Verification of the three-dimensional thermal-hydraulic models of the TRAC accident-analysis code. [PWR]

The Transient Reactor Analysis Code (TRAC) being developed at Los Alamos National Laboratory provides a best-estimate prediction of the response of light water reactors or test facilities to postulated accident sequences. One of the features of the code is the ability to analyze the vessel and its heated core in three dimensions. The code is being used to analyze the results of tests in a large-scale reflood test facility built in Japan, known as the Cylindrical Core Test Facility (CCTF). Two test runs have been analyzed that are useful for verification of the three-dimensional analysis capability of the TRAC code. One test began with an initial temperature skew across the heated core. The second test had a large radial power skew between the central and peripheral assemblies. The good agreement between the calculation and the experiment for both of these experiments demonstrates the three-dimensional analysis capability of the TRAC code.