power reactors and an emergency preparedness exercise. The IAEA's IRRS coordinator Gustavo Caruso said, ''This mission represents a milestone for the IRRS program because the U.S. regulatory system is the largest in the world and many nations look to it. The IRRS is a useful tool that allows host nations to gain guidance from experienced

Researches in nuclear safety

International Nuclear Information System (INIS)

Souchet, Y.

2009-01-01

This article comprises three parts: 1 - some general considerations aiming at explaining the main motivations of safety researches, and at briefly presenting the important role of some organisations in the international conciliation, and the most common approach used in safety researches (analytical experiments, calculation codes, global experiments); 2 - an overview of some of the main safety problems that are the object of worldwide research programs (natural disasters, industrial disasters, criticality, human and organisational factors, fuel behaviour in accidental situation, serious accidents: core meltdown, corium spreading, failure of the confinement building, radioactive releases). Considering the huge number of research topics, this part cannot be exhaustive and many topics are not approached; 3 - the presentation of two research programs addressing very different problems: the evaluation of accidental releases in the case of a serious accident (behaviour of iodine and B 4 C, air infiltration, fission products release) and the propagation of a fire in a facility (PRISME program). These two programs belong to an international framework involving several partners from countries involved in nuclear energy usage. (J.S.)

10 CFR 2.1403 - Authority and role of the NRC staff.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Authority and role of the NRC staff. 2.1403 Section 2.1403... ORDERS Expedited Proceedings with Oral Hearings § 2.1403 Authority and role of the NRC staff. (a) During the pendency of any hearing under this subpart, consistent with the NRC staff's findings in its own...

Effect of generic issues program on improving safety

International Nuclear Information System (INIS)

Fard, M. R.; Kauffman, J. V.

2010-01-01

The U.S. Nuclear Regulatory Commission (NRC) identifies (by its assessment of plant operation) certain issues involving public health and safety, the common defense and security, or the environment that could affect multiple entities under NRC jurisdiction. The Generic Issues Program (GIP) addresses the resolution of these Generic Issues (GIs). The resolution of these issues may involve new or revised rules, new or revised guidance, or revised interpretation of rules or guidance that affect nuclear power plant licensees, nuclear material certificate holders, or holders of other regulatory approvals. U.S. NRC provides information related to the past and ongoing GIP activities to the general public by the use of three main resources, namely NUREG-0933, 'Resolution of Generic Safety Issues, ' Generic Issues Management Control System (GIMCS), and GIP public web page. GIP information resources provide information such as historical information on resolved GIs, current status of the open GIs, policy documents, program procedures, GIP annual and quarterly reports and the process to contact GIP and propose a GI This paper provides an overview of the GIP and several examples of safety improvements resulting from the resolution of GIs. In addition, the paper provides a brief discussion of a few recent GIs to illustrate how the program functions to improve safety. (authors)

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

International Nuclear Information System (INIS)

2013-01-01

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

NRC high-level radioactive waste research at CNWRA, July--December 1992

International Nuclear Information System (INIS)

Sagar, B.; Ababou, R.; Ahola, M.

1993-07-01

Progress from July 1 to December 31, 1992 on the nine NRC-sponsored research projects conducted at the Center for Nuclear Waste Regulatory Analyses is described. Ion-exchange experiments between clinoptilolite and aqueous solutions of Na + and Sr 2+ and three applications of reaction-path modeling are described in the Unsaturated Mass Transport (Geochemistry) project. Numerical simulation of a laboratory-scale non-isothermal two-phase flow is discussed in the Thermohydrology chapter. Methods for estimating rock joint roughness coefficient are the focus of the Seismic Rock Mechanics project for which the Tilt Test, Tse and Cruden's equations, and fractal-based equations were tested and found to be unsatisfactory. In the Integrated Waste Package Experiments chapter, investigations of pit initiation and repassivation potential for alloys 825 and C-22 and stainless steel 304L and 316L are described. Testing of the BIGFLOW computer code and visualization of fracture topology is the theme of the Stochastic Hydrology project. Preliminary analysis of field data from the Akrotiri site in Greece is developed in the Geochemical Analogs project. Mechanistic modeling of sorption using the MINTEQA2 code is investigated as part of the Sorption project. Adaptive gridding and ''modified equations'' methods for solving the flow and transport equations are described in the Performance Assessment chapter. Finally, the Volcanism chapter focuses on using nonhomogeneous Poisson processes for estimating probability of volcanic events at the potential repository site

