Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

Barani, T.; Bruschi, E.; Pizzocri, D.; Pastore, G.; Van Uffelen, P.; Williamson, R. L.; Luzzi, Lelio

doi:10.1016/j.jnucmat.2016.10.051

Title: Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

Journal Article · Tue Jan 03 00:00:00 EST 2017 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2016.10.051· OSTI ID:1372265

^[1]; Bruschi, E. ^[1]; Pizzocri, D. ^[1]; Pastore, G. ^[2]; Van Uffelen, P. ^[3]; Williamson, R. L. ^[2]; Luzzi, Lelio ^[1]

Politecnico di Milano, Milano (Italy)
Idaho National Lab. (INL), Idaho Falls, ID (United States)
European Commission, Karlsruhe (Germany)

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.

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Research Organization:: Idaho National Laboratory (INL), Idaho Falls, ID (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: AC07-05ID14517

OSTI ID:: 1372265

Report Number(s):: INL/JOU-16-39206; PII: S002231151630544X

Journal Information:: Journal of Nuclear Materials, Vol. 486, Issue C; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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Cited by: 28 works

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