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Yano, Yasuhide; Uwaba, Tomoyuki; Tanno, Takashi; Yoshitake, Tsunemitsu; Otsuka, Satoshi; Kaito, Takeji
Journal of Nuclear Science and Technology, 9 Pages, 2023/00
Times Cited Count：1 Percentile：72.91(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.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Tanaka, Kosuke; Suto, Mitsuo; Onishi, Takashi; Akutsu, Yoko; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Sekioka, Ken*; Ishigamori, Toshio*; Obayashi, Hiroshi; Koyama, Shinichi
JAEA-Research 2013-036, 31 Pages, 2013/12
JAEA-Research-2013-036.pdf:3.31MB
In the accident of Fukushima Daiichi NPPs, the water ingress was performed in order to decrease the reactor temperature. At that time, sea water was temporarily used as a coolant and the water contacted with nuclear fuel directly. It can be supposed that fission products (FP) were easily migrated from the fuel to sea water in this situation and that affect the water quality. The knowledge of leaching behavior, therefore, is necessary for evaluating the integrity of reactor component materials such as steels for pressure containment vessel and for reactor vessel. In order to obtain the fundamental knowledge for leaching behavior of FP in the hot sea water, the leaching tests of irradiated fuel were performed and the leachates were subjected to chemical analysis. It is found that he leaching rate of each nuclides obtained in this study were similar to that of the leaching results simulating the underground water.
Yano, Yasuhide; Otsuka, Satoshi; Yamashita, Shinichiro; Ogawa, Ryuichiro; Sekine, Manabu; Endo, Toshiaki; Yamagata, Ichiro; Sekio, Yoshihiro; Tanno, Takashi; Uwaba, Tomoyuki; et al.
JAEA-Research 2013-030, 57 Pages, 2013/11
JAEA-Research-2013-030.pdf:48.2MB
It is necessary to develop the fast reactor core materials, which can achieve high-burnup operation improving safety and economical performance. Ferritic steels are expected to be good candidate core materials to achieve this objective because of their excellent void swelling resistance. Therefore, oxide dispersion strengthened (ODS) ferritic steel and 11Cr-ferritic/martensitic steel (PNC-FMS) have been respectively developed for cladding and wrapper tube materials in Japan Atomic Energy Agency. In this study, the effects of fast neutron irradiation on mechanical properties and microstructure of 9Cr-and 12Cr-ODS steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the CMIR-6 at temperatures between 420 and 835C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.
Yano, Yasuhide; Ogawa, Ryuichiro; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Inoue, Masaki; Yoshitake, Tsunemitsu; Tanaka, Kenya
Journal of Nuclear Materials, 419(1-3), p.305 - 309, 2011/12
Times Cited Count：20 Percentile：80.18(Materials Science, Multidisciplinary)The effects of fast neutron irradiation on ring tensile properties of oxide dispersion strengthened (ODS) steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the material irradiation rig at temperatures between 693 and 1108 K to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures. The experimental results showed that there was no significant change in tensile strengths after neutron irradiation below 923 K, but the tensile strengths at neutron irradiation above 1023 K up to 33 dpa were decreased by about 20%. On the other hand, uniform elongation after irradiation was more than 2% at all irradiation conditions. The ring tensile properties of these ODS claddings remained excellent within these irradiation conditions compared with conventional 11Cr ferritic/martensitic steel (PNC-FMS) claddings.
Matsui, Yoshinori; Takahashi, Hiroyuki; Yamamoto, Masaya; Nakata, Masahito; Yoshitake, Tsunemitsu; Abe, Kazuyuki; Yoshikawa, Katsunori; Iwamatsu, Shigemi; Ishikawa, Kazuyoshi; Kikuchi, Taiji; et al.
