4.6 Article

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part I: SCIANTIX

Journal

NUCLEAR ENGINEERING AND TECHNOLOGY
Volume 54, Issue 8, Pages 2771-2782

Publisher

KOREAN NUCLEAR SOC
DOI: 10.1016/j.net.2022.02.011

Keywords

Oxide nuclear fuel; Fission gas behaviour; Radioactive release; ANS 5.4; SCIANTIX

Funding

  1. Euratom research and training programme [847656]

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When assessing the radiological consequences of postulated accident scenarios, determining the amount of radioactive fission gas accumulated in the fuel rod free volume is crucial. This study reviewed the state-of-the-art semi-empirical approach and compared it with a mechanistic approach, providing a satisfactory agreement with measurement and demonstrating the model soundness. The research is important for evaluating the behavior of fuel rods under normal and off-normal conditions.
When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as inter-related phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS. (C) 2022 Korean Nuclear Society, Published by Elsevier Korea LLC.

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