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

[Licentiatavhandling]

In recent years, more realistic safety analyses of nuclear reactors have been based on best estimate (BE) computer codes. The need to validate and refine BE codes that are used in the predictions of relevant reactor safety parameters, led to the organization of international benchmarks based on high quality experimental data. The OECD/NRC BWR Full‐Size Fine‐Mesh Bundle Test (BFBT) benchmark offers a good opportunity to assess the accuracy of thermal‐hydraulic codes in predicting, among other parameters, single and two phase bundle pressure drops, cross‐sectional averaged void fraction distributions and critical powers under a wide range of system conditions. The BFBT is based on a multi‐rod assembly integral test facility which is able to simulate the high pressure, high temperature fluid conditions found in BWRs through electrically‐heated rod bundles. Since code accuracy is unavoidably affected by models and experimental uncertainties, an uncertainty analysis is fundamental in 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‐hydraulic features of the Westinghouse code POLCA‐T. This is achieved by computing a quantitative validation limit based on statistical uncertainty analysis. This validation theory is applied to some of the benchmark cases of the following macroscopic BFBT exercises: 1) Single and two phase bundle pressure drops, 2) Steady‐state cross‐sectional averaged 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 Hypercube Sampling (LHS) strategy over simple random sampling (SRS). LHS allows a much better coverage of the input uncertainties than SRS because it densely stratifies across the range of each input probability distribution. The aim here 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 pump trip scenario. It is shown that the replicated void fraction mean (either in steady‐state or transient conditions) has less 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 fraction uncertainty limits achieved with SRS by running 458 calculations (sample size required to cover 95% of 8 uncertain input parameters with a 95% confidence), result in the same uncertainty limits achieved by LHS with only 100 calculations. 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 uncertainties of important BWR parameters at a bundle level are assessed. Keywords: Thermal‐hydraulic codes, uncertainty and sensitivity analysis, BFBT benchmark, Latin Hypercube sampling, simple random sampling, reactor safety analysis

**Nyckelord: **Thermal‐hydraulic codes, uncertainty and sensitivity analysis, BFBT benchmark, Latin Hypercube sampling, simple random sampling, reactor safety analysis

Denna post skapades 2011-06-14.

CPL Pubid: 141699