JAEA-Technology 2009-072, 144 Pages, 2010/03
JAEA-Technology-2009-072.pdf:45.01MB
"R&D Project on Irradiation Damage Management Technology for Structural Materials of Long-life Nuclear Plant" was carried out from FY2006 in a fund of a trust enterprise of the Ministry of Education, Culture, Sports, Science and Technology. The coupled irradiations or single irradiation by JOYO fast reactor and JRR-3 thermal reactor were performed for about two years. The irradiation specimens are very important materials to establish of "Evaluation of Irradiation Damage Indicator" in this research. For the acquisition of the examination specimens irradiated by the JOYO and JRR-3, we summarized about the overall plan, the work process and the results for the study to utilize these reactors and some facilities of hot laboratory (WASTEF, JMTR-HL, MMF and FMF) of the Oarai Research-and-Development Center and the Nuclear Science Research Institute in the Japan Atomic Energy Agency.
Takaya, Shigeru; Nagae, Yuji; Yoshitake, Tsunemitsu; Nemoto, Yoshiyuki; Nakano, Junichi; Ueno, Fumiyoshi; Aoto, Kazumi; Tsukada, Takashi
E-Journal of Advanced Maintenance (Internet), 1(1), p.44 - 51, 2009/05
As the result of comparing the magnetic flux density and the IASCC susceptibility evaluated by SSRT test on neutron irradiated model alloys, it was shown that there is the relation without depending on dose level and chemical compositions as long as the contribution of neutron irradiation to SCC was seen. Furthermore, measuring the magnetic flux density of unirradiated simulated degraded materials indicates that not only change in chemical compositions but also some defects are needed for the magnetic flux density to increase. These results show the possibility of non-destructive estimation of susceptibility to IASCC by measuring magnetic flux density.
Yano, Yasuhide; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Takahashi, Heishichiro
Materia, 47(12), P. 625, 2008/12
no abstracts in English
Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Watanabe, Seiichi*; Takahashi, Heishichiro
Journal of Nuclear Science and Technology, 44(12), p.1535 - 1542, 2007/12
Times Cited Count：12 Percentile：63.69(Nuclear Science & Technology)The effects of fast neutron irradiation have been investigated on the mechanical properties of 11Cr-0.5Mo-2W, Nb, V ferritic/martensitic (F/M) stainless steel (PNC-FMS) and 10.5Cr-1.5Mo, Nb, V F/M stainless steel (HT9M) claddings, especially tensile and transient burst properties. These two F/M claddings were irradiated in the experimental fast reactor JOYO using the PFB090 fuel test assembly. Post irradiation tensile and temperature-transient-to-burst tests were carried out for defueled cladding specimens. The results of mechanical tests for PNC-FMS cladding showed that there was no significant degradation in tensile and transient burst strengths even after fast neutron irradiation. However, these strengths for HT9M cladding tended to shift to lower values than those of as-received specimens. This different behavior of tensile and transient burst strengths was attributed to martensite structural stability which was related to the stable solid solution elements.
Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Takahashi, Heishichiro*
Journal of Nuclear Materials, 367-370(1), p.127 - 131, 2007/08
Times Cited Count：11 Percentile：61.1(Materials Science, Multidisciplinary)The effects of fast neutron irradiation on tensile and transient burst properties of advanced ferritic/martensitic steel claddings were investigated. Specimens were irradiated in the experimental fast reactor JOYO using the material irradiation rig at temperatures between 773 and 1013 K to fast neutron doses ranging from 11 to 102 dpa. The post-irradiation tensile and temperature-transient-to-burst tests were carried out. The results of mechanical tests showed that there was no significant degradation in tensile and transient burst strengths after neutron irradiation below 873 K. This was attributed to grain boundary strengthening caused by precipitates that preferentially formed on prior-austenite grain boundaries. Both strengths at neutron irradiation above about 903 K up to 102 dpa decreased due to recovery of lath martensite structures and recrystallization.
Yano, Yasuhide; Oka, Keiichiro*; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Abe, Yasuhiro; Onuki, Somei
Journal of Nuclear Science and Technology, 43(6), p.648 - 654, 2006/06
Times Cited Count：4 Percentile：30.68(Nuclear Science & Technology)The embrittlement behavior of the modified ferritic/martensitic 11Cr-0.5Mo-2W, V, Nb steel (PNC-FMS) after neutron irradiation in JOYO was investigated by Charpy impact tests and TEM and SEM observations. The impact properties of the specimens after irradiation to 4.4 dpa at 773 K were similar to the as-received PNC-FMS. The ductile-brittle transition temperature (DBTT) remarkably decreased due to irradiation to 2.8 dpa at 923 K. The precipitates formed in the martensitic lath were still stable under neutron irradiation at 773 K, however they were unstable under irradiation at 923 K. The martensitic lath structure was also stable at the former irradiation temperature, but it was significantly changed at the latter. The decrease in the upper shelf energy after irradiation was related to the precipitate distribution. The changes of DBTT due to irradiation were attributed to decreased the dislocation recovery and to increased broadening of the martensitic lath.
