In nuclear safety, the Best-Estimate (BE) codes may be used in demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters, which comes mainly from the lack of knowledge of the input parameters, is evaluated by estimating the 95% percentile with a high degree of confidence. IRSN, technical support of the French Safety Authority, develop a method of uncertainty propagation and chose to apply it to the calculation of the Peak Cladding Temperature (PCT) with CATHARE-2 V2.5 code during a Large Break (LB) LOCA event for ZION, a 4-loop PWR of Westinghouse design. As a general rule the Global Sensitivity Analysis (GSA) is done with linear correlation coefficients. This paper presents a new approach to perform a more accurate GSA to determine and to classify the main uncertain parameters: the SOBOL methodology. This technique requires simulating many sets of parameters to propagate uncertainties correctly, which makes of it a time-consuming approach. Therefore, it is natural to replace the complex computer code by an approximate mathematical model, called response surface or surrogate model. Kriging methodology (with simulated annealing optimization) for its construction and the SOBOL methodology for the GSA are used. The paper presents the application of the previously described methodology on a LB-LOCA scenario in ZION reactor, associated with 54 input parameters. The output is the first maximum peak cladding temperature of the fuel. Results show that the methodology could be applied to both high-dimensional complex problems and real nuclear power plant calculations.
- Nuclear Engineering Division
Sensitivity Analysis by the Use of a Surrogate Model During Large Break LOCA on ZION Nuclear Power Plant With CATHARE-2 V2.5 Code
Fouet, F, & Probst, P. "Sensitivity Analysis by the Use of a Surrogate Model During Large Break LOCA on ZION Nuclear Power Plant With CATHARE-2 V2.5 Code." Proceedings of the 17th International Conference on Nuclear Engineering. Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications. Brussels, Belgium. July 12–16, 2009. pp. 205-212. ASME. https://doi.org/10.1115/ICONE17-75661
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