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 final report regarding a Multiloop Integral System Test, or MIST.

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

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."

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.

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)

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

Graphs presenting a summary of data for a fracture-mechanics parametric-type study conducted for a number of postulated transients, including critical fluences, pressures, and crack depth.

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)