Kasahara, Shigeki; Fukuya, Koji*; Fujimoto, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro
JAEA-Review 2018-013, 171 Pages, 2019/01
For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. When the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of pressurized water reactors (PWR) and boiling water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of PWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of PWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into tables.
Kasahara, Shigeki; Fukuya, Koji*; Koshiishi, Masato*; Fujii, Katsuhiko*; Chimi, Yasuhiro
JAEA-Review 2018-012, 180 Pages, 2018/11
For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. In the process of the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of boiling water reactors (BWR) and pressurized water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of BWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of BWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into the tables.
Nagai, Takayuki; Uehara, Akihiro*; Fujii, Toshiyuki*
KURRI Progress Report 2017, P. 128, 2018/08
In this study, to understand the structural change of a borosilicate glass by a neutron irradiation in detail, the neutron irradiation test was carried out for 50 min in the Pn-2 of the Kyoto University Research Reactor (KUR). The structural change of glass sample after the irradiation will be estimated in 2018FY. Before the irradiation test, the glass composition was selected to estimate a structural change accurately by Raman spectrometry and 2 kinds of Li free borosilicate glass for the irradiation test samples were prepared.
Nemoto, Yoshiyuki; Oishi, Makoto; Ito, Masayasu; Kaji, Yoshiyuki; Keyakida, Satoshi*
Hozengaku, 14(4), p.83 - 90, 2016/01
Authors previously reported that magnetic data obtained by using Eddy current method and AC magnetization method showed correlation with the increase of susceptibility of the irradiation assisted stress corrosion cracking (IASCC) on neutron irradiated austenitic stainless alloy specimens. To discuss the mechanism of the correlation, microstructure observation was conducted on the irradiated specimen, and magnetic permalloy phase (FeNi) formation along grain boundary was revealed in this work. From this result, the radiation induced magnetic phase formation along grain boundary seems to be a factor of the magnetic property change of the irradiated materials, and related to the correlation between magnetic data and IASCC susceptibility. In addition, sensor probe development was conducted in this work to obtain higher sensitivity and resolution. It was applied for magnetic measurement on type304 stainless steel irradiated up to different doses. In this case, magnetic ferrite phase was existed in the type304 stainless steel sample before irradiation therefore it was concerned that magnetic measurement on the irradiated ones would be disturbed by the magnetic signal from the pre-existing ferrite phase. In the magnetic measurements, increase of the magnetic data was clearly seen on the irradiated specimens. Thus, it was thought that the developed magnetic measurement technics can be applied for the irradiated austenite stainless steels which contain certain quantity of ferrite phase before irradiation.
Nagai, Takayuki; Kobayashi, Hidekazu; Okamoto, Yoshihiro; Uehara, Akihiro*; Fujii, Toshiyuki*
Photon Factory Activity Report 2015, Part B, 2 Pages, 2016/00
To investigate the characterization damage of a borosilicate glass by a neutron irradiation, the glass sample after neutron irradiation was estimated by using the Raman spectrophotometry and the synchrotron XAFS measurement. As a result, we confirmed that the Si-O bridge structure of a borosilicate glass and the containing element valence in the glass were changed by the neutron irradiation.
Nishikata, Kaori; Ishida, Takuya; Yonekawa, Minoru; Kato, Yoshiaki; Kurosawa, Makoto; Kimura, Akihiro; Matsui, Yoshinori; Tsuchiya, Kunihiko; Sano, Tadafumi*; Fujihara, Yasuyuki*; et al.
KURRI Progress Report 2014, P. 109, 2015/07
As one of effective applications of the Japan Materials Testing Reactor (JMTR), JAEA has a plan to produce Mo by (n,) method ((n,)Mo production), a parent nuclide of Tc. In this study, preliminary irradiation test was carried out with the high-density molybdenum trioxide (MoO) pellets in the hydraulic conveyer (HYD) of the Kyoto University Research Reactor (KUR) and the Tc solution extracted from Mo was evaluated. After the irradiation test of the high-density MoO pellets in the KUR, Tc was extracted from the Mo solution and the recovery rate of Tc achieved the target values. The Tc solution also got the value that satisfied the standard value for Tc radiopharmaceutical products by the solvent extraction method.
Yamada, Hirokazu*; Sato, Satoshi; Mori, Kensuke*; Nagao, Yoshiharu; Takada, Fumiki; Kawamura, Hiroshi
Fusion Engineering and Design, 81(1-7), p.631 - 637, 2006/02
This study estimated the neutron irradiation effect with 1.5 dpa on the mechanical properties of the SS/SS HIP joint materials jointed in the standard HIP joint condition. Results of this study showed that the HIP process in the standard HIP condition could make SS/SS HIP joint material of which tensile properties was equivalent to that of the SS base material. In addition, the effect of surface roughness at the HIP joint material on the mechanical properties of SS/SS HIP joint material was estimated.
