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Ota, Hirokazu*; Ogata, Takanari*; Yamano, Hidemasa; Futagami, Satoshi; Shimada, Sadae*; Yamada, Yumi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12
The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.
Ogura, Koya*; Hosoda, Masahiro*; Tamakuma, Yuki*; Suzuki, Takahito*; Yamada, Ryohei; Negemi, Ryoju*; Tsujiguchi, Takakiyo*; Yamaguchi, Masaru*; Shiroma, Yoshitaka*; Iwaoka, Kazuki*; et al.
International Journal of Environmental Research and Public Health, 18(3), p.978_1 - 978_16, 2021/02
Times Cited Count：7 Percentile：68.83(Environmental Sciences)Kubota, Ryuzaburo; Koyama, Kazuya*; Moriwaki, Hiroyuki*; Yamada, Yumi*; Shimakawa, Yoshio*; Suzuki, Toru; Kawada, Kenichi; Kubo, Shigenobu; Yamano, Hidemasa
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
This paper describes an analysis study on the initiating phase of the ATWS events with SAS4A in order to confirm the appropriateness of the core design for the medium-scale SFR (750MWe-1765MWt). Not using a conventional lumping method that multiple fuel sub-assemblies having a similar characteristic were assigned to one channel (representing fuel assembly in SAS4A), each channel represents only the sub-assemblies of identical operating condition. In addition, the detailed power and reactivity distribution were set reflecting the change of insertion position of control rods. Applying these detailed analysis conditions, the SAS4A analyses were performed for unprotected loss-of-flow (ULOF) and unprotected transient overpower (UTOP) during both of the nominal power and the partial power operation. As a result, more proper event progression including incoherency of events especially fuel dispersion after fuel failure was successfully evaluated and then this analysis study suggested that the power excursion with prompt criticality leading to large mechanical energy release can be prevented in the initiating phase of the current design.
Saito, Hiroyuki*; Yamada, Yumi*; Oyama, Kazuhiro*; Matsunaga, Shoko*; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
A self-actuated shutdown system (SASS) is a passive device, which can detach a control rod for reactor shutdown in response to excessive increase in coolant temperature. Since a detachment temperature, which triggers release of a control rod, and a response time are identified as important parameters for validity analyses, this study focused on investigation of the response time and the detachment temperature, and safety analysis to see feasibility of the SASS in low power. For this purpose, design modifications were made to shorten the response time and three-dimensional thermal-hydraulic analysis in a low power operation was carried out in order to confirm the response time. The resulting detachment temperature level is lower than previous studies, leading to improved safety parameters. Based on improved parameters, a safety analysis to see feasibility of the SASS in low power was carried out. From this safety evaluation, it was confirmed that core damage can be prevented by the SASS with flow collector in the case of LOF type ATWS event.
evaluation of F(
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I)KCCYSL for targeting HER2Sasaki, Ichiro; Watanabe, Shigeki; Ohshima, Yasuhiro; Sugo, Yumi; Yamada, Keiichi*; Hanaoka, Hirofumi*; Ishioka, Noriko
Peptide Science 2015, p.243 - 246, 2016/03
Sasaki, Ichiro; Hanaoka, Hirofumi*; Yamada, Keiichi*; Watanabe, Shigeki; Sugo, Yumi; Ohshima, Yasuhiro; Suzuki, Hiroyuki; Ishioka, Noriko
Peptide Science 2014, p.257 - 260, 2015/03
Sasaki, Ichiro; Yamada, Keiichi*; Watanabe, Shigeki; Hanaoka, Hirofumi*; Sugo, Yumi; Oku, Hiroyuki*; Ishioka, Noriko
JAEA-Review 2013-059, JAEA Takasaki Annual Report 2012, P. 96, 2014/03
Sasaki, Ichiro; Yamada, Keiichi*; Watanabe, Shigeki; Hanaoka, Hirofumi*; Sugo, Yumi; Oku, Hiroyuki*; Ishioka, Noriko
Peptide Science 2013, p.157 - 160, 2013/03
Hagiwara, Hiroyuki; Yamada, Yumi*; Eto, Masao*; Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
The self-actuated shutdown system (SASS), which is selected for Japan Sodium-cooled Fast Reactor (JSFR), is a passive reactor shutdown system utilizing a Curie point electromagnet (CPEM). With CPEM, an excessive fuel outlet temperature rise is sensed and the control rods are released into the core, and the reactor can be shutdown. Therefore it is important for feasibility of SASS to be established by assuring a quick response of CPEM to the coolant temperature rise. In this paper, a device named "flow collector", which collects flows discharged from six fuel subassemblies surrounding CPEM backup control rods, has been proposed to ensure a shorter response time.
Kubota, Ryuzaburo*; Yamada, Yumi*; Koyama, Kazuya*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kubo, Shigenobu; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
This paper describes a melt-down event progression revealed by a numerical simulation in the protected loss of heat sink (PLOHS) event for Japan Sodium-cooled Fast Reactor (JSFR). A multi-component multi-field computer code, MUTRAN, has been applied in order to simulate complicated core material motions and associated heat-transfer phenomena among the materials in a degradation core. The analyses with MUTRAN covered core degradation behaviors from the intact geometry and addressed the two initial states: one was the core without the coolant as the leakage type, and the other was the core covered by the coolant only up to the top of the fissile fuel as the boiling type. The analyses revealed representative event progression.
