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Onoda, Yuichi; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
The objective of this study is to develop an effectiveness evaluation methodology of the measures for improving resilience of nuclear structures at ultra-high temperature by using the failure mitigation technology. At the beginning, to identify the accident sequences having the potential to improve resilience, the characteristics of a next-generation loop-type sodium-cooled fast reactor (SFR) in Japan has been investigated by analyzing the event tree of level-1 and level-2 probabilistic risk assessment. As a result, event sequences of loss of heat removal systems (LOHRS) are identified. The effectiveness of the measures for improving resilience is evaluated by quantifying the reduction rate of core damage frequency before and after the introduction of the measures for improving resilience for all the accident sequences leading to LOHRS. To examine applicability of the developed methodology, a trial evaluation has conducted for a next-generation loop-type SFR in Japan. Through the applicability examined, the method for the effectiveness evaluation was developed successfully. The refinement of the conditional success probability of the measures for improving resilience is the future work.
Onoda, Yuichi; Yamano, Hidemasa
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 9 Pages, 2022/10
In Japan, sodium-cooled fast reactor design takes In-Vessel Retention (IVR) strategy to stably cool damaged core materials in the reactor vessel during a severe accident with various design measures. Although a possibility to fail IVR is extremely low, a probabilistic risk assessment study needs a wide variety of scenarios including the IVR failure. Therefore, in order to study a wide range of event spectra related to stable cooling of debris in the reactor vessel, this study numerically investigated the deformation and failure behavior of the reactor vessel due to the debris deposited onto the skirt of the core catcher using the FINAS-STAR structural analysis code. The analyses are conducted in two cases of power density with the aim of investigating failure conditions of the bottom of the reactor vessel. Reactor vessel deforms significantly when the temperature reaches about 1100 
C and the reactor vessel reaches the failure criteria in high-power-density case.
Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture. The target plant is a pool-type SFR of about 600 MWe class equipped with an axially homogeneous core currently under development in Japan. In the pool-type SFR, the primary system piping connects primary pump and the high-pressure sodium plenum located at the inlet of fuel sub-assemblies and called "pump/diagrid link". Because this piping is submerged in the reactor vessel, it is difficult to detect small scale sodium leakage in this piping, and thus a certain large pipe break like guillotine should be assumed and evaluated as a design basis event. In order to confirm the detectability of pump/diagrid link rupture by safety protection system signals, a series of analyses of the guillotine break for a pump/diagrid link were carried out. Sensitivity study had also been performed to consider the uncertainty of the reactivity coefficient in the analyses. The sufficiency of the function of the plant protection system against the pump/diagrid link rupture was confirmed by the analysis results that at least two signals are transmitted for the detection of the event, which is the development target of the plant protection system in pool-type SFR.
Onoda, Yuichi; John Arul, A.*; Klimonov, I.*; Danting, S.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 13 Pages, 2022/04
Arokiaswamy, J. A.*; Batra, C.*; Chang, J. E.*; Garcia, M.*; Herranz, L. E.*; Klimonov, I. A.*; Kriventsev, V.*; Li, S.*; Liegeard, C.*; Mahanes, J.*; et al.
IAEA-TECDOC-2006, 380 Pages, 2022/00
The IAEA coordinated research project on "Radioactive Release from the Prototype Sodium Cooled Fast Reactor under Severe Accident Conditions" was devoted to realistic numerical simulation of fission products and fuel particles inventory inside the reference sodium cooled fast reactor volumes under severe accident conditions at different time scales. The scope of analysis was divided into three parts, defined as three work packages (WPs): (1) in-vessel source term estimation; (2) primary system/containment system interface source term estimation; and, (3) in-containment phenomenology analysis. Comparison of the results obtained in WP-1 indicates that the release fractions of noble gases and cesium radionuclides, and fractions of radionuclides released to the cover gas are in a good agreement. In the analysis using a common pressure history in WP-2, the results were in good agreement indicating that the accuracy of the analysis method of each institution is almost the same. The standalone case, which uses a set of pre-defined release fractions, was defined for WP-3 which enables to decouple this part of analysis from previous WPs. There is broad consensus among the predicted results by all the participants in WP-3.