Yamada, Hirokazu*; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, G.*; Nagao, Yoshiharu; Takada, Fumiki; Nishikawa, Masahiro*
Journal of Nuclear Materials, 340(1), p.57 - 63, 2005/04
This study was performed to clear the effect of neutron re-irradiation on mechanical properties to welding material un-irradiated and irradiated stainless steel. The effect of re-irradiation to these weldaments were evaluated by tensile tests, metallographical observation and hardness test. The result of tensile tests shows that ultimate tensile strength of all joints specimen were almost similar, 0.2% yield strength and total elongation were depend on the irradiation damage and weldaments or not. Therefore, fracture mechanism was not change by re-irradiation. However, brittlement of material was depend on irradiation damage and the property of deformation was sensitive by the effect of irradiation damage or the effect of welding heat.
Yamada, Hirokazu*; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, G.*; Nagao, Yoshiharu; Sato, Satoshi; Mori, Kensuke*
Journal of Nuclear Materials, 335(1), p.33 - 38, 2004/10
no abstracts in English
Department of Materials Science; Department of Fusion Engineering Research (Tokai Site)
JAERI-Review 2004-018, 97 Pages, 2004/08
Extensive efforts for evaluating the irradiation performances of a reduced activation ferritic/martensitic steel (RAF/M) of F82H* and other several RAF/Ms have been made in recent several years. They are, examinations of the effects of neutron irradiation on (1) Ductile to brittle transition temperature (DBTT) up to a damage level of 20 dpa to explore lower temperature limit, (2) Enhanced He effect on DBTT shift for Ni/B doped heats (isotopic tailoring method was used for B doping), (3) Susceptibility to environmentally assisted cracking by the slow strain rate tensile tests (SSRT) in a high temperature pressurized water and (4) Flow stress-plastic strain relation obtained by measuring the profile of the specimen during tensile testing, together with the activities of (5) the development of the test methods after neutron irradiation and (6) other supporting researches. Results are summarized in the present report. They clearly indicate the good applicability of RAF/Ms to fusion machines.
Tanigawa, Hiroyasu; Hashimoto, Naoyuki*; Sakasegawa, Hideo*; Klueh, R. L.*; Sokolov, M. A.*; Shiba, Kiyoyuki; Jitsukawa, Shiro; Koyama, Akira*
Journal of Nuclear Materials, 329-333(1), p.283 - 288, 2004/08
Reduced-activation ferritic/martensitic steels (RAFs) were developed as candidate structural materials for fusion power plants. In a previous study, it was reported that ORNL9Cr-2WVTa and JLF-1 (Fe-9Cr-2W-V-Ta-N) steels showed smaller ductile-brittle transition temperature (DBTT) shifts compared to IEA modified F82H (Fe-8Cr-2W-V-Ta) after neutron irradiation up to 5 dpa at 573K. This difference in DBTT shift could not be interpreted as an effect of irradiation hardening, and it is also hard to be convinced that this difference was simply due to a Cr concentration difference. To clarify the mechanisms of the difference in Charpy impact property between these steels, various microstructure analyses were performed.
Miwa, Yukio; Jitsukawa, Shiro; Yonekawa, Minoru
Journal of Nuclear Materials, 329-333(Part2), p.1098 - 1102, 2004/08
Fatigue properties were examined on a reduced activation ferritic/martensitic steel, and preliminary results were presented. F82H steel was irradiated at 523 K to 3.8 dpa, and then fatigue-tested at 298-573 K in vacuum with total strain range of 0.4-1.0%. Effect of irradiation on fatigue lives was observed on test at 298 K with total strain range of 0.4%. The fatigue life of irradiated specimen was reduced to about 1/7 of unirradiated specimen. The reduction of the fatigue life was attributed to the occurrence of channel fracture. Effect of test temperature was discussed.
Wakai, Eiichi; Matsukawa, Shingo; Yamamoto, Toshio*; Kato, Yoshiaki; Takada, Fumiki; Sugimoto, Masayoshi; Jitsukawa, Shiro
Materials Transactions, 45(8), p.2641 - 2643, 2004/08
no abstracts in English
Tanigawa, Hiroyasu; Sakasegawa, Hideo*; Zinkle, S. J.*; Klueh, R. L.*; Koyama, Akira*
Fusion Materials Semiannual Progress Report for the Period Ending (DOE/ER-0313/35), p.30 - 32, 2004/04
Extraction residue was made from several HFIR 11J-irradiated RAFs, and the mass change was measured to investigate the irradiation-enhanced change in precipitation. Two different types of filter with coarse and fine pores were used in order to separate the difference of irradiation effects between larger and smaller precipitates. Unirradiated specimens were examined as well. Results suggest that during irradiation the mass of larger precipitates increased in F82H-IEA, Ni-doped F82H, JLF-1 and ORNL9Cr, fine precipitates disappeared in JLF-1, and fine precipitates increased in Ni-doped F82H.
