Uncertainty and Sensitivity Analysis Applied to the Validation of BWR Bundle Thermal-Hydraulic Calculations

Sammanfattning: In recent years, more realistic safety analyses of nuclear reactors have been based on best estimate (BE) computercodes. The need to validate and refine BE codes that are used in the predictions of relevant reactor safetyparameters, led to the organization of international benchmarks based on high quality experimental data. TheOECD/NRC BWR Full‐Size Fine‐Mesh Bundle Test (BFBT) benchmark offers a good opportunity to assess theaccuracy of thermal‐hydraulic codes in predicting, among other parameters, single and two phase bundle pressuredrops, cross‐sectional averaged void fraction distributions and critical powers under a wide range of systemconditions. The BFBT is based on a multi‐rod assembly integral test facility which is able to simulate the highpressure, high temperature fluid conditions found in BWRs through electrically‐heated rod bundles. Since codeaccuracy is unavoidably affected by models and experimental uncertainties, an uncertainty analysis is fundamentalin order to have a complete validation study.This work has two main objectives. The first one is to enhance the validation process of the thermal‐hydraulicfeatures of the Westinghouse code POLCA‐T. This is achieved by computing a quantitative validation limit based onstatistical uncertainty analysis. This validation theory is applied to some of the benchmark cases of the followingmacroscopic BFBT exercises: 1) Single and two phase bundle pressure drops, 2) Steady‐state cross‐sectionalaveraged void fraction, 3) Transient cross‐sectional averaged void fraction and 4) Steady‐state critical power tests.Sensitivity analysis is also performed to identify the most important uncertain parameters for each exercise.The second objective consists in showing the clear advantages of using the quasi‐random Latin HypercubeSampling (LHS) strategy over simple random sampling (SRS). LHS allows a much better coverage of the inputuncertainties than SRS because it densely stratifies across the range of each input probability distribution. The aimhere is to compare both uncertainty analyses on the BWR assembly void axial profile prediction in steady‐state,and on the transient void fraction prediction at a certain axial level coming from a simulated re‐circulation pumptrip scenario. It is shown that the replicated void fraction mean (either in steady‐state or transient conditions) hasless variability when using LHS than SRS for the same number of calculations (i.e. same input space sample size)even if the resulting void fraction axial profiles are non‐monotonic. It is also shown that the void fractionuncertainty limits achieved with SRS by running 458 calculations (sample size required to cover 95% of 8 uncertaininput parameters with a 95% confidence), result in the same uncertainty limits achieved by LHS with only 100calculations. These are thus clear indications on the advantages of using LHS.Finally, the present study contributes to a realistic analysis of nuclear reactors, in the sense that the uncertaintiesof important BWR parameters at a bundle level are assessed.Keywords: Thermal‐hydraulic codes, uncertainty and sensitivity analysis, BFBT benchmark, Latin Hypercubesampling, simple random sampling, reactor safety analysis

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