Nishino, Hiroyuki; Onoda, Yuichi; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 10 Pages, 2021/10
The objective of this study is to develop an effectiveness evaluations technology of the measures for improving resilience of nuclear structures against excessive earthquake by introducing the fracture control concept. After analyzing event tree in previous studies of PRA against earthquake, this study identified sequences of protected loss of heat sink and loss of reactor level induced from excessive earthquake as accident sequences in which improving resilience of nuclear structures become effective. This study focused on important components for safety (e.g., reactor vessel, air coolers, pipes of primary loops in decay heat removal systems, etc.) to be used as countermeasures for improving the resilience. Core damage frequency is selected as an index in evaluating effectiveness of the measures for improving the resilience. Seismic safety margin of the components is assumed to be enlarged when the measures for improving the resilience with the fracture control concept are implemented. Through the trial calculation, reduction effect of the core damage frequency was quantified. The result showed that the measures for improving the resilience are significantly effective for decreasing the core damage frequency in excessive earthquake twice higher than a design basis ground motion. The general concept for the effectiveness evaluations technology was formulated.
Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 11 Pages, 2021/10
The effectiveness evaluations technology of the measures for improving resilience by applying a fracture control concept under ultra-high temperature conditions has developed for prototype sodium-cooled fast reactor Monju as a model plant, and the trial evaluation has conducted using this technology in this paper. The important accident sequences to which the fracture control concept is expected to be applied under ultra-high temperature condition are identified by investigating the results of the existing researches of level-2 probabilistic risk assessment for Monju. Accident sequences categorized in protected loss of heat sink and loss of reactor level are both identified as such important accident sequences which has the potential to prevent core damage. This study has developed the technology to evaluate the effectiveness of improving resilience, where the headings which stand for success or failure of the measures to improve resilience are introduced into the event tree, the branch probability of them is set, and the effectiveness of improving resilience is expressed as the reduction of core damage frequency. As a result of the trial evaluation of the effectiveness for the measures to improve resilience, it is confirmed that core damage frequency can be reduced by applying fracture control concept. The branch probability of the measures to improve resilience proposed in this study is tentatively assigned based on the assumption. This value is expected to be quantified by the forthcoming analyses of the integrity for the reactor vessel structure at ultra-high temperature. The technology developed in this study will be applied for the evaluation of improving resilience of the next generation sodium-cooled fast reactor.
Onoda, Yuichi; Chikazawa, Yoshitaka; Nakamura, Hironori*; Barbier, D.*; Dirat, J.-F.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
no abstracts in English
Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru
Mechanical Engineering Journal (Internet), 4(3), p.16-00597_1 - 16-00597_14, 2017/06
Onoda, Yuichi; Kurisaka, Kenichi; Sakai, Takaaki
Journal of Nuclear Science and Technology, 53(11), p.1774 - 1786, 2016/11
Times Cited Count：3 Percentile：28.28(Nuclear Science & Technology)Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 10 Pages, 2016/06
Onoda, Yuichi; Kurisaka, Kenichi; Sakai, Takaaki
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
Onoda, Yuichi; Fukano, Yoshitaka; Sato, Ikken; Marquie, C.*; Duc, B.*
Journal of Nuclear Science and Technology, 48(2), p.188 - 204, 2011/02
Times Cited Count：10 Percentile：60.88(Nuclear Science & Technology)Two three-pin cluster tests simulating the Unprotected Loss-of-Flow (ULOF) accident of Sodium-cooled Fast Reactors (SFRs) were conducted focusing on fuel relocation and freezing behavior. Based on detailed data evaluation and theoretical interpretation for the three-pin cluster tests, it is concluded that axial fuel relocation and freezing are dominated by local fuel enthalpy, and the relation between fuel dispersal and fuel enthalpy observed in these CABRI tests is basically applicable to the pin-bundle condition. It is also clarified that a fuel/steel mixture tends to create tight blockages near the axial ends of the relocating fuel. Part of the fission gas released from the fuel is expected to be trapped within the bottled-up region between the upper and lower blockages and will keep this region pressurized for a relatively long period.
Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken
Journal of Nuclear Science and Technology, 47(4), p.396 - 410, 2010/04
Times Cited Count：7 Percentile：45.04(Nuclear Science & Technology)Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, J.*
Journal of Nuclear Science and Technology, 46(11), p.1049 - 1058, 2009/11
Times Cited Count：17 Percentile：72.88(Nuclear Science & Technology)Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, J.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 13 Pages, 2009/10
Onoda, Yuichi; Fukano, Yoshitaka; Sato, Ikken; Marquie, C.*; Duc, B.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 15 Pages, 2009/09
Yamano, Hidemasa; Onoda, Yuichi; Tobita, Yoshiharu; Sato, Ikken
Nuclear Technology, 165(2), p.145 - 165, 2009/02
Times Cited Count：27 Percentile：85.04(Nuclear Science & Technology)In the TPA2 test of the CABRI-RAFT program which is part of a fast reactor safety study, fuel-to-steel heat transfer characteristics within a molten fuel/steel mixture system were investigated. This test was performed in the French CABRI reactor and used a test capsule containing fresh 12.3% enriched UO
pellets with embedded stainless steel balls. Following a pre-heating phase, the capsule was subjected to a transient overpower resulting in fuel melting and steel vaporization. The observed steel vapor pressure build-up was quite low suggesting presence of a mechanism significantly limiting the fuel-to-steel heat transfer. A detailed experimental data evaluation by SIMMER-III led to one possible interpretation that the steel vaporization at the surface of the steel ball blanketed the steel from the molten fuel.
Nakai, Satoru; Aoyama, Takafumi; Ito, Chikara; Yamamoto, Masaya; Iijima, Minoru; Nagaoki, Yoshihiro; Kobayashi, Atsuko; Onoda, Yuichi; Ohgama, Kazuya; Uwaba, Tomoyuki; et al.
Kosoku Jikkenro "Joyo" Rinkai 30-Shunen Kinen Hokokukai Oyobi Gijutsu Koenkai, 154 Pages, 2008/06
no abstracts in English
Yamano, Hidemasa; Onoda, Yuichi; Tobita, Yoshiharu; Sato, Ikkenn
JNC TN9400 2005-045, 123 Pages, 2005/06
JNC-TN9400-2005-045.pdf:18.37MB
During the course of core disruptive accidents in liquid-metal fast reactors, a boiling pool of molten fuel/steel mixture could be formed. The stability of this boiling-pool, for which in-pile experimental data with real reactor materials are very limited, plays an important role in the determination of the accident scenarios. In the TPA2 test of the CABRI-RAFT program (from 1996 to 2002), the fuel-to-steel heat transfer characteristic governing the pool behavior was investigated as a joint study with the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN). This test was performed in the CABRI reactor in 2001 using a test capsule that contains fresh 12.3% enriched UO pellets with embedded stainless steel balls. Following a pre-heating phase, the capsule was submitted to a transient overpower resulting in fuel melting and steel vaporization. The steel vapor-pressure build-up observed during the transient was quite weak, suggesting the presence of a strong mechanism to limit the fuel-to-steel heat transfer. The detailed experimental data evaluation suggested a phenomenon that the steel vaporization at the surface of steel ball blanketed the steel from molten fuel. This vapor blanketing seems to be a mechanism reducing the fuel-to-steel heat transfer. An analysis with the SIMMER-III code, a multi-component multi-phase thermal-hydraulics code, was performed in this study. This code simulation could well reproduce the pressure buildup and boiling pool behavior which occurred in the test by applying specifically reduced heat transfer coefficients.
