Report presenting data plots for selected parameters measured during Test PCM-4 of the Power-Cooling-Mismatch Test Series.

"A comparison of the calculated and measured fuel rod behavior (cladding elongation, fuel centerline temperature, and fuel rod internal pressure) determined during the steady state power calibration, onset of departure from nucleate boiling (DNB), and post-DNB testing is presented in this appendix" (p. 184).

"A summary of the methods used to establish fuel rod cladding temperatures associated with the axial region of film boiling for Test PCM-4 and the results of the calculations are presented in this appendix." (p. 219)

This report presents and discusses the results from Test PCM-4 of the Power-Cooling-Mismatch Test Series.

"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."

Bouguer gravity anomaly map derived from the data set and accompanying the report to document the data sources and the procedures used in the reduction of the data.

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.

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.

Report presenting preparation of the Power-Cooling-Mismatch Test Series PCM-5 fuel rods and fuel rod sections for metallographic examination and the metallographic examination results from Rods 205-1, -2, -3, -4, -5, -6, and -8.

Report presenting cladding axial and circumferential temperature estimates, fuel rod radial temperature estimates, and the techniques used to estimate the cladding and fuel temperatures for Power-Cooling-Mismatch Test Series PCM-5.

Report presenting microprobe examination results of fuel Rods 205-1, 205-2, and 205-8 for test Power-Cooling-Mismatch Test Series PCM-5.

Report presenting fuel rod particle sizes and size distributions as determined for the two failed Test PCM-5 fuel rods.

Report presenting scanning electron microscopy (SEM) examinations used to analyze one cladding fracture surface of Rod 205-4 for 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.

Dynamic test data from an experiment conducted to determine the puncture resistance of lead- and uranium-backed type AISI 316 stainless steel (SS) plates. The data is presented in line graphs. The parameters that were varied in this investigation were punch diameter, test plate thickness, backing material, and test temperature.

Report on the development of improved analytical methods for predicting puncture of radioactive material shipping containers.

Graphical presentations of data from the 1/5-scale Mark I BWR Pressure Suppression Experiment (PSE) air transient tests, including HVLF Characterization Data, Impulse Characterization, Computed Load Cell Response Histories and Frequency Analyses, and Enthalpy Flux Evaluation (see Table E-1. Summary of microfiche descriptions)

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.

MAthematical-modeling studies were carried out to determine effects of isothermal and anisothermal anneals on recovery of mechanical properties of irradiated Zircaloy.

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.

Computer-generated graphical and tabular presentations of transient test and posttest deformation data for Multirod Burst Test (MRBT) Bundle B-6 (see pages 131-141 of report). "The primary objective of this 8 x 8 multirod burst test was to investigate cladding de-formation in the alpha-plus-beta-Zircaloy temperature range under simulated light-water-reactor (LWR) loss-of-coolant accident (LOCA) conditions. B-6 test conditions simulated the adiabatic heatup (reheat) phase of an LOCA and produced very uniform temperature distributions." - from Abstract