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 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.

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.

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."

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.

Appendix to report "Test PCM-5 Fuel Rod Materials Behavior" presenting pretest physical characteristics of the Power-Cooling-Mismatch Test Series PCM-5 fuel rod components.

Report presenting details of Test-PCM-5 fuel rod axial power profiles as determined by two methods: gamma scanning of the fuel bundle flux wires and radiochemical burnup analysis of fuel samples from Rods 205-1, -2, -3, -5, -7, -8, and -9. (p. B-2)

Appendix presenting postest bundle and individual fuel rod photographs for report "Test PCM-5 Fuel Rod Materials Behavior."

Report presenting rod-to-rod spacing measurements and individual rod diameter measurements for fuel rods used in the Power-Cooling-Mismatch Test Series PCM-5.

Test PCM-5 was the first nine-rod bundle test in the Power-Cooling-Mismatch (PCM) Test Series. The test objectives were to evaluate the behavior of the center rod operated in an environment typical of a commercial reactor rod subjected to PCM transient and to investigate the potential for rod-to-rod interactions.