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.

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.

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.

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, …

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 …

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.

The objective of this program was to provide a mechanical-property database that could be used to predict the performance of Zircaloy cladding under various off-normal, transient, and accident conditions in a power reactor.

"This report presents the results of the evaluation of trench subsidence and possible future stabilization techniques at the Sheffield Low-Level Radioactive Waste Disposal Facility, Sheffield, Illinois."

This report examines the adequacy of existing fire-hazards analysis methodology in the context of nuclear power plant safety

SABRES II is an individual resolution computer simulation of combat between two small groups using small arms weapons.

This report discusses the results of a comparative review of the selection, training, qualification, and requalification of Maitenncace personnel, operators, shift supervisors, and senior onsite managers involved in the operation and/or maintenance of nuclear power plants.

The purposes of this report are to document the data sources and the procedures used in the reduction of the data and to present a Bouguer gravity anomaly map derived from the data set.

WRAITH is a FORTRAN computer code which calculates the doses received by a standard man exposed to an accidental release of radioactive material

The following data provides the code for the SABRES II model.

The following data is a code listing and sample case output calculating the steady state thermal-mechanical behavior of oxide fuel rods.

The compilation, assessment, and expansion of the Strong Earthquake Ground Motion Data Base.

A catalog had been prepared which contains information for world-wide, ground-motion accelerograms, the accelerogaph sites where these records were obtained, and the seismological parameters of the causative earthquakes.

Computer printout listings of three data tables (Digitized Accelerogram, Recording Station, and Earthquake) organized alphabetically by country.

Report presenting the results of nondestructive and destructive examinations of four fuel rods from Test RIA 1-2.

Report detailing the computer code used to calculate the cladding temperatures (WIZARD code) from the observed oxidation of the cladding for Test RIA 1-2.

Report listing the table of contents for appendices and figures included with the report: "Reactivity Initiated Accident Test Series Test RIA 1-2 Fuel Behavior Report'

This report presents and discusses the results from the Reactivity Initiated Accident Test RIA1-2 conducted in the Power Burst Facility at the Idaho National Engineering Laboratory.

A report presenting the results from Power-Cooling-Mismatch Test CM-3