Validation of frey for the safety analysis of LWR fuel using transient fuel rod experiments
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The analysis and comparison of severe light water reactor transient experiments are presented from the FREY verification and validation effort. The purpose of this study was to validate the predictive capabilities of the code for severe transient analysis. The FREY code, developed under the sponsorship of the Electric Power Research Institute, uses a two-dimensional finite-element computational method for the thermomechanical analysis of LWR fuel rods under steady state and transient conditions. A total of 10 test fuel rods from experimental programs conducted in both the Power Burst Facility and the Transient Reactor Test Facility have been used in this study. The fuel rods were selected from the following test programs: Power Coolant Mismatch Tests, PCM-2 and PCM-4: Reactivity Initiated Accident Test, RIA 1-2; Loss-of-Coolant Accident Test, LOC-3; First Fuel Rod Failure Test, FRF-1; and Irradiation Effects Test, IE-3. The test programs used in this study cover a large range of code applications for severe transient analysis. The methods used to model the fuel, cladding, and coolant geometry are discussed in addition to experimental data comparisons. The results of the PCM-2, RIA 1-2, and FRF-1 analyses are presented to highlight the full two-dimensional modeling capabilities of FREY and to compare the thermal and mechanical measurements with FREY's prediction. The comparisons show good general agreement, with a tendency for FREY to overpredict the peak cladding surface temperature for a few cases where strong three-dimensional effects have been identified. 1990.
Nuclear Engineering and Design
author list (cited authors)
Montgomery, R. O., Rashid, Y. R., George, J. A., Peddicord, K. L., & Lin, C. L.
complete list of authors
Montgomery, RO||Rashid, YR||George, JA||Peddicord, KL||Lin, CL