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Takamizawa, Hisashi; Lu, K.; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2022-006, 221 Pages, 2023/02
As a part of the structural integrity assessment research for aging light water reactor (LWR) components, a probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in Japan Atomic Energy Agency. The PASCAL code can evaluate failure probabilities and failure frequencies of core region in reactor pressure vessel (RPV) under transients by considering the uncertainties of influential parameters. The continuous development of the code aims to improve the reliability by introducing the analysis methodologies and functions base on the state-of-the-art knowledge in fracture mechanics and domestic data. In the first version of PASCAL, which was released in FY2000, the basic framework was developed for analyzing failure probabilities considering pressurized thermal shock events for RPVs in pressurized water reactors (PWRs). In PASCAL Ver. 2 released in FY 2006, analysis functions including the evaluation methods for embedded cracks and crack detection probability models for inspection were introduced. In PASCAL Ver. 3 released in FY 2010, functions considering weld-overlay cladding on the inner surface of RPV were introduced. In PASCAL Ver. 4 released in FY 2017, we improved several functions such as the stress intensity factor solutions, probabilistic fracture toughness evaluation models, and confidence level evaluation function by considering epistemic and aleatory uncertainties related to influential parameters. In addition, the probabilistic calculation method was also improved to speed up the failure probability calculations. To strengthen the practical applications of PFM methodology in Japan, PASCAL code has been improved since FY 2018 to enable PFM analyses of RPVs subjected to a broad range of transients corresponding to both PWRs and boiling water reactors, including pressurized thermal shock, low-temperature over pressure, and normal operational transients. In particular, the stress intensi
Hayakawa, Sho*; Yamamoto, Yojiro*; Okita, Taira*; Itakura, Mitsuhiro; Suzuki, Katsuyuki*
Computational Materials Science, 218, p.111987_1 - 111987_10, 2023/02
Times Cited Count:1 Percentile:14.66(Materials Science, Multidisciplinary)Tsugawa, Kiyoto*; Hayakawa, Sho*; Okita, Taira*; Aichi, Masaatsu*; Itakura, Mitsuhiro; Suzuki, Katsuyuki*
Computational Materials Science, 215, p.111806_1 - 111806_8, 2022/12
Times Cited Count:2 Percentile:29.01(Materials Science, Multidisciplinary)Takamizawa, Hisashi; Hata, Kuniki; Nishiyama, Yutaka; Toyama, Takeshi*; Nagai, Yasuyoshi*
Journal of Nuclear Materials, 556, p.153203_1 - 153203_10, 2021/12
Times Cited Count:3 Percentile:31.78(Materials Science, Multidisciplinary)Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 10 n/cm were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Ni, Mn, and Si atoms like a core-shell structure. In low-Cu bearing materials, Ni, Mn, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius () and number density () decreased and increased, respectively, with increasing nominal Si content. The shift in the reference temperature for nil-ductility transition (RT) showed a good correlation with the square root of volume fraction () multiplied by r (). This suggested that the dislocation cutting through the particles mechanism dominates the precipitation hardening responsible for irradiation embrittlement. The negative relation between the nominal Si content and RT indicated that increasing of nominal Si content reduces the degree of embrittlement.
Takamizawa, Hisashi; Nishiyama, Yutaka
Journal of Pressure Vessel Technology, 143(5), p.051502_1 - 051502_8, 2021/10
Times Cited Count:3 Percentile:30.36(Engineering, Mechanical)no abstracts in English
Hata, Kuniki; Takamizawa, Hisashi; Hojo, Tomohiro*; Ebihara, Kenichi; Nishiyama, Yutaka; Nagai, Yasuyoshi*
Journal of Nuclear Materials, 543, p.152564_1 - 152564_10, 2021/01
Times Cited Count:12 Percentile:91.16(Materials Science, Multidisciplinary)Reactor pressure vessel (RPV) steels for pressurized water reactors (PWRs) with bulk P contents ranging from 0.007 to 0.012wt.% were subjected to neutron irradiation at fluences ranging from 0.3 to 1.210 n/cm (E 1 MeV) in PWRs or a materials testing reactor (MTR). Grain-boundary P segregation was analyzed using Auger electron spectroscopy (AES) on intergranular facets and found to increase with increasing neutron fluence. A rate theory model was also used to simulate the increase in grain-boundary P segregation for RPV steels with a bulk P content up to 0.020wt.%. The increase in grain-boundary P segregation in RPV steel with a bulk P content of 0.015wt.% (the maximum P concentration found in RPV steels used in Japanese nuclear power plants intended for restart) was estimated to be less than 0.1 in monolayer coverage at 1.010 n/cm (E 1 MeV). A comparison of the PWR data with the MTR data showed that neutron flux had no effect upon grain-boundary P segregation. The effects of grain-boundary P segregation upon changes in irradiation hardening and ductile-brittle transition temperature (DBTT) shifts were also discussed. A linear relationship between irradiation hardening and the DBTT shift with a slope of 0.63 obtained for RPV steels with a bulk P content up to 0.026wt.%, which is higher than that of most U.S. A533B steels. It is concluded that the intergranular embrittlement is unlikely to occur for RPV steels irradiated in PWRs.
