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 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 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.

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

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.

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 first three power-cooling-mismatch tests - Tests 8-1 RS CHF Scoping, and 8-1 RF - performed in the PCM Test Series.

This report presents the analysis of the fuel rod materials behavior based on the postirradiation examination of the Test PCM-1 fuel rod from the Power-Cooling-Mismatch Test Series.

This report presents the fuel rod behavior results of the Power-Cooling-Mismatch (PCM) Test Series, Test PCM-1.

From Abstract. "This report describes a linear elastic finite element computer program (NOZ-FLAW) which is used for the direct evaluation of mixed-mode stress intensity factors (K1, KII K111) along user-defined crack flaws at pressure-vessel nozzle corners." From Forward: "The work reported here was performed at the Oak Ridge National Laboratory and at the Georgia Institute of Technology under UCC-ND Subcontract No. 7565."

A report discussing work performed at the Oak Ridge National Laboratory in support of the ORNL Design Criteria for Piping and Nozzle Program being conducted for the U.S. Nuclear Regulatory Research.

This report presents recorded test data for Tests LLR-S0, LLR-3, LLR-4A, and LLR-5 of the thermal Fuels Behavior Program PBF/LOFT Lead Rod Test Series.

This report presents data acquired from Instrumental Fuel Assembly 429.

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

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

FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

Discussing sensitivity studies of the response surface developed for a seven-variable, sparsely-noded set of RELAP calculations.

This report was compiled at the Hanford Engineering Development Laboratory operated by Westinghouse Hanford Company, a subsidiary of Westinghouse Electric Corpoation, for the United States Department of Energy and the Nuclear Regulatory Commission, under DOE contract number DE-AC14-76FF02170 and NRC service request number TV-0176.