Validating the BISON fuel performance code to integral LWR experiments

• Published in: Nuclear engineering and design Vol. 301; no. C; pp. 232 - 244
• Main Authors: Williamson, R.L., Gamble, K.A., Perez, D.M., Novascone, S.R., Pastore, G., Gardner, R.J., Hales, J.D., Liu, W., Mai, A.
• Format: Journal Article
• Language: English
• Published: United States Elsevier B.V 01.05.2016; Elsevier
• Subjects: BISON; Fission; Fuels; Light water reactors; Mathematical analysis; Mathematical models; MATHEMATICS AND COMPUTING; NUCLEAR FUEL CYCLE AND FUEL MATERIALS; Nuclear power generation; Ramps; validation; verification; USA, Idaho
• DOI: 10.1016/j.nucengdes.2016.02.020
• ISSN: 0029-5493

•The BISON multidimensional fuel performance code is being validated to integral LWR experiments.•Code and solution verification are necessary prerequisites to validation.•Fuel centerline temperature comparisons through all phases of fuel life are very reasonable.•Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties.•Rod diameter comparisons are not satisfactory and further investigation is underway. BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons are not satisfactory and have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.