Li, Y.; Hirota, Takatoshi*; Itabashi, Yu*; Yamamoto, Masato*; Kanto, Yasuhiro*; Suzuki, Masahide*; Miyamoto, Yuhei*
JAEA-Review 2020-011, 130 Pages, 2020/09
For the improvement of the structural integrity assessment methodology on reactor pressure vessels (RPVs), the probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed and improved in Japan Atomic Energy Agency based on the latest knowledge. The PASCAL code evaluates the failure probabilities and frequencies of Japanese RPVs under transient events such as pressure thermal shock considering neutron irradiation embrittlement. In order to confirm the reliability of the PASCAL as a domestic standard code and to promote the application of PFM on the domestic structural integrity assessments of RPVs, it is important to perform verification activities, and summarize the verification processes and results as a document. On the basis of these backgrounds, we established a working group, composed of experts on this field besides the developers, on the verification of the PASCAL module and the source program of PASCAL was released to the members of working group. This report summarizes the activities of the working group on the verification of PASCAL in FY2016 and FY2017.
Nagai, Takayuki; Okamoto, Yoshihiro; Akiyama, Daisuke*; Uehara, Akihiro*; Fujii, Toshiyuki*; Sekimoto, Shun*
KURNS Progress Report 2019, P. 257, 2020/08
To understand the influence of glass structural change by neutron irradiation and boron isotope composition, glass samples were made from enrichment boric acid reagents and neutron irradiation of those samples was carried out in Pn-2 of KUR. The structural change of glass sample after the irradiation will be estimated in 2020FY. Before neutron irradiation test of glass samples, the Si-O bridging structure difference by boron isotope composition compared by using a Raman spectrometry.
Saito, Shigeru; Suzuki, Kazuhiro; Hatakeyama, Yuichi; Suzuki, Miho; Dai, Y.*
Journal of Nuclear Materials, 534, p.152146_1 - 152146_16, 2020/06
A post-irradiation examination (PIE) was performed on the tensile specimens prepared from the MEGAPIE (MEGAwatt Pilot Experiment) target which were irradiated in flowing lead-bismuth eutectic (LBE). Thicknesses of the specimens were over two times larger than that of the standard specimen. The PIE revealed that the T91 specimens showed a 1.5-2.0 times larger total elongation (TE) compared to the literature values for a specimen with standard t/w (ratio of thickness to width). It could be suggested that the t/w and TE were strongly correlated. Then, we tried to investigate the effects of the t/w on the TE by comparing unirradiated specimens. We found that there was no t/w dependence on the strength and uniform elongation. On the other hand, the TE increases with increasing t/w. Based on the experimental data, we correlated the TE with various specimens t/w to estimate appropriate TE values, including that for the standard specimen.
Nagai, Takayuki; Kobayashi, Hidekazu; Okamoto, Yoshihiro; Akiyama, Daisuke*; Sato, Nobuaki*; Uehara, Akihiro*; Fujii, Toshiyuki*; Sekimoto, Shun*
KURNS Progress Report 2018, P. 105, 2019/08
To understand this structural change of a borosilicate glass by a neutron irradiation in detail, the irradiation test was carried out in KUR in 2017FY. The glass structure was estimated by using Raman spectrometry in 2018FY. Comparing with the Raman spectra of glass samples before and after irradiation, it could be observed the change of peak height of Si-O bridging structure by the irradiation.
Kasahara, Shigeki; Chimi, Yasuhiro; Hata, Kuniki; Fukuya, Koji*; Fujii, Katsuhiko*
Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), p.1345 - 1355, 2019/08
This paper describes empirical equation development of crack growth rates (CGR) in consideration of IASCC of neutron irradiated stainless steel to contribute to structural integrity assessment of BWR reactor internals. Empirical equations of CGR (da/dt) were developed based on a formula of da/dt = MK, assuming that "M" and "n" tend to be saturated with increasing neutron fluence. To obtain the empirical equations for normal water chemistry (NWC) and hydrogen water chemistry (HWC) of BWR, a data fitting with least square method was applied to the datasets consisting of F, K and CGR from post irradiation examinations of neutron irradiated stainless steel under simulated NWC and HWC conditions from open literature. As a result, calculated results by the equation for NWC showed good agreement with measured CGR data, meanwhile those for HWC did not. The above difference was seemed to be attributed that CGR data obtained under HWC conditions were scattered extensively.
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.
Takamizawa, Hisashi; Katsuyama, Jinya; Ha, Yoosung; Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 8 Pages, 2019/07
no abstracts in English
Sekio, Yoshihiro; Sakaguchi, Norihito*
Materials Transactions, 60(5), p.678 - 687, 2019/05
The quantitative evaluation of vacancy migration energies in high nickel model alloy was conducted by analyzing the void denuded zone (VDZ) width formed near grain boundaries under neutron and electron irradiation. The microstructures of Fe-15Cr-xNi (x=15, 20, 25, 30 mass%) alloys that were neutron irradiated at 749 K and electron irradiated at 576 K-824 K were examined. The VDZ widths increased with increasing Ni content in both irradiation experiments, which implies an increase of the vacancy mobility. The vacancy migration energies were estimated from the temperature dependence of the VDZ widths, and the energies were 1.09, 0.97, 0.90, and 0.77 eV for the alloys containing 15, 20, 25, and 30 mass% Ni, respectively. From the obtained energies, the effective vacancy diffusivity and excess vacancy concentration were estimated using the analytical equation of the VDZ width, which quantitatively confirmed the increase of the vacancy mobility with increasing Ni content.
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.