The validation of heat transfer models of safety analysis codes such as TRACE is very important due to the strong interaction of the thermal hydraulics parameters with the core neutronics. TRACE is the reference system code of the US NRC for LWR. It is being developed and extensively validated within the international CAMP-program. In this paper, the validation of heat transfer models of TRACE related to the prediction of the critical power is presented. The validation is based on a large number of critical power tests performed in the NUPEC BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility in Japan. These tests were analysed with the TRACE Version 5 RC 2. The comparison of predictions with the experimental data shows good agreement. The developed TRACE model and the comparison of experimental data with code results will be presented and discussed.
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17th International Conference on Nuclear Engineering
July 12–16, 2009
Brussels, Belgium
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4353-6
PROCEEDINGS PAPER
Post-Test Investigations of BFBT Critical Power Tests With Trace
Marc Thieme,
Marc Thieme
Westinghouse Electric Germany GmbH, Mannheim, Germany
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Wolfgang Tietsch,
Wolfgang Tietsch
Westinghouse Electric Germany GmbH, Mannheim, Germany
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Rafael Macian,
Rafael Macian
Technical University Munich, Garching, Germany
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Victor Hugo Sanchez Espinoza
Victor Hugo Sanchez Espinoza
Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany
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Marc Thieme
Westinghouse Electric Germany GmbH, Mannheim, Germany
Wolfgang Tietsch
Westinghouse Electric Germany GmbH, Mannheim, Germany
Rafael Macian
Technical University Munich, Garching, Germany
Victor Hugo Sanchez Espinoza
Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany
Paper No:
ICONE17-75594, pp. 503-514; 12 pages
Published Online:
February 25, 2010
Citation
Thieme, M, Tietsch, W, Macian, R, & Sanchez Espinoza, VH. "Post-Test Investigations of BFBT Critical Power Tests With Trace." Proceedings of the 17th International Conference on Nuclear Engineering. Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance. Brussels, Belgium. July 12–16, 2009. pp. 503-514. ASME. https://doi.org/10.1115/ICONE17-75594
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