Nuclear safety research

International Nuclear Information System (INIS)

1999-01-01

The NNSA checked and coordinated in 1999 the research project of the Surveillance Technology on Nuclear Installations under the National 9th-Five-Year Program to promote the organizations that undertake the research work on schedule and lay a foundation of obtaining achievements and effectiveness for the 9th-five-year plan on nuclear safety research

NRC study of control room habitability

International Nuclear Information System (INIS)

Hayes, J.J. Jr.; Muller, D.R.; Gammill, W.P.

1985-01-01

Since 1980, the Advisory Committee on Reactor Safeguards (ACRS) has held several meetings with the NRC staff to discuss the subject of control room habitability. Several meetings between the ACRS and the staff have resulted in ACRS letters that express specific concerns, and the staff has provided responses in reports and meetings. In June of 1983, the NRC Executive Director for Operations directed the Offices of Nuclear Reactor Regulation and Inspection and Enforcement to develop a plan to handle the issues raised by the ACRS and to report to him specific proposed courses of action to respond to the ACRS's concerns. The NRC control room habitability working group has reviewed the subject in such areas as NRR review process, transformation of control room habitability designs to as-built systems, and determination of testing protocol. The group has determined that many of the ACRS concerns and recommendations are well founded, and has recommended actions to be taken to address these as well as other concerns which were raised independent of the ACRS. The review has revealed significant areas where the approach presently utilized in reviews should be altered

NRC TLD Direct Radiation Monitoring Network

International Nuclear Information System (INIS)

Struckmeyer, R.; McNamara, N.

1992-06-01

The US Nuclear Regulatory Commission (NRC) Direct Radiation Monitoring Network is operated by the NRC in cooperation with participating states to provide continuous measurement of the ambient radiation levels around licensed NRC facilities, primarily power reactors. Ambient radiation levels result from naturally occurring radionuclides present in the soil, cosmic radiation constantly bombarding the earth from outer space, and the contribution, if any, from the monitored facilities and other man-made sources. The Network is intended to measure radiation levels during routine facility operations and to establish background radiation levels used to assess the radiological impact of an unusual condition, such as an accident. This report presents the radiation levels measured around all facilities in the Network for the first quarter of 1992. All radiation measurements are made using small, passive detectors called thermoluminescent dosimeters (TLDs), which provide a quantitative measurement of the radiation levels in the area in which they are placed. Each site is monitored by arranging approximately 40 to 50 TLD stations in two concentric rings extending to about five miles from the facility. All TLD stations are outside the site boundary of the facility

Program nuclear safety research: report 2000

International Nuclear Information System (INIS)

Muehl, B.

2001-09-01

The reactor safety R and D work of forschungszentrum karlsruhe (FZK) had been part of the nuclear safety research project (PSF) since 1990. In 2000, a new organisational structure was introduced and the Nuclear Safety Research Project was transferred into the nuclear safety research programme (NUKLEAR). In addition to the three traditional main topics - Light Water Reactor safety, Innovative systems, Studies related to the transmutation of actinides -, the new Programme NUKLEAR also covers Safety research related to final waste storage and Immobilisation of HAW. These new topics, however, will only be dealt with in the next annual report. Some tasks related to the traditional topics have been concluded and do no longer appear in the annual report; other tasks are new and are described for the first time. Numerous institutes of the research centre contribute to the work programme, as well as several external partners. The tasks are coordinated in agreement with internal and external working groups. The contributions to this report, which are either written in German or in English, correspond to the status of early/mid 2001. (orig.)

