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Onoda, Yuichi; Tobita, Yoshiharu; Okano, Yasushi
IAEA-TECDOC-2079, p.215 - 225, 2025/00
The analysis methodologies for the evaluation of unprotected loss of flow accident on sodium-cooled fast reactor in Japan Atomic Energy Agency (JAEA) are briefly explained focusing on the mechanical consequences during expansion phase of the accident. JAEA developed the analysis methodologies for the evaluation of energetics and divided the analysis process into following three: 1) analysis of converting the heat generated into the mechanical energy with SIMMER code, 2) analysis of the structural response of the reactor vessel with AUTODYN code, and 3) analysis of the amount of sodium ejected onto the top shield through the gaps between shield plugs with PLUG code. Pressure-volume relation of the CDA bubble, which is the mixture of gas (fuel, steel vapor and fission gas) and molten core material, obtained by SIMMER calculation is used as the input for structural response analysis with AUTODYN. Pressure history exerted on the lower surface of the top shield obtained by SIMMER calculation is used as the input for PLUG. These analysis codes are validated by simulating the dominant phenomena that significantly affect the results in each calculation. We applied these analysis methodologies developed by JAEA to the reactor case analyses and confirmed their applicability.
Tobita, Yoshiharu; Tagami, Hirotaka; Ishida, Shinya; Onoda, Yuichi; Sogabe, Joji; Okano, Yasushi
IAEA-TECDOC-2079, p.72 - 84, 2025/00
Since the fast reactor core is not in the maximum reactivity configuration, a hypothetical core disruptive accident could lead to the prompt criticality due to a change in the core material configuration, and the resulting energy generation has been one of the issues in fast reactor safety, and therefore appropriate measures are needed to mitigate and contain the effect of energy generated in the accident. In order to assess the effectiveness of these mitigative measures, a set of computer codes to analyze the event progressions and energy generation behavior in the ATWS of fast reactors have been developed, maintained, and improved under international collaboration in JAEA. Since the important physical phenomena, which govern the event progression, vary along with the progression of the accident, the whole accident process is divided into several phases in the analysis of accident, and dedicated analysis methods for each phase are provided to analyze the event progression in each phase. The organization and overview of these analysis methods are described in this paper. As a representative example of the validation approaches in applying these analysis methods to the reactor safety assessment in licensing procedure in Japan, the validation studies to confirm the applicability to reactor analysis of the SIMMER code for analyzing core material movement and reactor power, which is important to analyze the energy generation in the accident, are presented in the paper. The validation studies of the SIMMER code confirmed the applicability of SIMMER to the reactor analysis, while the critical phenomena that the effect of their uncertainty in the reactor analysis should be checked were also recognized.
Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Akaev, A.*; Mikisha, A.*; Baklanov, V.*; Vurim, A.*
no journal, ,
Yamano, Hidemasa
no journal, ,
Four phases are categorized in the core degradation process: Initiating, transition, material relocation and cooling phase. In this report, specific design features for each phase are shown as well as the safety assessment results to demonstrate their effectiveness.
