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Isobe, Nobuhiro*; Kawasaki, Nobuchika; Ando, Masanori; Sukekawa, Masayuki*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
Evaluation of local strain at structural discontinuities is an important technology in high temperature design of fast reactors because the failure mode in high temperature fatigue or creep fatigue damage is usually crack initiation and growth from such a locally high strained area. A rationalized strain concentration evaluation method was discussed experimentally in this study. The stress redistribution locus (SRL) method had been proposed to improve the accuracy of local stress and strain evaluation for structural discontinuities. High temperature fatigue tests of circumferentially notched specimens were conducted accompanying with local strain measurement by a capacitance type strain gage. Measured strain was compared with the prediction by the SRL method and the applicability of the method is discussed.
Isobe, Nobuhiro*; Kawasaki, Nobuchika; Ando, Masanori; Sukekawa, Masayuki*
Journal of Nuclear Science and Technology, 48(4), p.567 - 574, 2011/04
A rationalized strain concentration evaluation method was discussed experimentally in this study. High temperature fatigue tests of circumferentially notched specimens were conducted accompanying with local strain measurement by a capacitance type strain gage. Measured strain was compared with the prediction by the SRL method and the applicability of the method is discussed. The SRL method improves the accuracy of inelastic strain estimation with keeping conservativeness in comparison with the Neuber's rule which is used in high temperature design codes.
Okajima, Satoshi; Isobe, Nobuhiro*; Kawasaki, Nobuchika; Sukekawa, Masayuki*
Nihon Kikai Gakkai 2009-Nendo Nenji Taikai Koen Rombunshu, Vol.1, p.125 - 126, 2009/09
no abstracts in English
Ando, Masanori; Isobe, Nobuhiro*; Kawasaki, Nobuchika; Sukekawa, Masayuki*; Kasahara, Naoto*
Proceedings of 2009 ASME Pressure Vessels and Piping Division Conference (PVP 2009) (CD-ROM), 10 Pages, 2009/07
Wakai, Takashi; Sukekawa, Masayuki*; Date, Shingo*; Asayama, Tai; Aoto, Kazumi; Kubo, Shigenobu*
International Journal of Pressure Vessels and Piping, 85(6), p.352 - 359, 2008/06
Times Cited Count：5 Percentile：44.12(Engineering, Multidisciplinary)This paper presents the provisional material specification and characteristics of the high chromium (Cr) steel for the sodium-cooled fast reactor (SFR) in Japan and creep-fatigue assessment of the steel. Based on the mechanical test and metallurgical examination results, it is clarified that tungsten (W) should be diminished to achieve better ductility and toughness. Then the provisional specifications of the high Cr steel for SFR components are proposed. Material characteristics, e.g. cyclic stress-strain relationship and creep strain curve, are also presented based on the material test results. Using these characteristics, a creep-fatigue strength assessment of the steel was performed. Conservative predictions were obtained and it was clarified that such conservativeness were resulted from over estimation of creep damage caused by too large initial stress at the beginning of dwell. It can be pointed out there are some rooms for improvement in the assessment procedure.
Wakai, Takashi; Aoto, Kazumi; Sukekawa, Masayuki*; Date, Shingo*; Shibamoto, Hiroshi
Nuclear Engineering and Design, 238(2), p.399 - 407, 2008/02
Times Cited Count：5 Percentile：35.07(Nuclear Science & Technology)This paper presents the establishment of the provisional specifications and material strength standard of the high chromium (Cr) steels for Fast Breeder Reactor (FBR) components. For the improvement of toughness and ductility of the steels, a series of mechanical tests and metallurgical examinations are performed for several kinds of high Cr steels controlled the balance of tungsten (W) and molybdenum (Mo). In addition, the effects of heat treatment conditions on material properties are also investigated. Based on these results, it is revealed that W should be diminished to achieve better ductility and toughness and that it is difficult to improve the long term properties by changing heat treatment conditions. Then the provisional specifications of the high Cr steel for FBR components are given and the provisional material strength standard is proposed for the specifications of the steel. The standard is utilized in the study on the FBR plant design.
