Conference Proceedings Paper
Comparative Analysis of CTF and Trace Thermalhydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database
NURETH-14 - 2011 September 25-30
Maria Avramova (The Pennsylvania State University)
Alexander Velazquez-Lozada (U.S. Nuclear Regulatory Commission)
Adam Rubin (The Pennsylvania State University)
The international OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark has been established to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFD) codes and to encourage advancement in the analysis of fluid flow in rod bundles. The aim is to improve the reliability of the nuclear reactor safety margin evaluations. The benchmark is based on one of the most valuable databases identified for the thermal-hydraulics modeling, which was developed by the Nuclear Power Engineering Corporation (NUPEC) in Japan. The database includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurized Water Reactor (PWR) fuel assembly. Part of this database is made available for the international PSBT benchmark activity. The PSBT benchmark team is organized based on the collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety organization (JNES) including the participation and support of the U.S. Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. On behalf of the PSBT benchmark team, PSU in collaboration with US NRC is performing supporting calculations of the benchmark exercises using its in-house advanced thermalhydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of the well-known and widely used code COBRA-TF whose models have been continuously improved and validated over the last years at the Reactor Dynamics and Fuel Management Group (RDFMG) at PSU. TRACE is a reactor systems code developed by the U.S. Nuclear Regulatory Commission to analyze transient and steady-state thermal-hydraulic behavior in Light Water Reactors (LWRs) and it has been designed to perform best-estimate analyses of loss-of-coolant accidents (LOCAs), operational transients, and other accident scenarios in PWRs and boiling light-water reactors (BWRs). The paper presents the CTF and TRACE models for the exercises of the void distribution phase of the OECD/NRC PSBT benchmark. Code-to-code and code-to-data comparisons are provided along with a discussion of the void generation and void distribution models available in the two codes.
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