France-Japan collaboration on severe accident studies in sodium-cooled fast reactors, 1; Severe accident scenarios assessment

Onoda, Yuichi; Ishida, Shinya; Fukano, Yoshitaka; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Shibata, Akihiro*; Bertrand, F.*; Seiler, N.*

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05

https://doi.org/10.1080/00223131.2023.2243949

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00395_1 - 16-00395_9, 2017/04

https://doi.org/10.1299/transjsme.16-00395

no abstracts in English

Improvement of transient analysis method of a sodium-cooled fast reactor with FAIDUS fuel sub-assemblies

Ohgama, Kazuya; Kawashima, Katsuyuki*; Oki, Shigeo

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

In order to evaluate transient behavior of Japan sodium-cooled fast reactor (JSFR) with fuel sub-assemblies with the innerduct structure (FAIDUS) precisely, a new model for a plant dynamics code HIPRAC was developed. In this new model, inner core and outer core channels can be divided into three channels, respectively, such as interior, edge and near innerduct channel, and calculate coolant redistribution and coolant temperature in each channel. Coolant temperature distribution of interior and edge channels calculated by this model was compared with previous study by the general-purpose thermal-hydraulics code -FLOW. Coolant temperature behavior inside the innerduct was analyzed by a commercial thermal hydraulics code STAR-CD ver. 3.26. Based on this result, horizontally-uniformed coolant temperature in the innerduct was assumed as a heat transfer model of the innderduct. Reactivity coefficients for 750 MWe JSFR with low -decontaminated transuranic (TRU) fuel were evaluated. Transient behaviors of an unprotected loss-of-flow (ULOF) accident for JSFR with 750 MWe output calculated by previous and new models were compared. The results showed that the detailed evaluation of coolant temperature improved overestimation of the coolant temperature and coolant temperature feedback reactivity of the peripheral channels including coolant inside the innerduct and in the inter-wrapper gap.

A Computer code for analysis of core transient behavior in a Na-cooled metal fuel fast reactor; Modification of the code EXCURS and sample calculation

Okajima, Shigeaki; ; Mukaiyama, Takehiko

JAERI-M 92-031, 81 Pages, 1992/03

JAERI-M-92-031.pdf:2.29MB

no abstracts in English

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 2; Assessment of initiating phase in ULOF

Kawada, Kenichi; Ishida, Shinya

no journal, , 

no abstracts in English

Development of multi-level simulation system for sodium-cooled fast reactor; Application of coupled 1D-CFD simulation to ULOF test of EBR-II

Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Murakami, Satoshi*; Tanaka, Masaaki

no journal, , 

The multi-level simulation system with 1D-CFD coupling method which enables to evaluate various phenomena from the whole plant dynamics to the local thermal hydraulics has been developed. The numerical simulation of the ULOF test in the experimental fast reactor EBR-II in the U.S. is performed for validation study of the 1D-CFD coupling method, which combines a one-dimensional plant dynamics analysis (1D) code with a computational fluid dynamics (CFD) code. Through the numerical simulation, it was shown that the whole plant response and the multi-dimensional thermal hydraulics in the core upper plenum could be simulated. And the applicability of the 1D-CFD coupling method to plant scale analysis was confirmed in comparison with the experimental results.

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 4; Assessment of transition phase in ULOF

Tobita, Yoshiharu; Suzuki, Toru; Tagami, Hirotaka

no journal, , 

The event progression in the transition phase of ULOF (Unprotected Loss of Flow), which is the representative event in ATWS (Anticipated Transient without Scram) of fast breeding reactor. The existing experimental knowledges on the important phenomena, which dominates the event progression, were adopted in the nominal case. The reactivity lowered with accordance to the fuel discharge through control rod guide tube and the event terminated without prompt criticality. If the uncertainty was considered in the analysis, the reactivity slightly exceeded the prompt criticality, but no mechanical energy was produced.

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 11; Best estimate and uncertainty assessment of PAMR/PAHR phase in ULOF

Sogabe, Joji; Wada, Yusaku*; Suzuki, Toru*; Tobita, Yoshiharu

no journal, , 

no abstracts in English

Model development for blockage of disrupted core materials in flow path

Sogabe, Joji; Kamiyama, Kenji; Tobita, Yoshiharu; Okano, Yasushi

no journal, , 

During severe accidents by an anticipated transient without scram, it is important to evaluate multiphase multi-component flow behavior, when a part of the disrupted core material is discharged outside the disrupted core region through control rod guide tubes. In particular, the blockage behavior of the disrupted core material in a flow path is an important phenomenon that affects the amount of relocated fuels (the fuel discharged outside the disrupted core region and the fuel remaining in the disrupted core region). A fast reactor safety analysis code, SIMMER, is currently being developed for application to the post-accident material relocation (PAMR) phase. In the paper, aiming at actual reactor analyses for the PAMR phase of the SIMMER code, a model for the blockage in the flow path for possible phenomena in the PAMR phase. The model improves the applicability of the SIMMER code to the PAMR phase on the actual reactors.

Development of fundamental numerical simulation system for integrated safety evaluation in various innovative sodium-cooled fast reactor, 18; Analysis of SA progress in PRISM-type reactor

Fuchita, Sho*; Fujimata, Kazuhiro*; Abe, Takashi*; Nakahara, Hirotaka*; Aoyagi, Mitsuhiro; Ishida, Shinya

no journal, , 

no abstracts in English