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Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00393_1 - 16-00393_10, 2017/04
The achievement of In-Vessel Retention (IVR) of the accident consequences in an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, is effective and rational approach in enhancing the safety characteristics of sodium-cooled fast reactor. The objective of the present study is to show that the decay heat generated from the relocated fuels would be stably removed in post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phase, where the relocated fuels mean fuel discharged from the core into the low-pressure plenum through control-rod guide tubes, and fuel remnant in the disrupted core region (non-discharged fuel). As a result of the present assessments, it should be concluded that the stable cooling of the relocated fuels was confirmed and the prospect of IVR was obtained.
Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 3 Pages, 2016/06
no abstracts in English
Wada, Yusaku
Heisei-21-Nendo Nihon Kinzoku Gakkai Kanto Shibu Koshukai "Hakai No Genin O Hamen Kansatsu Kara Saguru" Tekisuto, p.6_1 - 6_6, 2009/09
In December 1995, a thermocouple well was broken and liquid sodium leaked out of the secondary heat transport system of Monju, which was operated in 40% power for the general plant performance tests. The sodium leakage was caused by the breakage of a thermowell that was installed on the main pipe. The thermowell had been suffered from flow induced vibration leading to high cycle fatigue failure. This vibration was an in-line oscillation associated with symmetric vortex shedding. The evidence of failure cause analysis was based on the fractographic examination, and microstructure of fracture surface showed the features of high cycle fatigue. Furthermore crack arrests were observed. The well was broken at the neck where a stress concentration was large by the geometric discontinuity of diameter transition. Crack initiation and growth analyses were carried out considering the deterioration of natural frequency of well with crack depth increase.
Wada, Yusaku; Okubo, Toshiyuki; Miyazaki, Hitoshi; none; Donomae, Yasushi
JNC TN9410 2005-007, 94 Pages, 2005/03
JNC-TN9410-2005-007.pdf:6.07MB
None
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi
JNC TN2400 2003-003, 225 Pages, 2004/02
JNC-TN2400-2003-003.pdf:40.45MB
The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies.
Misawa, Naoto; Wada, Yusaku; Kato, Shoichi*
JNC TN2400 2003-001, 38 Pages, 2004/01
JNC-TN2400-2003-001.pdf:1.08MB
When rolled steels for welded structure SM400B was used under temperature change environment, we can apply the method of soundness evaluation of heat strain generated into this material conservatively and appropriately, by the ductility exhaution in two stages in temperature falling process, or by the strain criterion in two stages in temperature rising process.
Saito, Naoshi*; Tsuruda, Takashi*; Wada, Yusaku
JNC TY9200 2004-002, 61 Pages, 2003/03
JNC-TY9200-2004-002.pdf:2.95MB
On the research related to combustion behavior in coolant sodium leak in the fast reactor, it is important to phenomenologically clarify the behavior in addition to conventional engineering challenge. National Research Institute of Fire and Disaster(NRIFD) and Japan Nuclear Cycle Development Institute(JNC) have done cooperative research since the 1998 fiscal year for the purpose of deepening understanding on the sodium combustion behavior by the information exchange on basic experiment and analysis of sodium combustion behavior carried out in each institute. This report coordinated results such as information exchange conference in the 2001 - 2002 fiscal year.
Wada, Yusaku
Kazafusutan Genshiryoku Senta Soritsu 10-Shunen Kinen Kokusai Kaigi, 0 Pages, 2002/00
None
Komine, Ryuji; Wada, Yusaku
PNC TN9410 98-086, 135 Pages, 1998/08
A sodium-water reaction drived from the single tube break in steam generator might overheat nabor tubes rapidly under internal pressure loadings. If the temperature of tube wall becomes too high, it has to be evaluated that the stress of tube does not exceed the material strength limit to prevent the propagation of tube rupture. In the present study this phenomenon was recognized as the fracture of cylindrical tube with the large deformation due to overheating, and the evaluation method was investigated based on both of experimental and analytical approaches. The results obtained are as follows. (1)As for the nominal stress estimation, it was clarified through the experimental data and the detailed FEM elasto-plastic large deformation analysis that the formula used in conventional designs can be applied. (2)Within the overheating temperature limits of tubes, the creep effect is dominant, even if the loading time is too short. So the strain rate on the basis of JIS elevated temperature tensile test method for steels and heat-resisting alloys is too late and almost of total strain is composed by creep one. As a result the time dependent effect cannot be evaluated under JIS strain rate condition. (3)Creep tests in shorter time condition than a few minutes and tensile tests in higher strain rate condition than 10%/min of JIS are carried out for 2.25Cr-1Mo(NT) steel, and the standard values for tube rupture strength evaluation are formulated. (4)The above evaluation method based on both of the stress estimation and the strength standard values application is justified by using the tube burst test data under internal pressure. (5)The strength standard values on Type 321 ss is formulated in accordance with the procedure applied for 2.25Cr-1Mo(NT) steel.
Aoto, Kazumi; Wada, Yusaku
IAEA-TECDOC-817, p.79 - 87, 1995/00
None
Aoto, Kazumi; Abe, Yasuhiro; Shibahara, Itaru; Wada, Yusaku
IAEA-TECDOC-817, p.27 - 37, 1995/00
None
1MoAoto, Kazumi; ; Ueno, Fumiyoshi; Kawasaki, Hirotsugu; Wada, Yusaku
Nuclear Engineering and Design, 153(1), p.97 - 110, 1994/12
Times Cited Count：56 Percentile：96.55(Nuclear Science & Technology)None
Asayama, Tai; Hasebe, Shinichi; Wada, Yusaku
PNC TN9410 94-307, 43 Pages, 1994/11
None
Aoki, Norichika*; Yoshida, Eiichi; Wada, Yusaku
Nihon Kinzoku Gakkai-Shi, 58(2), p.124 - 131, 1994/00
None
Watashi, Katsumi; Aoto, Kazumi; Aoki, M; Komine, Ryuji; Ito, Takushi; Hasebe, Shinichi; Kato, Shoichi; Koi, Mamoru; Wada, Yusaku
PNC TN9410 93-142, 120 Pages, 1993/06
Much progress has been made in improving established creep properties of Type 316 stainless steel and to develop a new structural material named "FBR Grade Type 316 Stainless Steel", 316FR, with superior creep properties. This report includes a draft of Material Strength Standard of 316FR and its interpretation on the basis of the major result of research and development conducted so far. The draft includes identical items described in the "Standards for the Strength of Materials" for Monju, and was carefully prepared to have an identical style for convenience in design evaluation. Creep damage evaluation diagrams, which are depicted in the "Structural Design Guide for Class 1 Components of Prototype Fast Breeder Reactor for Elevated Temperature Service" (ETSDG) for individual materials, are also included in this report.
