Validating the BISON fuel performance code to integral LWR experiments

Journal Article · Thu Mar 24 00:00:00 EDT 2016 · Nuclear Engineering and Design

DOI:https://doi.org/10.1016/j.nucengdes.2016.02.020· OSTI ID:1357498

Williamson, R. L. ^[1]; Gamble, K. A. ^[1]; Perez, D. M. ^[1]; Novascone, S. R. ^[1]; Pastore, G. ^[1]; Gardner, R. J. ^[1]; Hales, J. D. ^[1]; Liu, W. ^[2]; Mai, A. ^[2]

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1. Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation
2. ANATECH Corporation, San Diego, CA (United States)

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the 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. Our 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. In the initial rod diameter comparisons they were unsatisfactory and have lead 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.

• Research Organization: Idaho National Laboratory, Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE; USDOE Office of Nuclear Energy (NE)
• Grant/Contract Number: AC07-05ID14517
• OSTI ID: 1357498
• Alternate ID(s): OSTI ID: 22561559; OSTI ID: 1488994
• Report Number(s): INL/JOU--15-36919; PII: S0029549316000789
• Journal Information: Nuclear Engineering and Design, Journal Name: Nuclear Engineering and Design Journal Issue: C Vol. 301; ISSN 0029-5493
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
97 MATHEMATICS AND COMPUTING
BISON
validation
verification