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FRAP-T5 uncertainty study of five selected accidents and transients. [PWR; BWR]

Description: The FRAP-T5 transient fuel rod behavior code, with its automated uncertainty analysis subcode, was used to estimate the uncertainty of best estimate calculations simulating five selected reactor accident and transient events. Code input was supplied by the four US reactor vendors. Response surface methodology was used to estimate calculation uncertainty. Fifteen fuel rod design, model, and operation parameters were perturbed in this study. The uncertainty of both rod thermal and mechanical perf… more
Date: January 1, 1981
Creator: Laats, E.T.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Influence of FRAPCON-1 evaluation models on fuel behavior calculations for commercial power reactors. [PWR; BWR]

Description: A preliminary set of nine evaluation models (EMs) was added to the FRAPCON-1 computer code, which is used to calculate fuel rod behavior in a nuclear reactor during steady-state operation. The intent was to provide an audit code to be used in the United States Nuclear Regulatory Commission (NRC) licensing activities when calculations of conservative fuel rod temperatures are required. The EMs place conservatisms on the calculation of rod temperature by modifying the calculation of rod power his… more
Date: January 1, 1981
Creator: Chambers, R. & Laats, E. T.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

FRAP-T5 predictions during reactor shutdown events. [PWR; BWR]

Description: The Transient Fuel Rod Analysis Program, FRAP-T5, was recently assessed by EG and G Idaho, Inc. As part of this assessment, the measured and FRAP-T5 predicted fuel centerline temperature response during reactor shutdown events were compared. For these events either forced convection or nucleate boiling boundary conditions existed, resulting in a negligible effect on fuel behavior from cladding temperature and deformation uncertainties. This allowed the assessment of internal heat transfer to be… more
Date: January 1, 1980
Creator: Peeler, G.B. & Laats, E.T.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

FRAP-S2: a computer code for the steady state analysis of oxide fuel rods

Description: This document describes the Fuel Rod Analysis Program--Steady State code (FRAP-S) and the verification of FRAP-S as a fuel rod analysis tool. FRAP-S2 (Fuel Rod Analysis Program--Steady State, Version 2) is a FORTRAN IV computer code which can be used to solve for the long term burnup response of a light water reactor fuel rod. The coupled effects of fuel and cladding deformation, temperature distribution, internal gas accumulation, and pressure and material property on the behavior of the fuel … more
Date: July 1, 1977
Creator: Dearien, J. A.; Bohn, M. P.; Berna, G. A.; Coleman, D. R. & Laats, E. T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

USNRC-OECD Halden Project fuel behavior test program: experiment data report for test assemblies IFA-226 and IFA 239

Description: The experimental data which were obtained from the IFA-226 and IFA-239 test assemblies during operation in the Halden Boiling Water Reactor are reported. Included are cladding elongation, fuel centerline temperature, internal gas pressure, and power history data from IFA-226 which were obtained from November 1971 through April 1974, and cladding elongation, diametral profile, and power history data from IFA-239 covering the period from March 1973 through April 1974. The data, presented in the f… more
Date: December 1, 1975
Creator: Laats, E. T.; MacDonald, P. E. & Quapp, W. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

RELAP5 based engineering simulator

Description: The INEL Engineering Simulation Center was established in 1988 to provide a modern, flexible, state-of-the-art simulation facility. This facility and two of the major projects which are part of the simulation center, the Advance Test Reactor (ATR) engineering simulator project and the Experimental Breeder Reactor II (EBR-II) advanced reactor control system, have been the subject of several papers in the past few years. Two components of the ATR engineering simulator project, RELAP5 and the Nucl… more
Date: January 1, 1990
Creator: Charlton, T.R.; Laats, E.T. & Burtt, J.D.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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