The systems computer code is a key part of the evaluation model for safety analysis of nuclear reactors. The systems code utilizes a set of governing equation that is simplified from the fundamental Navier-Stokes equations and closure models to describe the transport of mass, momentum, and energy of single phase or multiphase fluid throughout the reactor coolant systems. Following the Evaluation Model Development and Assessment Process, an assessment matrix is established where Separate Effects Tests and Integral Effects Tests are selected based on phenomena identification and ranking table. The purpose of the assessment matrix is to validate the systems code against the important phenomena for the safety analysis. The code biases and uncertainties are established and the effect of scale could then be determined. The assessment matrices of major systems codes, RELAP5/MOD3, TRACE Ver.5.0 and WCOBRA/TRAC-TF2, for the reactor safety analysis are reviewed and compared in this study for the Loss of Coolant Accident (LOCA) safety analysis perspectives. The scenarios are divided into small break LOCA and large break LOCA. The phenomena bases of the separate effects tests in those assessment matrices are discussed following its PIRT. The comparison demonstrates the capability of each systems code.
A Thermal Hydraulic Phenomenon Based Review on the Validation Matrix of Several Reactor Safety Analysis Systems Codes
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Liao, J, & Fortune, SE. "A Thermal Hydraulic Phenomenon Based Review on the Validation Matrix of Several Reactor Safety Analysis Systems Codes." Proceedings of the 2016 24th International Conference on Nuclear Engineering. Volume 3: Thermal-Hydraulics. Charlotte, North Carolina, USA. June 26–30, 2016. V003T09A088. ASME. https://doi.org/10.1115/ICONE24-61136
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