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Fault Tree Handbook

Description: Introduction: Since 1975, a short course entitled "System Safety and Reliability Analysis" has been presented to over 200 NRC personnel and contractors. The course has been taught jointly by David F. Haasl, Institute of System Sciences, Professor Norman H. Roberts, University of Washington, and members of the Probabilistic Analysis Staff, NRC, as part of a risk assessment training program sponsored by the Probabilistic Analysis Staff. This handbook has been developed not only to serve as text for the System Safety and Reliability Course, but also to make available to others a set of otherwise undocumented material on fault tree construction and evaluation. The publication of this handbook is in accordance with the recommendations of the Risk Assessment Review Group Report (NUREG/CR-0400) in which it was stated that the fault/event tree methodology both can and should be used more widely by the NRC. It is hoped that this document will help to codify and systematize the fault tree approach to systems analysis.
Date: January 1981
Creator: Vesely, W. E.

Federal/State Regulatory Permitting Actions in Selected Nuclear Power Station Licensing Cases

Description: Abstract: This report documents the Federal/State regulatory permitting actions in 12 case histories of nuclear power station licensing in nine different states. General observations regarding Federal/State siting roles in the siting process include: new regulations, with the exceptions of those imposed by NEPA, were not found to be the source of significant delay; interventions were the sources of significant delay in only two cases; in only two cases was a local agency a source of delay; no one factor was found to be a source of delay, rather several factors often combined to cause delay; it is still too early to assess the influence of State power plant siting laws on the licensing process; clarification of the word "delay" is needed; water related issues predominate in State permitting requirements associated with delay; generalizations on the sources and nature of delay in the licensing process are difficult to make because of site specific characteristics; and frequently problems outside the Federal/State realm have had, or can have, a delaying effect on the process. Eleven of the case histories are illustrated with a logic network that gives the actions of the utilities in addition to the Federal/State permits.
Date: June 1977
Creator: U.S. Nuclear Regulatory Commission. Office of State Programs.

Final Environmental Statement by the U.S. Nuclear Regulatory Commission for Greene County Nuclear Power Plant

Description: Abstract: A Final Environmental Statement for the Power Authority of the State of New York for the construction of the Greene County Nuclear Power Plant (Docket No. 50-549) located in Greene County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission (NRC). This statement provides (1) a summary of environmental impact and adverse effects of the proposed facility, and (2) a consideration of principal alternatives. Also included are comments of governmental agencies and other organizations on the Draft Environmental Statement for the project and staff responses to these comments. The NRC staff has concluded, based on a weighing of environmental, economic, technical, and other benefit against environmental costs and available alternatives, that a construction permit should be denied because the alternative sites available to the applicant are environmentally preferable. If the permit is granted, the applicant will be required to take the necessary mitigating actions to decrease the aesthetic impact by using alternative closed cycle cooling systems and to undertake monitoring programs to identify, evaluate and mitigate construction related community and public services impacts in the immediate three-county impact area.
Date: January 1979
Creator: U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.

Final Environmental Statement by the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation for Montague Nuclear Power Station, Units 1 and 2

Description: The proposed project: Pursuant to the Atomic Energy Act, as amended, the U.S. Nuclear Regulatory Commission's regulations in Title 10, Code of Federal Regulations, an application with an accompanying Environmental Report, was filed by Northeast Utilities (hereinafter referred to as the applicant) for construction permits for two generating units designated as the Montague Nuclear Power Station, Units 1 and 2 (Docket Nos. 50-496 and 50-497), each of which is powered by a boiling water reactor (BWR) and is designed for initial operation at approximately 3579 megawatts thermal (MWt) with a net electrical output of 1150 megawatts electric (MWe). A safety design rating of 3759 (MWt) has been used in assessing the impact in this report. Condenser cooling will be accomplished through the use of natural-draft cooling towers. Makeup water for the cooling towers will be obtained from the Connecticut River, and the tower discharge (blowdown) will be returned to the Connecticut River. The proposed facilities will be located on the 1900-acre Montague Plain in the Town of Montague, Franklin County, in northwestern Massachusetts about 1.8 miles east of the Connecticut River and about 3.5 miles east-southeast of the Town of Greenfield, Massachusetts, the largest community within 10 miles with a population of about 15,000. Integration of the power from the Montague Nuclear Power Station will be accomplished by individual routes for each unit, requiring the construction of approximately 118 miles of 345-kV circuit transmission lines into existing electrical systems. A 345-kV switchyard will be located on the Montague site in proximity to the generating units and will constitute the terminus of the 345-kV circuits over which the output of the station will be delivered to the load centers. The route for Unit 1 will terminate at the Ludlow, Massachusetts, substation, and the route for Unit 2 will terminate at the ...
Date: February 1977
Creator: U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.

Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants

Description: From abstract: As a result of review by the Nuclear Regulatory Commission (NRC) staff and extended collegial consultations and investigations within the NRC, the Commission has designated four new Unresolved Safety Issues (USIs). This report describes the process used to evaluate the large number of concerns and recommendations which resulted from the major investigations of the Three Mile Island-2 accident, as well as other events and investigations of the past year, and it identifies the four new USIs.
Date: March 1981

Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, Appendices 3 and 4

Description: From section 1: In the quantitative system probability estimates performed in this study, component behavior data in the form of failure rates and repair times are required as inputs to the system models. Since the goal of this study is risk assessment, as opposed to reliability analysis, larger errors (e.g. order of magnitude type accuracy) can be tolerated in the quantified results. This has important implications on the treatment of available data. In standard reliability analysis, point values (i.e., "best-estimates") are generally used for both data and results in quantifying the system model. In risk assessment, since results accurate to about an order of magnitude are sufficient, data and results using random variable and probabilistic approaches, can be usefully employed. The base of applicable failure rate data is thus significantly broadened since data with large error spreads and uncertainties can now be utilized. The data and associated material that were assembled for use in this study and that are presented here are to be used in the random variable framework (which will be described). The data and the accompanying framework are deemed sufficient for the study's needs. Care must be taken, however, since this data may not be sufficiently detailed, or accurate enough for use in general quantitative reliability models.
Date: October 1975
Creator: U.S. Nuclear Regulatory Commission

Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, Appendix 1

Description: From introduction: In conventional safety analyses, a suitable design basis, including redundancy, is specified to assure a minimum level of operability of ESFs, and the likelihood or consequences of total failure of ESFs are not considered further. In this study all failures are considered possible, but appropriate probabilities are assigned to them. Thus, many potential accident sequences are described in the following discussions as if they will surely occur, with no reservations expressed as to their likelihood or significance. However, most of these sequences have such low probability that they do not contribute to the overall risk from reactor accidents. In fact, in order to make an overall risk assessment, a major task of this study was to identify the sequences that are the dominant contributors to risk. In this study the initial failures or initiating events that could lead to significant consequences were examined to varying degrees. Those that seemed to contribute significantly to potential risks were analyzed in considerable detail; those that did not, received less detailed consideration. This is discussed more fully in section 3 of this appendix.
Date: October 1975
Creator: U.S. Nuclear Regulatory Commission

Report of the Advisory Committee on Construction During Adjudication

Description: Abstract: In the Nuclear Regulatory Commission's Seabrook Opinion of January 6, 1978, the Commission directed that an internal study be made of: (1) the effect which would be achieved by relaxation of NRC's stay standards so that site-related issues in potentially troublesome nuclear power plant licensing proceedings could be taken up before utilities invest large sums of money and sites are irrevocably altered; and (2) ways in which the NRC's appellate administrative procedures could assure earlier resolution of all the issues arising out of nuclear power plant licensing and cut relitigation and piecemeal review to a minimum. In December of 1978, the Commission chose Gary Milhollin, a faculty member at the University of Wisconsin Law School to chair a study group composed of nine other members. These members were chosen by the heads of various offices within the Commission. This report describes the Committee's work, summarizes the data gathered by the Committee, and recommends action by the Commission.
Date: January 1980
Creator: Milhollin, Gary

Safety Evaluation Report Related to the Operation of San Onofre Nuclear Generating Station, Units 2 and 3

Description: From introduction: This report is a safety evaluation report on the application for operating licenses for the San Onofre Nuclear Generating Station, Units 2 and 3 (San Onofre 2 and 3 or the facility). This report was prepared by the United States Nuclear Regulatory Commission staff (the NRC staff or the staff), and summarizes the results of our radiological safety review of the facility.
Date: February 1981
Creator: U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.

Safety Evaluation Report: Related to the Preliminary Design of the Standard Reference System RESAR-414

Description: From introduction: This Safety Evaluation Report summarizes the results of the technical evaluation of the proposed RESAR-414 design performed by the NRC staff, and delineates the scope of the technical matters considered in evaluating the radiological safety aspects of the RESAR-414 design. Environmental aspects were not considered in our review of RESAR-414, but will be addressed in each utility application for a construction permit which references RESAR-414.
Date: 1978
Creator: U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.