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 scanning electron microscopy (SEM) examinations used to analyze one cladding fracture surface of Rod 205-4 for 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 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 detailing postirradiation examinations performed on the three pressurized water-reactor-type fuel rods following Tests PCM 8-1 RS, CHF Scoping, and PCM 8-1 RF.

Report discussing the results of experimental methods for determining fuel rod power needed for three single-rod Power-Cooling Mismatch (PCM) tests.

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.

Report presenting time history plots of selected parameters measured during the three tests: PCM 8-1 RS, CHF Scoping, and PCM 8-1 RF.

Report presenting pretest measurement characteristics of the fuel rods used in three, single-rod, power-cooling-mismatch tests: PCM 8-1 RS, CHF Scoping, and PCM 8-1 RF.

Appendix detailing the hardware and instrumentation of test assemblies for the three single-rod Power-Cooling Mismatch (PCM) tests.

Charts analyzing Multirod Burst Test (MRBT) Bundle B-5 test data produced at Oak Ridge National Laboratory.

Sample input and output data from a linear elastic finite element computer program (NOZ-FLAW).

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."

Tables presenting the results of COPS computer model analyses, which has been used to estimate local law enforcement agency (LLEA) officer availability in the neighborhood of 60 selected nuclear reactor sites. The tables show the unrestricted response and restricted response for both state and local police within a given radius for each neighborhood in the study (see page 10 of report).

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

"This report describes a systematic study to determine the relationship between seismicity and geologic or geophysical features for the central and eastern United States in order to delineate seismic hazard."

"This report presents results of the investigation to characterize the existing surface conditions at the Sheffield Low-Level Waste Disposal Facility Sheffield Illinois."

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.

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.

The following data contains the line printer output for the terminal session for sample run #2 as described in its parent document, WRAITH: a Computer Code for Calculating Internal and External Doses Resulting From an Atomospheric Release of Radioactive Material.

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.

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