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Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Mori, Takero; Yoshimura, Kazuo; Yoshikawa, Ryuji; Kikuchi, Norihiro; Matsushita, Kentaro; Mochinaga, Shota; Ezure, Toshiki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR26) (Internet), 9 Pages, 2026/05
This paper presents the outline of the evaluation support program on the evaluations of the thermal-hydraulics related issues in the conceptual design study of the sodium-cooled fast reactor for demonstration in Japan, utilizing the technical basis on the simulations and evaluations configured in the integrated evaluation system named ARKADIA that has been developed with experiences through the development of the SFRs and related R&Ds in Japan Atomic Energy Agency. Moreover, as the extension of the function of the ARKADIA, the possible framework for SFR lifecycle development introducing the Digital-Twin for the MBD with ARKADIA is presented with an example of a whole plant coupled analysis model developed in ARKADIA.
Mochinaga, Shota; Doda, Norihiro; Iwasaki, Takashi*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
We have been developing the core hot spot (CHS) evaluation (e.g., maximum cladding temperature in a fuel pin) for the safety evaluation of a pool-type demonstration sodium-cooled fast reactor. Since changing analysis conditions due to uncertainty parameters during natural circulation decay heat removal affects the generation of buoyancy, the CHS evaluation based on the statistical method is effective. In this study, we conducted the phenomena identification and ranking table (PIRT) in case of the loss of external power events to extract important phenomena and those related uncertainty factors, and we quantitatively evaluated the important factors with sensitivity analyses.
Mochinaga, Shota; Doda, Norihiro; Iwasaki, Takashi*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
A core hot spot (CHS) evaluation method (e.g., maximum cladding temperature in a fuel pin) is being developed for the safety evaluation of a pool-type demonstration sodium-cooled fast reactor. During decay heat removal by natural circulation (NC), buoyancy drives the coolant flow, making the influences of uncertainty factors complicated relationship with thermal-hydraulic behavior. Therefore, a statistical approach to the CHS evaluation is effective in the NC transient. In this study, the fuel pin model including effects of coolant flow distribution and buoyancy within the assembly was newly constructed and the applicability of the statistical CHS method for several NC events was confirmed.
Tanaka, Masaaki; Ezure, Toshiki; Yoshikawa, Ryuji; Doda, Norihiro; Mori, Takero; Hamase, Erina; Yoshimura, Kazuo; Kikuchi, Norihiro; Matsushita, Kentaro; Mochinaga, Shota
no journal, ,
This brief presents the outline of the evaluation support program on the evaluations of the thermal-hydraulics related issues in the conceptual design study of the sodium-cooled fast reactor for demonstration in Japan over the five years, utilizing the technical basis on the simulations and evaluations configured in the integrated evaluation system named ARKADIA that has been developed with experiences through the development of the SFRs and related R&Ds in Japan Atomic Energy Agency.
Mochinaga, Shota; Doda, Norihiro; Iwasaki, Takashi*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
We have been developing the core hot spot (CHS) evaluation (e.g., maximum cladding temperature in a fuel pin) for the safety evaluation of a pool-type demonstration sodium-cooled fast reactor (SFR). Since changing analysis conditions due to uncertainty parameters during natural circulation decay heat removal affects the generation of buoyancy, the CHS evaluation based on the statistical method is effective. In this study, CHS evaluation procedure was organized for a pool-type SFR and its applicability was investigated. First trial results of applying the statistical CHS evaluation procedure to the SFR demonstration reactor (tentative specification) during rated operation are reported.
Mochinaga, Shota
no journal, ,
The conceptual design of a pool-type demonstration sodium-cooled fast reactor (SFR) has been carried out. The core hot spot (CHS) evaluation (e.g., maximum cladding temperature in a fuel pin) considering various uncertainties related to the temperature rise in the core and abnormal transients is required in the safety design. Since changing analysis conditions due to uncertainty parameters affects the generation of buoyancy, the CHS evaluation based on the statistical method is effective. In this study, CHS evaluation procedure was organized for a poo-type SFR and its applicability was investigated. Establishing the statistical CHS evaluation method with the concrete procedure can ensure reliability of the safety assessment processes and results.