Plutonium air transportable package Model PAT-1. Safety analysis report

International Nuclear Information System (INIS)

1978-02-01

The document is a Safety Analysis Report for the Plutonium Air Transportable Package, Model PAT-1, which was developed by Sandia Laboratories under contract to the Nuclear Regulatory Commission (NRC). The document describes the engineering tests and evaluations that the NRC staff used as a basis to determine that the package design meets the requirements specified in the NRC ''Qualification Criteria to Certify a Package for Air Transport of Plutonium'' (NUREG-0360). By virtue of its ability to meet the NRC Qualification Criteria, the package design is capable of safely withstanding severe aircraft accidents. The document also includes engineering drawings and specifications for the package. 92 figs, 29 tables

Compilation of contract research for the Materials Engineering Branch, Division of Engineering: Annual report for FY 1987

International Nuclear Information System (INIS)

1988-06-01

This compilation of annual reports by contractors to the Materials Engineering Branch of the NRC Office of Research concentrates on achievements in safety research for the primary system of commercial light water power reactors, particularly with regard to reactor vessels, primary system piping, steam generators, nondestructive examination of primary components, and in safety research for decommissioning and decontamination, on-site storage, and engineered safety features. This report, covering research conducted during Fiscal Year 1987 is the sixth volume of the series of NUREG-0975, ''Compilation of Contractor Research for the Materials Engineering Branch, Division of Engineering.''

The use of research results for effective aging management

International Nuclear Information System (INIS)

Gunther, W.E.; Taylor, J.H.

1989-01-01

The study of the degradation of structures, components, and systems due to aging is an important ongoing area of research in the nuclear industry. Efforts by the U.S. Nuclear Regulatory Commission (NRC) and the utility industry, through organizations such as the Electric Power Research Institute (EPRI), have produced substantial research results that can be used by inspectors and operators to effectively understand and manage the aging of nuclear power plants. One of the primary objectives of the NRC's Nuclear Plant Aging Research (NPAR) Program is to determine how aging affects the safety of nuclear power plants. This program uses operating experience, testing, and engineering analysis to identify failures caused by age-related degradation. Useful information on aging has also resulted from research being performed by the industry to support plant-life extension (PLEX). The EPRI program, for instance, is directed toward the resolution of issues related to materials and components. Degradation of equipment and systems due to aging can occur which, if unmitigated, could result in reduction of the nuclear power plant safety margin as the plant ages. This paper describes how aging research results may be used by plant operating management to effectively address the aging issue and by inspectors responsible for monitoring plant activities and programs

77 FR 32146 - Safety Evaluation Report, International Isotopes Fluorine Products, Inc., Fluorine Extraction...

Science.gov (United States)

2012-05-31

...-3119; email: [email protected] . SUPPLEMENTARY INFORMATION: I. Introduction By letter dated..., organization and administration, integrated safety analysis (ISA) and ISA summary, radiation protection..., you can access the NRC's ADAMS, which provides text and image files of NRC's public documents. The...

Risk as a target of safety research

International Nuclear Information System (INIS)

Krueger, W.

1986-01-01

Job creation is not the idea behind the demand for risk studies to be intensified in safety research. Risks are not only a target safety research should investigate, they are a subject that actually can be most adequately investigated by safety research. Assuming a neutral position between irrational fears and interest-minded problem minimization, that is the central approach and the ethics of a safety scientist. The Babylonian confusion of terminology experienced after the Chernobyl accident is a good example proving the necessity of fostering the neutral professionalism in safety research. (orig./DG) [de

Energy Technology Division research summary 2001

International Nuclear Information System (INIS)

2001-01-01

The Energy Technology Division provides materials and engineering technology support to a wide range of programs important to the U.S. Department of Energy. As shown on the preceding page, the Division is organized into eight sections, four with concentrations in the materials area and four in engineering technology. Materials expertise includes fabrication, mechanical properties, corrosion, friction and lubrication, and irradiation effects. Our major engineering strengths are in heat and mass flow, sensors and instrumentation, nondestructive testing, transportation, and electromechanics and superconductivity applications. The Division Safety Coordinator, Environmental Compliance Officer, Quality Assurance Representative, Financial Administrator, and Communication Coordinator report directly to the Division Director. The Division Director is personally responsible for cultural diversity and is a member of the Laboratory-wide Cultural Diversity Advisory Committee. This Overview highlights some major ET research areas. Research related to the operational safety of commercial light water nuclear reactors (LWRs) for the U.S. Nuclear Regulatory Commission (NRC) remains a significant area of interest for the Division. We currently have programs on environmentally assisted cracking, steam generator integrity, and the integrity of high-burnup fuel during loss-of-coolant accidents. The bulk of the NRC research work is carried out by three ET sections: Corrosion and Mechanics of Materials; Irradiation Performance; and Sensors, Instrumentation, and Nondestructive Evaluation

