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.

An aeromagnetic map of the East-Central Midcontinent of the United States

A report regarding experiment data for a multirod burst test, to investigate cladding deformation in the alpha-plus-beta-Ziracloy temperature range under light-water-reactor (LWR) loss-of-coolant accident (LOCA) conditions."

A parametric analysis regarding PWR pressure vessel integrity during overcooking accidents.

The results of an in-pile nuclear blowdown test series, designated TC-1, are presented in this report.

Graphs plotting the thermal-hydraulic response of the TC-1 tests characterized by the system depressurization, volumetric flow, temperature (in Kelvin) over time within each fuel rod flow shroud used in the study.

Graphs presenting peak power history, ratio of local rod power to rod average power (P/A), total heat transfer coefficient history, and coolant pressure history conditions in the test assembly during the transient phase of the TC-1 tests.

Tables presenting design variables for TC-1 fuel rods including composite powder analysis, pellet and fuel stack dimensions, and fuel rod and flow shroud dimensions.

Report presenting the design of the Test TC-1 test train and Power Burst Facility (PBF) blowdown system and the test conduct in tables, text, and illustrations.

Analyses, interpretations, and discussions of results from the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-1, conducted in the Power Burst Facility reactor are presented.

Report discussing and comparing four techniques used for cladding temperatures achieved during Test RIA 1-1: metallorgraphic examination of cladding microstructures, computations based on isothermal oxidation kinetics, BUID5 computer code calculations, and weight gain calculations for metal-water reaction (p. 305)

"A summary of the experimental irradiation and energy deposition measurements of fuel enthalpy, useful in assessing the test fuel rod thermal history, is presented in this appendix." (p. 116)

Report presenting the results of the metallographic examination for Rods 801-1, 801-2, 801-3, and 801-5 used to supplement the visual inspection and posttest description of the Test RIA 1-1 fuel rods. (p. 177)

Tables, text, and diagrams describing the design, instrumentation, and assembly of fuel rods uses in Test RIA 1-1.

"The pretest characterization of the four Test RIA 1-1 fuel rods and flow shrouds is presented in this appendix." (p. 147)

Appendix to report discussing the SINGLE computer code used to determine radial fuel temperature profiles.

Report presenting the pretest fuel rod characterization data for four rods used in Test RIA 1-2.

Report describing four different methods of measuring the total radially averaged fission energy deposited in PBF (Power Burst Facility) fuel rods, as well as a summary evaluation of each method.

Report providing a detailed description of the Power Burst Facility (PBF) design and capabilities, including an illustration of the axial cross section of the PBF in-pile tube.

Report summarizing the design of Power Burst Facility (PBF) test rods, test train hardware, the individual rod assembly instrumentation, and test conduct.

"This report describes the methods, analyses, results, and conclusions of two different aging studies."