Fujita, Kaoru; Yamano, Hidemasa; Kubo, Shigenobu*; Eto, Masao*; Yamada, Yumi*; Toyoshi, Akira*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 6 Pages, 2012/12
no abstracts in English
Ishikawa, Nobuyuki; Chikazawa, Yoshitaka; Fujita, Kaoru; Yamada, Yumi*; Okazaki, Hitoshi*; Suzuki, Shinichi*
Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.483 - 489, 2012/06
The development of safety protection system for JSFR is progressed in terms of logic circuits, selection of trip signals and its setting values for reactor trip. In addition, it is necessitated to evaluate the satisfaction for requirements of the safety protection system by safety analyses considering comprehensive parameter ranges. For this purpose, we will report the current status of the development focusing on the evaluation results for satisfaction of safety protection system based on safety standard.
Nagaishi, Ryuji; Yamada, Reiji; Kumagai, Yuta; Sugo, Yumi
JAEA-Review 2008-055, JAEA Takasaki Annual Report 2007, P. 160, 2008/11
no abstracts in English
Nagaishi, Ryuji; Yamada, Reiji; Aoyagi, Noboru; Sugo, Yumi
JAEA-Review 2007-060, JAEA Takasaki Annual Report 2006, P. 161, 2008/03
From the standpoints of utilization of radioactive wastes, and of sophistication of separation process of spent fuels, we have been investigating promotion or inhibition of radiation-induced reactions in immiscible heterogeneous systems: solutions coexisting/contacting with solid oxides, solvent system with aqueous and organic phases, etc.. We have recently report that the reactions of reduction of metal ions and of hydrogen production in aqueous solution were promoted by adding oxide particles to the solution, and that the radiolysis of amides in n-dodecane was dependent on aqueous solution contacting with the n-dodecane. In this report, we illustrate recover of platinum-group elements from aqueous solution, and non-toxic treatment of chrysotile asbestos using ionizing radiations as the experimental results found in fiscal 2006.
*; Kishida, Masako*; *
JNC TJ9440 2000-013, 80 Pages, 2000/03
JNC-TJ9440-2000-013.pdf:4.93MB
In this work, analyses on sodium-water reaction temperaturc in the new SWAT-1(SWAT-1R) test were completed by the CHAMPAGNE code in order to understand void and velocity distribution in sodium system, which was difficult to be measured in experiments. The application method of the RELAP5/Mod2 code was investigated to LMFBR steam generator(SG)blow down analysis, too. The following results were obtained. (1)Analyses on sodium-water reaction temperature in the SWAT-1R test. (a)Analyses were carried out for the SWAT-1R test under the condition water leak rate 600 g/s by treating thc pressure loss coefficient, the interface friction coefficient and the coefficient related to reaction rate as parameters. The effect and mechanism of each parameter on the shape of rcaction zone were well understood by these analyses. (b)The void and velocity distribution in sodium system were estimated by use of the most suitable parameters. These analytical results are expected to be useful for planning of the SWAT-1R test and evaluation of test result. (2)Investgation of the RELAP5/Mod2 code. (a)The items to be improved in the RELAP5/Mod2 code were clarified to apply this code to the FBR SG blow down analysis. (b)One of these items was an addition of the shell-side (sodium-side) model. A sodium-side model was designed and added to the RELAP5/Mod2 code. Test calculations were carried out by this improved code and the basic function of this code was confirmed.
Moribe, Takeshi; Sakai, Takaaki; Ikeda, Hirotsugu*; Yamada, Yumi*; Kurome, Kazuya*
no journal, ,
no abstracts in English
Niwa, Hajime; Kurisaka, Kenichi; Sato, Ikken; Tobita, Yoshiharu; Kamiyama, Kenji; Yamano, Hidemasa; Miyahara, Shinya; Ohno, Shuji; Seino, Hiroshi; Ishikawa, Hiroyasu; et al.
no journal, ,
In order to develop the core damage evaluation technology (level 2 PSA) for sodium-cooled fast reactors, we develop the new analysis codes of post accident material relocation phase and of ex-vessel events, and we develop the technical bases that is necessary for level 2 PSA. In this presentation, summary and scope of the entire study is introduced as a part of the 4 series presentations.
Nagaishi, Ryuji; Yamada, Reiji; Aoyagi, Noboru; Sugo, Yumi
no journal, ,
no abstracts in English
Nakai, Ryodai; Kurisaka, Kenichi; Sato, Ikken; Tobita, Yoshiharu; Kamiyama, Kenji; Yamano, Hidemasa; Miyahara, Shinya; Ohno, Shuji; Seino, Hiroshi; Ishikawa, Hiroyasu; et al.
no journal, ,
To develop a core damage evaluation technology (level-2 PSA) in sodium-cooled fast reactors, a new analysis method is developed for core-material relocation phase and internal containment vessel event. This study also develop technical basis necessary for the level-2 PSA.