Trend analysis of incidents involving setpoint drift in safety or safety/relief valves at U.S. LWRs

International Nuclear Information System (INIS)

Watanabe, Norio

2008-01-01

Since the beginning of the 1980's, in the United States, there have been many licensee event reports (LERs) involving setpoint drift in safety or safety/relief valves. The United States Nuclear Regulatory Commission (NRC) has issued a lot of generic communications on this issue and the industry has made efforts to resolve the issue. However, the NRC staff recently highlighted that over 70 LERs involved instances where safety or safety/relief valves failed to meet the allowed setpoint tolerance from 2001 through August 2006. In the present study, we analyzed the U.S. experience with setpoint drift in safety/relief valves (SRVs) at BWRs, pressurizer safety valves (PSVs), and main steam safety valves (MSSVs) at PWRs by reviewing approximately 90 LERs from 2000 to 2006 and examined the trend focusing on causes and setpoint deviation ranges. This study indicates that for SRVs and MSSVs, disc-seat bonding is a dominant cause of the setpoint drifting high and has a tendency to result in a relatively large deviation of the setpoint. This means that disc-seat bonding might be a safety concern from the view point of overpressure protection. For PSVs, the deviation of setpoints is generally small, although its causes are not specified in many instances. (author)

Degraded voltage protection at nuclear plant safety buses

International Nuclear Information System (INIS)

Haddad, S.Z.; Berger, W.E.

1989-01-01

This paper reviews the events that led to the NRC's degraded voltage protection requirement and outlines the related NRC positions. It describes problems that have occurred with the protection scheme and identifies the measures required to minimize the occurrence of these problems. The paper recommends new guidelines for degraded voltage protection and proposes a protection scheme logic that aims at enhancing the reliability of the safety systems

Shifting to non-explosive fuels for research reactors

International Nuclear Information System (INIS)

Gilinsky, Victor

1984-01-01

The RERTR program is not just an American program, it is an international program and it can succeed only with wide support and participation. Excellent work underway at research laboratories in several countries is making a vital contribution. The cooperative spirit shown by all participants is particularly gratifying. Some practical difficulties may be encountered with the safety regulatory agencies in different countries when modifying the reactors. The US NRC intends to demonstrate that conversion to low enriched fuel is not a difficult process by taking steps to enable domestic reactors operating in the United States to convert to low enriched fuels. A proposed regulation is being prepared limiting the use of highly enriched uranium in domestic reactors. In this connection, The US NRC will be prepared tp cooperate with reactor operators from other countries in the safety area as it relates to the conversion process

Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants

International Nuclear Information System (INIS)

O'Hara, J.M.; Higgins, J.; Fleger, Stephen

2011-01-01

The U.S. Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) programs of applicants for nuclear power plant construction permits, operating licenses, standard design certifications, and combined operating licenses. The purpose of these safety reviews is to help ensure that personnel performance and reliability are appropriately supported. Detailed design review procedures and guidance for the evaluations is provided in three key documents: the Standard Review Plan (NUREG-0800), the HFE Program Review Model (NUREG-0711), and the Human-System Interface Design Review Guidelines (NUREG-0700). These documents were last revised in 2007, 2004 and 2002, respectively. The NRC is committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool. To this end, the NRC is updating its guidance to stay current with recent research on human performance, advances in HFE methods and tools, and new technology being employed in plant and control room design. This paper describes the role of HFE guidelines in the safety review process and the content of the key HFE guidelines used. Then we will present the methodology used to develop HFE guidance and update these documents, and describe the current status of the update program.