Nuclear fuel cladding failure in small-break loss of coolant accidents safety analysis for three-loop pressurized water reactors
Safety Instruction (IS-37) was issued by the Spanish Nuclear Regulatory Commission on January 21st 2015, and plays a crucial role in nuclear safety by providing guidelines for accident analysis in nuclear power plants. Historically, Condition III accidents (Operational events anticipated to occur wi...
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| Tipo de recurso: | tesis de maestría |
| Fecha de publicación: | 2025 |
| País: | España |
| Institución: | Universitat Politècnica de Catalunya (UPC) |
| Repositorio: | UPCommons. Portal del coneixement obert de la UPC |
| Idioma: | inglés |
| OAI Identifier: | oai:upcommons.upc.edu:2117/431877 |
| Acceso en línea: | https://hdl.handle.net/2117/431877 |
| Access Level: | acceso embargado |
| Palabra clave: | Nuclear power plants -- Safety measures Pressurized water reactors Nuclear engineering -- Safety measures Centrals nuclears -- Mesures de seguretat Reactors nuclears d'aigua a pressió Enginyeria nuclear -- Mesures de seguretat Àrees temàtiques de la UPC::Energies::Energia nuclear |
| Sumario: | Safety Instruction (IS-37) was issued by the Spanish Nuclear Regulatory Commission on January 21st 2015, and plays a crucial role in nuclear safety by providing guidelines for accident analysis in nuclear power plants. Historically, Condition III accidents (Operational events anticipated to occur with a frequency ranging from 0.1/reactor-year to 0.01/reactor-year), such as Small Break Loss of Coolant Accidents (SBLOCAs), were assumed to be inherently bounded by more severe Category IV accidents (Unexpected accidents during the operational life of the facility that can result in the release of significant amounts of radioactive material) like Large Break Loss of Coolant Accident (LBLOCA) and therefore, their radiological consequences had not been directly assessed in Westinghouse Pressurized Water Reactors (W-PWR) under Spanish regulatory standards. However, the assumptions applied to Category IV accidents are usually very conservative and penalizing in nature, which, for Category III accidents, becomes unrealistic and extremely over-penalizing from the radiological point of view. As a consequence of the newly established regulatory requirements, Spanish Westinghouse facilities were required to conduct detailed computational analyses to demonstrate compliance with dose criteria for an SBLOCA event, among other accidents. For the simulations to be successful, the Spanish facilities should not exceed a rod burst fraction of 33%, as established by the EUR 19256 EN report. This was achieved by studying the peak cladding temperature and the evolution of rod internal pressure along the accident transient, in order to calculate the possible rod failure. However, this simulations where only carried out at Beginning of Life (BOL) conditions, and the calculations were then extrapolated to other core burn-up conditions. This thesis addresses this gap by performing a sensitivity analysis to assess the impact of SBLOCAs on key accident parameters, particularly peak cladding temperature and internal fuel pressure, across various break sizes and fuel burn-up stages. To achieve this, computational simulations have been performed using advanced modeling techniques and simulation programs in order to analyze different SBLOCA scenarios at different break sizes and core burn-up levels. These transients also combine various fuel compositions, cladding materials and decay heat profiles, which will be explained in detail. The findings reveal that NOTRUMP results, which describe system-wide thermal-hydraulic behavior, are largely independent of burn-up, cladding material, and fuel composition due to core reactivity control and the geometric size of the studied system. In contrast, SBLOCTA results exhibit significant dependencies on these factors, influencing critical parameters such as peak cladding temperature and fuel rod internal pressure. The predominant factor in the evolution of the transients is found to be the decay heat profile. To be precise, two decay heat models are applied in this work: ANSI/ANS-5.1-1979+2σ and ANSI/ANS5.1-1971+20%, the latter being the standard decay heat profile used in safety analysis. One of the analysis objectives was to determine whether previous assumptions regarding SBLOCA radiological consequences were overly conservative by using a very penalizing decay heat model, and whether a more relaxed and less penalizing decay heat profile could provide more realistic, yet equally safe, assessments. A comparison of residual heat models further demonstrates that the ANSI/ANS-5.1-1971+20% model leads to highly conservative estimates, resulting in fuel cladding rupture across a wide range of conditions. On the contrary, the ANSI/ANS-5.1-1979+2σ model provides a more balanced assessment, avoiding overly pessimistic results while maintaining necessary safety margins. Under this refined approach, no cladding rupture is observedfor any of the cases studied, ensuring compliance with the established regulatory limit of 33% cladding failure in SBLOCTA scenarios. |
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