Ueno, Fumiyoshi*; Nagae, Yuji; Nemoto, Yoshiyuki*; Miwa, Yukio*; Takaya, Shigeru; Hoshiya, Taiji; Tsukada, Takashi*; Aoto, Kazumi; Ishii, Toshimitsu*; Omi, Masao*; et al.
JNC TY9400 2005-013, 150 Pages, 2005/09
JNC-TY9400-2005-013.pdf:37.33MB
None
Ueno, Fumiyoshi; Nagae, Yuji*; Nemoto, Yoshiyuki; Miwa, Yukio; Takaya, Shigeru*; Hoshiya, Taiji*; Tsukada, Takashi; Aoto, Kazumi*; Ishii, Toshimitsu; Omi, Masao; et al.
JAERI-Research 2005-023, 132 Pages, 2005/09
JAERI-Research-2005-023.pdf:33.03MB
JAERI and JNC have started a JAERI-JNC joint research program in fiscal year 2003, which has been aimed for efficient progress and synergistic effect on the research activities in both Institutes. This study has been chosen one of the joint research themes because it has been our common objective in the field of structural materials of FBR and LWR components. The purpose of the study is to clarify damage mechanism of structural materials used under irradiation, and then to develop the methods for damage evaluation and detection in earlier stage of progressing process of damage. In fiscal year 2004 and 2005, micro-corrosion measurement, electrochemical corrosion test and leakage magnetic flux density measurement apparatuses were developed and equipped in two hot facilities and irradiated and unirradiated crept specimens, irradiated high purity model austenitic stainless alloys were also prepared and applied to this study. These apparatuses and specimens were used for damage evaluation, and these feasibilities for nuclear power plant materials were studied.
Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Ukai, Shigeharu; Kimura, Akihiko*
Materials Transactions, 46(3), p.493 - 497, 2005/05
Times Cited Count：5 Percentile：43.89(Materials Science, Multidisciplinary)Oxide dispersion strengthened (ODS) steels, which are candidate materials for water-cooled solid breeder blankets, were fabricated with several manufacturing parameters, and then irradiated in JOYO to evaluate their irradiation behavior. Engineering stress strain curve of ODS steels irradiated at 673 K exhibited superior material response, i.e., increased tensile strength due to irradiation hardening and no loss of total elongation. Also, their temperature dependence on tensile property indicated that degradation of the tensile property at elevated temperature, which is closely related to phase stability during irradiation, could be avoided due to optimal combination of manufacturing parameters, such as chemical composition, type of inert gas during mechanical alloying, heat-treatment temperature and initial phase of the matrix.
Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Ukai, Shigeharu; Onuki, Somei
Vol.475-479(2005)1467, 0 Pages, 2005/00
Three ODS steel claddings, one martensitic phase and two ferritic ones in initial structures, were neutron-irradiated in the experimental fast reactor, JOYO, and those nano-meso structural evolutions during irradiations and those ring-tensile behaviors at as close conditions as it should be in service were evaluated. As for the structural features, no significant structural changes in macro scale such as grain morphology and dislocation structure were recognizable, but those in nano scale such as carbide and oxide dispersoid emerged in all neutron-irradiated ODS claddings. With respect to the ring-tensile properties, on the other hand, a manifested variation of an irradiation temperature dependence among all the ODS claddings was revealed, reflecting the each microstructural evolution of ODS claddings during neutron irradiation.
