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.
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.
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.
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 ...
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.
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.
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.