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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
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
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
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
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.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; ; Miyakawa, Akira; Okabe, Ayao;
JNC TN9400 2001-130, 235 Pages, 2002/03
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. Major results obtained in the studies are as follows: (1)To evaluate the structural integrity of tube material, the strength standard for 2.25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200C) creep data. This standard has been validated with the tube rupture simulation test data. (2)The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. (3)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. (4)The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. (5)The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system.
Wada, Yusaku
Kazafusutan Genshiryoku Senta Soritsu 10-Shunen Kinen Kokusai Kaigi, 0 Pages, 2002/00
None
; ; Yamaguchi, Akira;
JNC TN9400 2000-088, 67 Pages, 2000/09
The 1995 sodium leak incident of the secondary main piping of the prototype fast breeder reactor MONJU stemmed from the failure of a thermometer weU, which was due to flow-induced vibration. JNC, Japan Nuclear Fuel Cycle Development lnstitute, formerly PNC, Power Reactor and Nuclear Fuel Development Corporation, prepared the Flow-induced ibration Design Guide for Thermometer Wells (FIV-DG) in 1997, to prevent the recurrence of the similar phenomena. This document provides a detailed background for the FIV-DG, covering the basic design philosophy against flow-induced vibration, the explanation of each provision, the design methods and procedures employed, and supporting experimental data. Recent experimental results and findings, which were obtained after the establishment of the FIV-DG, are utilized to enrich this background document.
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.
Hamada, Hirotsugu; *; *; *; Hiroi, Hiroshi*
PNC TN9410 98-029, 122 Pages, 1998/05
The following items have been studies to evaluate overheating failure of FBR generator heat transfer tubes: (1)To establish a structural integrity analysis method. The strength standard values for 2.25Cr-1Mo steel was established taking account of time dependent effect to overheating failure mechanism based on high temperature (700 - 1200C) creep data and was validated by tube rupture simulation test data. (2)To improve and validate blow down analytical method. The analytical result by use of BLOOPH, the FBR blow down code, was compared with that by use of RELAP-5, the general purpose thermo-hydraulic code, and a good agreement was obtained. (3)To quantitatively validate the entire overheating analysis model by sodium water reaction data Sodium-water reaction tests of SWAT-3 and LLTR were analyzed using above mentioned analytical method. The ductile fracture occurred earlier than the creep fracture in the analysis and the comparison of tube failure times with the experiments showed sufficient conservativeness. Based on the above studies, the analytical method was applied to PFR superheater leak event and the Monju steam generator accidental analysis. The followings were quantatitively shown through the analysis: (1)The most important cause that multi-tube failure occurred in the 1987 PFR superheater-2 leak is that the superheater did not equip a fast steam dump system at the time of the leak event. (2)Overheating failure will not occur under any operational conditions of Monju in both steady state and transient phases such as water/steam blow-down. (3)Although safety margin becomes small when the water/steam flow rate becomes small during the blow-down, the modification of the plant such as hastening blow-down by equipping more relief valves will drastically improve the safety margin.
Iwata, Koji; *; *; *; *; *; *
PNC TN9410 97-042, 8 Pages, 1997/03
A design guide for flow-induced vibration of thermometer wells is proposed to prevent the recurrence of the failure of thermometer wells, which was the direct cause of the 1995 sodium leak incident of the secondary main piping of the prototype fast breeder reactor MONJU. As a supplement to the technical standards in force for MONJU, the design guide specifies the methods of evaluation and the design criteria on structural integrity against flow-induced vibration for thermometer wells, which are inserted into pipes of fast breeder reactors. The design guide is a PNC's (Power Reactor and Nuclear Fuel Development Corporation) internal guide for MONJU, which is to be used, with the permission of outside authorities, to confirm the integrity of the existing equipments as well as to make an improved design of thermometer wells. The proposed design guide was prepared by the Special Working Group on Thermometer Design Guide, organized in PNC during the period from May to November, 1996.