Tanigawa, Hiroyasu; Sakasegawa, Hideo*; Payzant, E. A.*; Zinkle, S. J.*; Klueh, R. L.*; Koyama, Akira*
Fusion Materials Semiannual Progress Report for the Period Ending (DOE/ER-0313/35), p.37 - 40, 2004/04
XRD analyses were performed on the extraction residue of HFIR 11J-irradiated RAFs to investigate the overall precipitate character. Un-irradiated and aged specimens were examined as well. Results suggested that the distinctive peaks of M23C6 (M; Cr, Fe, W) were observed on all specimens. Peaks possibly related to MX (M;Ta,Ti,V : X ; C, N) were observed on the specimens extracted from un-irradiated JLF-1 and ORNL9Cr, but those peaks were not observed on irradiated specimens.
Kaminaga, Katsuo; Tsuboi, Kazuaki; Kusunoki, Hidehiko; Asano, Norikazu; Yanai, Tomohiro
JAERI-Tech 2003-051, 26 Pages, 2003/05
The sixth replacement of beryllium frame has been accomplished for the west and north frame, while the old east frame of fifth generation is preserved for use. JMTR achieved its first criticality in 1968 and has been in operation until today, with an integrated thermal output of 141.454.5 MWd. During that period replacement was carried out 4 times for all the frames, east, west and north. This time, differently from the previous replacement, west and north frames only were manufactured and installed, when the old east frame was left for continuous use, based on the past experience. Such a strategy contributed to reduction of budget and the waste associated.
Ooka, Norikazu*; Ishii, Toshimitsu
Hihakai Kensa, 52(5), p.235 - 239, 2003/05
no abstracts in English
Tanigawa, Hiroyasu; Hirose, Takanori; Ando, Masami; Jitsukawa, Shiro; Kato, Yudai*; Koyama, Akira*
Journal of Nuclear Materials, 307-311(Part1), p.293 - 298, 2002/12
It has been a key issue to get the mechanical understanding of fracture process on microstructure basis, especially on neutron-irradiated materials, but not yet to be understood well enough as for the difficulty of making transmission electron microscope (TEM) thin film sample from mechanical-tested specimen. To solve this technical problem, the focused ion beam (FIB) micro-sampling system was installed to the Research Hot Laboratory of Japan Atomic Research Institute (JAERI), Japan. This system makes it possible to fabricate the TEM specimens from the critical points of mechanical-tested radioactive specimens, such as the crack initiation points of fatigue fracture on neutron irradiated specimen. In this paper, the microstructure of mechanical-tested specimen of Reduced Activation Ferritic/martensitic steels, RAFs are investigated focusing on the helium effects to fatigue fracture.
Saito, Shigeru; Fukaya, Kiyoshi*; Ishiyama, Shintaro; Amezawa, Hiroo; Yonekawa, Minoru; Takada, Fumiki; Kato, Yoshiaki; Takeda, Takashi; Takahashi, Hiroyuki*; Nakahira, Masataka
Journal of Nuclear Materials, 307-311(Part2), p.1573 - 1577, 2002/12
no abstracts in English
Sawai, Tomotsugu; Wakai, Eiichi; Jitsukawa, Shiro; Hishinuma, Akimichi
Journal of Nuclear Materials, 307-311(Part1), p.389 - 392, 2002/12
A Ti-35Al-10V alloy was fabricated from mechanically alloyed powder by a hot-isostatic-pressing. The microstructure consists of 2, , and phases. Specimens were irradiated in Japan Research Reactor No. 3 Modified (JRR-3M) up to 3.5 10 n/cm at 400C and 600C. Unirradiated tensile specimens showed total elongation of 3 to 15 % at 400C-tests, while 400C-irradiated or 600C-irradiated specimens showed no plastic deformation before fracture. At 600C tensile tests, unirradiated specimens showed total elongation of more than 60 %, while irradiated ones showed 10 % or less elongation. The low ductility of irradiated specimens suggests embrittlement due to phase decomposition, but electron diffraction using a transmission electron microscope results of irradiated specimens will be also discussed.