Isobe, Nobuhiro*; Sukekawa, Masayuki*; Nakayama, Yasunari*; Date, Shingo*; Otani, Tomomi*; Takahashi, Yukio*; Kasahara, Naoto; Shibamoto, Hiroshi*; Nagashima, Hideaki*; Inoue, Kazuhiko*
Nuclear Engineering and Design, 238(2), p.347 - 352, 2008/02
Times Cited Count：21 Percentile：78.82(Nuclear Science & Technology)The effect of ratcheting on fatigue strength was investigated in order to rationalize the strain limit as a design criterion of commercialized fast reactor systems. Ratcheting fatigue tests were conducted at 550
C. Duration of the ratchet straining was set for a certain number of strain cycles taking the loading condition of fast reactors into account, and the number of cycles for strain accumulation was defined as the ratchet-expired cycle. Fatigue lives decrease as the accumulated strain by ratcheting increases. Fatigue life reduction was negligible when the maximum mean stress was less than 25 MPa, corresponding to an accumulated strain of 2.2 percent. Accumulated strain is limited to 2 percent in the present design guidelines and this strain limit is considered effective to avoid reducing fatigue life by ratcheting. Micro-crack growth behaviors were also investigated in these tests in order to discuss the life reduction mechanisms in ratcheting conditions.
Sukekawa, Masayuki*; Isobe, Nobuhiro*; Shibamoto, Hiroshi; Tanaka, Yoshihiko*; Kasahara, Naoto
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 5 Pages, 2006/00
For expansion of non-creep design area and simplification of design procedures, a rational identification method of creep design area by negligible creep (NC) curves was studied. NC curves of six kinds of stainless and ferrite steels for fast reactors were determined at 1.5Sm (Sm: design stress intensity). These NC curves are based on domestic material data. NC curves provide the relation between temperature and time that does not induce meaningful creep strain under the constant primary stress. As for 316FR steel, which is used for reactor vessel in Japanese fast reactor, non-creep design area is identified with comparing the highest temperature and 425C (constant upper limit for austenite stainless steal) by existing Japanese Guides. However, this temperature limit can be enhanced by NC curve concept when operating (thermal transient) time is long. NC curves under higher primary stress, and the curves under secondary stress were also studied. However, at the present stage, NC curves for stress level 1.5Sm were adopted to identify creep design area. The concept of NC curve was introduced into the interim FDS (fast reactor design standard for commercialized fast reactors in Japan) to simplify the creep design of fast reactor systems. Utilizing these curves, design becomes easier for components which are employed at comparatively lower temperature under normal condition and short holding time at high temperature.
Wakai, Takashi; Aoto, Kazumi; Inoue, Kazuhiko; Sukekawa, Masayuki*; Date, Shingo*
Dai-30-Kai MPA Semina, 28 Pages, 2004/10
None
Sukekawa, Masayuki*; Imo, Kazumichi*; Kawakami, Mitsuo*
JNC TJ9430 2001-002, 98 Pages, 2002/03
JNC-TJ9430-2001-002.pdf:2.68MB
None
*; *; Sukekawa, Masayuki*; *
PNC TJ9124 88-001, 103 Pages, 1988/04
Strength tests of SUS304 stainless steel welded joint and Mod.9Cr-1Mo forging base metal at elevated temperature were carried out for the purpose of getting data needed for prototype Fast Breeder Reactor and Demonstration plant. Same as last year's, examined materials were welded joints of SUS304 plate (40t) and SUS304 forging (350t). For base metal test, Mod.9Cr-1Mo forging (550t) were applied. Creep test, bending creep and fatigue test of welded joint of SUS304 plate and SUS304 forging, and creep test of weld metal were carried out. Tensile test, creep test and fatigue test of Mod.9Cr-1Mo forging were carried out. Results of these tests are as follows. (1)308 weld metal and welded joint of 304 plate and forging show that their creep strength are higher than those of Material Strength Standard, PNC. (2)Bending creep fatigue strength of welded joint of plate and forging are a little higher than those strength on axial load at 500C and are a considerable higher than those at 500C. (3)Tensile strength, creep strength and fatigue strength of Mod.9Cr-1Mo forging steel are same as Material Strength Standard, PNC (preliminary). And there are no significant difference between two data of surface and middle of forging.
Fujimata, Kazuhiro*; Nagashima, Hideaki*; Sukekawa, Masayuki*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto
no journal, ,
no abstracts in English