Kurishita, Hiroaki*; Yamamoto, Takuya*; Narui, Minoru*; Yoshitake, Tsunemitsu; Akasaka, Naoaki
JNC TY9400 2004-006, 48 Pages, 2004/04
JNC-TY9400-2004-006.pdf:1.95MB
Radiation embrittlement in high-strength ferritic/martensitic steels of 2WFK and 63WFS and oxide dispersion strengthened (ODS) martensitic steel of H-35 that were irradiated in the experimental fast reactor JOYO is evaluated by instrumented Charpy impact tests for miniaturized (1.5 x 1.5 x 20 mm) and half-sized Charpy V-notch (CVN) specimens. Effects of thermal aging and microstructural evolution during irradiation on radiation embrittlement are described. Next, in order to clarify the specimen size effects on the ductile-to-brittle transition temperature (DBTT) in Charpy impact testing, a method to evaluate the plastic constraint loss for differently sized CVN specimens that may be responsible for the size effects is proposed and applied to 2 WFK. In the method, the constraint factor, a, that is an index of the plastic constraint is defined as a = s*/sy*. Here, s* is the critical cleavage fracture stress which is a material constant and sy* is the uniaxial yield stress at the DBTT at the strain rate generated in the Charpy impact test. The procedures for evaluating each of s* and sy* are described and the result of s* and sy*, thus the value of a, is presented for various types of miniaturized and full-sized CVN specimens of 2 WFK. It is shown that there is the following relationship between a and the specimen size factor, (A*/b2). a=a0-k(A*/b2)0.4 Here, A* is the critical area for cleavage fracture and b is the ligament size. a0 and k are constants depending on a /W (a is the notch depth and W is the specimen width). a increases with increasing a /W.
Yano, Yasuhide; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Abe, Yasuhiro
JNC TN9400 2003-118, 30 Pages, 2004/03
JNC-TN9400-2003-118.pdf:4.37MB
In order to evaluate the effects of microstructural changes during irradiation on the Charpy impact properties of the PNC-FMS, microstructural characterization were performed with transmission electron microscopy on ruptured halves of the half-sized Charpy specimens of PNC-FMS irradiated in the JOYO/MARICO-1.
Yamashita, Shinichiro; Oka, K.*; Yoshitake, Tsunemitsu; Akasaka, Naoaki; Ukai, Shigeharu; Onuki, Somei
Effects of Radiation on Materials (ASTM STP 1447), 391- Pages, 2003/00
None
Akasaka, Naoaki; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Ukai, Shigeharu; Kimura, Akihiko*
Proceedings of 11th International Conference on Fusion Reactor Materials (ICFRM-11), 0 Pages, 2003/00
Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.
Yano, Yasuhide; Yoshitake, Tsunemitsu; Abe, Yasuhiro
JNC TN9400 2003-028, 44 Pages, 2002/12
JNC-TN9400-2003-028.pdf:1.75MB
The effect of fast neutron irradiation on impact properties of the PNC-FMS (2WFK) and ODS martensitic steel (H35), which were irradiated in the JOYO/CMIR-5 and MARICO-1, were investigated, because it was well known that irradiation-embrittlement was one of the most important issue to apply these ferritic steels to FBR core materials. The post irradiation half-sized and miniaturized Charpy impact tests were carried out to understand the impact properties of these ferritic steels. The results obtained in this study are as follows: (1) In all irradiated 2WFK specimens, the impact properties depended on irradiation temperature rather than neutron fluence under these irradiation conditions. The impact properties of 2WFK specimens irradiated at the temperature range of 400 to 500C almost scarcely change, however, those of 2WFK specimens irradiated at 659C obviously decreased. (2) The impact properties of 2WFK specimens irradiated at 495C were similar to those of as-received 2WFK specimen. This indicates that the effects of thermal aging improvement and irradiation degradation canceled each other. Moreover, the impact properties of 2WFK specimens irradiated at 659C decreased more remarkably than those of thermal aged specimens at 650C. It is suggested that these 2WFK impact properties seemed to be understood by microstructure behaviors, especially precipitation behaviors, by means of irradiation and thermal aging. (3) It is suggested that the anisotropy of H35 ODS steel was considerably improved by the 
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phase transformation, and H35 specimens had superior impact properties even after irradiation because no radiation hardening and no change of impact properties under these irradiation conditions were observed.
