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Methods for the radioactive release estimation under DEC-A conditions

•The generic BWR-4 model for the ASTEC code was improved to analyze thermal hydraulic processes, fission product release and its transportation.•The verification of the ASTEC calculations was provided using TRANSURANUS fuel performance code.•The ASTEC core nodalisation was updated by recalculating t...

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Bibliographic Details
Published in:Annals of nuclear energy 2024-01, Vol.195, p.110143, Article 110143
Main Authors: Kaliatka, Tadas, Kačegavičius, Tomas, Kaliatka, Algirdas, Povilaitis, Mantas, Tidikas, Andrius, Slavickas, Andrius
Format: Article
Language:English
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Summary:•The generic BWR-4 model for the ASTEC code was improved to analyze thermal hydraulic processes, fission product release and its transportation.•The verification of the ASTEC calculations was provided using TRANSURANUS fuel performance code.•The ASTEC core nodalisation was updated by recalculating the relative power and the number of fuel assemblies in each concentric ring.•Fission product release was reduced compared to the results obtained with the initial model. After the Fukushima Daiichi Nuclear Power Plant accident, the importance was raised to strengthen the assessment of the safety level of Nuclear Power Plants (NPP) by considering situations more severe than those integrated during the design of the plants. Design Extension Conditions (DEC) term was introduced by IAEA and WENRA. One of the most important subjects in the analysis of these conditions is the evaluation of radiological consequences. The Reduction of Radiological Accident Consequences (R2CA) collaborative project started in 2019 in the frame of the Horizon-2020 Program of the European Commission. The project addresses a broad scope of LWR designs (Gen II, III and III +) through the analyses of bounding scenarios of Loss Of Coolant Accidents (LOCA) and Steam Generator Tube Rupture (SGTR) transients and explores DBA and DEC-A conditions. The Lithuanian Energy Institute is taking part in this project as well. The part of work provided in the frame of R2CA project is presented in this article. The generic BWR-4 type reactor with MARK-I containment was analyzed in the case of DEC-A conditions. In this analysis the main thermal hydraulic processes, fission product release and its transport from the broken loop through containment to the environment were investigated. The number of ruptured fuel assemblies is the key parameter determining the fission product release from the core. The methodology which would help more precisely evaluate the number of failed fuel assemblies is presented in this work. The methodology consists of the application of integral severe accident code ASTEC and fuel performance code TRANSURANUS. Analysis using ASTEC code showed the importance of the reactor core nodalisation scheme. Taking iterative calculations, the relative power to which fuel assemblies remain intact at the selected DEC-A condition scenario was obtained. These results were verified with TRANSURANUS code calculations considering the uncertainties of the most relevant parameters. Based on TRANSURANUS
ISSN:0306-4549
1873-2100
DOI:10.1016/j.anucene.2023.110143