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Yano, Yasuhide; Uwaba, Tomoyuki; Tanno, Takashi; Yoshitake, Tsunemitsu; Otsuka, Satoshi; Kaito, Takeji
Journal of Nuclear Science and Technology, 61(4), p.521 - 529, 2024/04
Times Cited Count：3 Percentile：55.74(Nuclear Science & Technology)The effects of fast neutron irradiation on tensile properties of modified 316 stainless steel (PNC316) claddings and wrappers for fast reactors were investigated. PNC316 claddings and wrappers were irradiated in the experimental fast reactor Joyo at irradiation temperatures between 400 and 735 
C to fast neutron doses ranging from 21 to 125 dpa. The post-irradiation tensile tests were carried out at room and irradiation temperatures. Elongations of PNC316 measured by the tensile tests were maintained at an engineering level, although the material incurred significant irradiation hardening and softening. The maximum swelling of PNC316 wrappers was about 2.5 vol.% at irradiation temperature between 400 and 500C up to 110 dpa. Japanese 20% cold-worked austenitic steels, PNC316 and 15Cr-20Ni, had sufficient ductility and work-hardenability even after above 10 vol.% swelling, while they had very weak plastic instabilities.
Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Kaito, Takeji
Journal of Nuclear Materials, 555, p.153105_1 - 153105_8, 2021/11
Times Cited Count：4 Percentile：29.36(Materials Science, Multidisciplinary)The aim of this study was to evaluate the tensile properties and microstructures of dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging at temperatures between 400 and 600C up to 30,000 h. Characterization of microstructure was carried out by scanning electron microscopy and transmission electron microscopy. Microstructural analysis showed that the microstructure in the weld metals consisted of lath martensite containing a small amount of residual austenite. Thermal aging hardening of WMs occurred at 400 and 450C due to the effects of both a-a' phase separation and G-phase precipitation. However, there was no significant change in the total elongation, and fracture surfaces indicated that very fine dimpled rupture was predominant rather than the cleavage rupture. It was suggested that lath martensite phases enhanced the tensile strength due to phase separation, while residual austenite played a role in keeping elongation as a soft phase.
Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Kaito, Takeji; Ukai, Shigeharu*
Materials Transactions, 62(8), p.1239 - 1246, 2021/08
Times Cited Count：11 Percentile：51.45(Materials Science, Multidisciplinary)The FeCrAl-ODS alloy claddings were manufactured and Vickers hardness, ring tensile tests and TEM observations of these claddings were performed to investigate the effects of thermal aging at 450 C for 5,000 and 15,000 h. The age-hardening of all FeCrAl-ODS alloy cladding was found. In addition, the significant increase in tensile strength was accompanied by much larger loss of ductility. It was suggested that this age-hardening behavior was attributed to the (Ti, Al)-enriched phase (
' phase) and the 
' phase precipitates (content of Al is 
7 wt%). In comparison with FeCrAl-ODS alloys with almost same chemical compositions, there was significant age-hardening in both alloys. However, the extrusion bar with no-recrystallized structures was keeping good ductility. It was suggested that this different behavior of reduction ductility was attributed to the effects of grain boundaries, dislocation densities and specimen preparation direction.
Fukuya, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro; Hata, Kuniki
Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), p.523 - 531, 2019/08
For structural integrity assessment on reactor internals of light water reactors, empirical equations of tensile properties as a function of neutron dose, and trend curves of stress-strain relations of neutron-irradiated austenitic stainless steels was proposed by fitting to recently developed database. The data in the database were obtained from reports of national projects in Japan and open literature, which was summarized in the form of data sheets. The empirical equations for tensile properties were formulated by using a saturation-type formulae. The equations were for CW 316 and SA 304/316 stainless steels in the temperature range of 280-350C and the dose range up to 80 dpa. Stress-strain relation curves were reproduced based on the Swift model. Obtained calculated results by the empirical equations and stress-strain relations were reasonably well fitted to experimental data. The effects of composition and cold-working, etc. on tensile properties were discussed.
Saito, Toru; Okubo, Toshikazu*; Izumi, Keisuke*; Okawa, Yoshinao*; Kobayashi, Norihiro*; Yamazaki, Toru; Kawano, Katsumi; Isono, Takaaki
Teion Kogaku, 50(8), p.400 - 408, 2015/08
Aramid fiber-reinforced plastic (AFRP) has been developed as a structural material that has the advantages of light weight and high strength. In this study, tensile tests were carried out to measure the tensile properties of AFRP rod on the market for reinforcement of concrete at room temperature, 77 K and 4.2 K. Especially at cryogenic temperatures, it is difficult to perform a tensile test of the bar because the specimen slips through the jig grip. To prevent the rod from slipping, tensile tests were carried out with some filling conditions. The applicable and appropriate tensile test conditions were established by modifying the jig grip, treating the surface of the rod and using cryogenic epoxy infill to grip the rod. They were more than 1100 MPa. Additionally, the AFRP rod included a temperature dependence in which the Young's modulus increased as the test temperature decreased. It was confirmed that the Young's modulus increased because aramid fiber was more dominant than epoxy.
