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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 = M
K
, 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 280-350
C 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.
Kasahara, Shigeki; Fukuya, Koji*; Fujimoto, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro
JAEA-Review 2018-013, 171 Pages, 2019/01
JAEA-Review-2018-013.pdf:6.89MB
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.
Ara, Katsuyuki*; Ebine, Noriya
Denki Gakkai Magunetikkusu Kenkyukai Shiryo (MAG-00-182), p.23 - 31, 2000/09
no abstracts in English
Kaji, Yoshiyuki; Tsukada, Takashi; Tsuji, Hirokazu; Nakajima, Hajime
Proceedings of 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems, p.987 - 995, 1999/00
no abstracts in English
Sakai, Akihiro; Yoshimori, Michiro; Okoshi, Minoru; Abe, Masayoshi
Proceedings of International Waste Management Symposium '99 (Waste Manegement '99) (CD-ROM), 14 Pages, 1999/00
no abstracts in English
Tsukada, Takashi
Proceedings of Seminar on Water Chemistry of Nuclear Reactor Systems '99, p.26 - 32, 1999/00
no abstracts in English
Tsukada, Takashi
JAERI-Research 98-007, 187 Pages, 1998/03
JAERI-Research-98-007.pdf:17.46MB
no abstracts in English
*; Ioki, Kimihiro*; F.Elio*; Kodama, T.*; S.Chiocchio*; D.Williamson*; M.Roccella*; P.Barabaschi*; R.S.Sayer*
Fusion Technology 1998, 2, p.1389 - 1392, 1998/00
no abstracts in English
RANDEC Nyusu, 0(27), p.6 - 7, 1995/10
no abstracts in English
RANDEC Nyusu, 0(13), p.5 - 7, 1992/05
no abstracts in English
Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni
Journal of Nuclear Science and Technology, 28(8), p.767 - 776, 1991/08
no abstracts in English
*; Tone, Tatsuzo
JAERI-M 85-203, 35 Pages, 1985/12
no abstracts in English
Kasahara, Shigeki; Fukuya, Koji*; Chimi, Yasuhiro; Fujii, Katsuhiko*; Koshiishi, Masato*
no journal, ,
no abstracts in English
Fukuya, Koji*; Chimi, Yasuhiro; Kasahara, Shigeki; Fujii, Katsuhiko*; Fujimoto, Koji*
no journal, ,
no abstracts in English
Chimi, Yasuhiro; Fukuya, Koji*; Kasahara, Shigeki; Fujii, Katsuhiko*; Hanawa, Satoshi
no journal, ,
To contribute to materials degradation assessment in core internals of light water reactors, we have widely surveyed the literature on irradiation properties of austenitic stainless steels (SSs) and summarized the data. In the present investigation, in terms of mechanical properties (tensile properties, hardness, and fracture toughness), IASCC properties (IASCC susceptibility, IASCC initiation, and IASCC growth), stress relaxation - creep - swelling, and microstructural properties (microstructures and grain boundary segregation) in irradiated SSs, we have collected the data and made the spread sheets separately on PWR and BWR according to the test conditions and the target materials. We have also considered the trend curves on the mechanical properties, the IASCC properties, etc. based on the knowledge obtained through surveying the literature on dose dependence of the irradiation properties. In the presentation, we will report the overview of the present investigation.
Kasahara, Shigeki; Fukuya, Koji*; Chimi, Yasuhiro; Fujii, Katsuhiko*; Koshiishi, Masato*
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.
Kasahara, Shigeki; Fukuya, Koji*; Chimi, Yasuhiro; Fujii, Katsuhiko*; Hata, Kuniki
no journal, ,
Disposition curves of crack growth rates (CGR) of stainless steel in appropriate consideration of IASCC are necessary for structural integrity assessment of reactor internals of BWRs. This paper describes empirical equations development of CGR (da/dt), as functions of stress intensity factors (K) and neutron dose (F) to contribute to improvement of the structural integrity assessment. Development started from a formula of da/dt=MK, and on the assumption that "M" and "n" tend to be saturated with increasing F. Datasets for fitting were prepared consisting of CGR, F and K from the results of PIE under simulated NWC conditions. Data fitting with least square method was applied to the datasets to obtain the equation. The results from the empirical equation were compared with the measured crack growth data, and validity of the equations were discussed from the viewpoints of statistical analysis.
Shimodaira, Masaki; Hata, Kuniki; Iwata, Keiko; Ha, Yoosung; Kasahara, Shigeki; Katsuyama, Jinya
no journal, ,
no abstracts in English
Hata, Kuniki; Iwata, Keiko; Shimodaira, Masaki; Ha, Yoosung; Takamizawa, Hisashi; Katsuyama, Jinya
no journal, ,
no abstracts in English
