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Journal Articles

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

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.

Journal Articles

In-vessel retention of unprotected accident in a fast reactor; Assessment of material-relocation and heat-removal behavior in ULOF

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

Journal Articles

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

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

Journal Articles

Practical experiences of incident cause study by fractography analysis, 4; Cause analysis on Monju's sodium leak incident

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.

JAEA Reports

None

Wada, Yusaku; none; Miyazaki, Hitoshi; none; Donomae, Yasushi

JNC TN9410 2005-007, 94 Pages, 2005/03

JNC-TN9410-2005-007.pdf:6.07MB

None

JAEA Reports

The Development and Application of Overheating Failure Model of FBR Steam Generator Tubes (IV)

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.

JAEA Reports

Evaluation method of the thermal strain for the steel of SM400B

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.

JAEA Reports

Study on sodium combustion behavior(III); NRIFD/JNC Joint Research Project

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.

Journal Articles

Status on R&D Activities for Key Technologies of FR Systems in OEC/JNC, Japan

Wada, Yusaku

Kazafusutan Genshiryoku Senta Soritsu 10-Shunen Kinen Kokusai Kaigi, 0 Pages, 2002/00

None

Journal Articles

None

Koi, Mamoru; Wada, Yusaku

Oarai FBR Saikuru Shimpojiumu, 0 Pages, 2001/00

None

JAEA Reports

Study on tube rupture strength evaluation method for rapid overheating

Komine, Ryuji; Wada, Yusaku

PNC TN9410 98-086, 135 Pages, 1998/08

PNC-TN9410-98-086.pdf:8.3MB

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.

Journal Articles

Concept of design criteria of low dose irradiation for FBR structural materials

Aoto, Kazumi; Wada, Yusaku

IAEA-TECDOC-817, p.79 - 87, 1995/00

None

Journal Articles

Effects of neutron irradiation on creep properties of FBR grade 316 stainless steel

Aoto, Kazumi; Abe, Yasuhiro; Shibahara, Itaru; Wada, Yusaku

IAEA-TECDOC-817, p.27 - 37, 1995/00

None

Journal Articles

None

Aoto, Kazumi; Wada, Yusaku

Zairyo, 44(496), p.23 - 28, 1995/00

None

Journal Articles

None

Wada, Yusaku; Aoto, Kazumi; Ueno, Fumiyoshi

Zairyo, 44(496), p.29 - 34, 1995/00

None

Journal Articles

Creep-fatigue evaluation of normalized and tempered modified 9Cr$$cdot$$1Mo

Aoto, Kazumi; ; Ueno, Fumiyoshi; ; Wada, Yusaku

Nuclear Engineering and Design, 153(1), p.97 - 110, 1994/12

 Times Cited Count:53 Percentile:96.45(Nuclear Science & Technology)

None

Journal Articles

Effect of Cold Work on Decarburization of 2%Cr・1Mo Steel in High Temperature Sodium

Aoki, Norichika*; Yoshida, Eiichi; Wada, Yusaku

Nihon Kinzoku Gakkai-Shi, 58(2), p.124 - 131, 1994/00

None

Journal Articles

None

Wada, Yusaku; Aoto, Kazumi; Ueno, Fumiyoshi

Donen Giho, (87), p.19 - 33, 1993/09

None

JAEA Reports

Material strength standard of FBR grade type 316 stainless steel(Draft)

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

PNC-TN9410-93-142.pdf:6.08MB

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.

55 (Records 1-20 displayed on this page)