Shiba, Kiyoyuki; Klueh, R. L.*; Miwa, Yukio; Igawa, Naoki; Robertson, J. P.*
Fusion Materials Semiannual Progress Report (DOE/ER-0313/28), p.131 - 135, 2000/06
no abstracts in English
; ; ; ; Aoto, Kazumi;
JNC TN9400 2000-059, 43 Pages, 2000/05
JNC-TN9400-2000-059.pdf:2.08MB
The purpose of this study is to understand the material properties of manufacturable high-purity iron and high-purity iron-based alloy in present technology and to get an applicable prospect for the structural and functional material of the frontier fast reactor. Then the about 10kg high-purity iron and iron-based alloy were melted using a cold-crucible induction melting furnace under the ultra-high vacuum. Subsequent to that, the compatibility between the melted material and the high-temperature sodium environment which is a special feature of the fast reactor and tensile property at room and elevated temperatures were investigated using the melted materials. Also, the creep test using the high-purity 50%Cr-Fe alloy at 550C in air in order to understand the high temperature creep property. ln addition, the material properties such as thermal expansion coefficient, specific heat and electrical resistance were measured and to evaluate the outlook for the structural material for the fast reactor. The following results were obtained based on the property test and evaluation. (1)lt was possible to melt the about 10kg high-purity ingot and high-purity 50%Cr-Fe alloy ingot using a cold-crucible induction melting furnace under the ultra-high vacuum. (2)The tensile tests of the high-purity 50%Cr-Fe alloy were performed at room and elevated temperatures in order to understand the deformation behavior. From the experimental results, it was clear that the high-purity 50%Cr-Fe alloy possesses high strength and good ductility at elevated temperatures. (3)The physical properties (the thermal expansion coefficient and specific heat etc.) were measured using the high-purity 50%Cr-Fe alloy. lt was clear that the thermal expansion coefficient of high-purity 50%Cr-Fe alloy was smaller than that of SUS304. (4)From the corrosion test in liquid sodium, the ordinary-purity iron showed the weight loss after corrosion test. However the high-purity iron showed ...
Hishinuma, Akimichi; Tabuchi, Masayuki; Sawai, Tomotsugu
Intermetallics, 7(8), p.875 - 879, 1999/00
Times Cited Count：11 Percentile：60.77(Chemistry, Physical)no abstracts in English
Ioka, Ikuo; Naito, Akira; Shiba, Kiyoyuki; Jitsukawa, Shiro; J.P.Robertson*; Hishinuma, Akimichi
Journal of Nuclear Materials, 258-263, p.1664 - 1668, 1998/00
Times Cited Count：2 Percentile：24.08(Materials Science, Multidisciplinary)no abstracts in English
Hishinuma, Akimichi; Jitsukawa, Shiro
Annales de Physique, 22(SUPPL.3), p.163 - 170, 1997/06
no abstracts in English
Watanabe, Katsutoshi; Shindo, Masami; Nakajima, Hajime; Koikegami, Hajime*; Higuchi, Makoto*; Nakanishi, Tsuneo*; Sahira, Kensho*; Marushichi, Koki*; Takeiri, Toshiki*; Saito, Teiichiro*; et al.
JAERI-Research 97-009, 62 Pages, 1997/02
JAERI-Research-97-009.pdf:4.82MB
no abstracts in English
Hishinuma, Akimichi
Journal of Nuclear Materials, 239(1-3), p.267 - 272, 1996/00
Times Cited Count：21 Percentile：82.33(Materials Science, Multidisciplinary)no abstracts in English
; Watanabe, Katsutoshi; Tsuji, Hirokazu; Nakajima, Hajime
JAERI-M 93-148, 34 Pages, 1993/07
no abstracts in English
Fujita, Mitsutane*; ; Nakajima, Hajime; ; Ueno, Fumiyoshi*; Kano, Shigeki*; Iwata, Shuichi*
Computer Aided Innovation of New Materials,II,Pt. 1, p.81 - 84, 1993/00
no abstracts in English
Tsuji, Hirokazu; ; Tsukada, Takashi; Nakajima, Hajime
Proc. of the 4th Int. Symp. on Advanced Nuclear Energy Research (JAERI-CONF 1/JAERI-M 92-207), p.426 - 433, 1992/12
no abstracts in English
Kurata, Yuji; Nakajima, Hajime
JAERI-M 90-157, 38 Pages, 1990/09
no abstracts in English
; ; Ueda, Shuzo
Bull.JSME, 20(141), p.278 - 284, 1977/00
no abstracts in English
Fukuya, Koji*; Chimi, Yasuhiro; Kasahara, Shigeki; Fujii, Katsuhiko*; Fujimoto, Koji*
no journal, ,
no abstracts in English
Fukuya, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro
no journal, ,
To evaluate the effects of neutron irradiation on the tensile properties of austenitic stainless steels for reactor core internals in light water reactors, we have revised the database on tensile properties of irradiated stainless steels and investigated the dose dependence. Based on the fundamental equation, which can express the tendency toward saturation of tensile properties with increasing the dose, we have proposed the appropriate trend equations of dose dependence for the classified data in terms of irradiation conditions, materials, work and thermal treatment conditions, etc. In the presentation, we will report the results of investigation on the effects of each condition on the irradiation behavior of tensile properties.
Wakui, Takashi; Saito, Shigeru; Wakai, Eiichi; Sakai, Tomoki*; Mori, Kotaro*; Futakawa, Masatoshi
no journal, ,
One of dominant factors to determine the lifetime of the mercury target in J-PARC is the radiation damage. Authors suggested the tensile properties evaluation technique from numerical tensile tests using material properties estimated from inverse analyses on indentation tests. The technique was applied to ion-irradiated materials, and the validity of the technique was investigated by comparing the result with results of the PIE on the targets of SNS. By conducting indentation tests on samples cut out from used targets, it is expected that the residual lifetime estimation can be conducted considering various effects; fatigue, temperature, LME, etc. superimposed on the radiation damage from evaluating hardness and tensile properties obtained by the technique. The technique and comparison results will be discussed.