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.
Takamizawa, Hisashi; Nishiyama, Yutaka; Hirano, Takashi*
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08
no abstracts in English
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
Katsuyama, Jinya; Masaki, Koichi; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2017-015, 229 Pages, 2018/03
As a part of the structural integrity research for aging light water reactor components, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAEA. The PASCAL code can evaluate the conditional failure probabilities and failure frequencies for core region in reactor pressure vessels under the pressurized thermal shock events. In this study, we improved many functions such as the stress intensity factor solutions, the fracture toughness models, or confidence level evaluation function by considering epistemic and aleatory uncertainties related to influence parameters in the structural integrity assessment. We also developed the analysis module PASCAL-Manager which calculates the failure frequency for the entire core region taking into consideration the failure probabilities obtained from PACAL-RV. Based on these improvements, the new analysis code is upgraded to PASCAL Ver.4. This report provides the user's manual and theoretical background of PASCAL Ver.4.
Takamizawa, Hisashi; Ito, Hiroto; Nishiyama, Yutaka
Journal of Nuclear Materials, 479, p.533 - 541, 2016/10
Times Cited Count:6 Percentile:49.29(Materials Science, Multidisciplinary)To understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters (such as mean and standard deviation) for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). Clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel, neutron flux, neutron fluence, and irradiation temperatures. It was found through numerous examinations that the measured shifts of DBTT correlated well with calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were significantly disparate among the results. This indicates that slowly developing or late-onset embrittlement mechanisms were not evident in the present study.
Takamizawa, Hisashi; Nishiyama, Yutaka
Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 5 Pages, 2016/07
It has been accepted that neutron irradiation embrittlement of reactor pressure vessel is caused by irradiation-induced formation of solute clusters (SCs) and matrix damages (MDs). In the present study, to analyze the contribution of chemical composition contained in SCs to irradiation embrittlement at high fluence region, statistical analysis using the Bayesian nonparametric (BNP) method was performed for Japanese PWR surveillance data. The significance of P, Si and Mn contents, which are not necessarily included in embrittlement correlations unlike the Cu and Ni content, was evaluated. The BNP method can learn the complexity of the statistical model itself from the input data and infer the predicted data with individual probability distribution of predict condition. The result suggested that irradiation embrittlement was most affected by the Si content at high fluence region.
Wakai, Eiichi; Jitsukawa, Shiro; Tomita, Hideki*; Furuya, Kazuyuki; Sato, Michitaka*; Oka, Keiichiro*; Tanaka, Teruyuki*; Takada, Fumiki; Yamamoto, Toshio*; Kato, Yoshiaki; et al.
Journal of Nuclear Materials, 343(1-3), p.285 - 296, 2005/08
Times Cited Count:48 Percentile:93.91(Materials Science, Multidisciplinary)The dependence of helium production on radiation-hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel doped with B, B and B+B irradiated at 250C to 2.2 dpa. The total amounts of doping boron were about 60 massppm. The range of He concentration produced in the specimens was from about 5 to about 300 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He irradiation was also performed to implant about 85 appm He atoms at 120C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain DBTT. Radiation-hardening of the neutron-irradiated specimens increased slightly with increasing He production. The 100 MPam DBTT for the F82H+B, F82H+B+B, and F82H+B were 40, 110, and 155C, respectively. The shifts of DBTT due to He production were evaluated as about 70C by 150 appmHe and 115C by 300 appmHe. The DBTT shift in the small punch testing was evaluated as 50C.
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
Times Cited Count:19 Percentile:75.21(Materials Science, Multidisciplinary)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.
Ooka, Norikazu*; Ishii, Toshimitsu
Hihakai Kensa, 52(5), p.235 - 239, 2003/05
no abstracts in English
Hasegawa, Masayuki*; Nagai, Yasuyoshi*; Tang, Z.*; Yubuta, Kunio*; Suzuki, Masahide
JAERI-Tech 2003-015, 137 Pages, 2003/03
no abstracts in English
Suzuki, Masahide; Nishiyama, Yutaka
Kinzoku, 71(8), p.42 - 45, 2001/08
no abstracts in English
Ishii, Toshimitsu; Ooka, Norikazu; Niimi, Motoji; Kobayashi, Hideo*
Kinzoku, 71(8), p.20 - 24, 2001/08
no abstracts in English
Onizawa, Kunio; Suzuki, Masahide
Effects of Radiation on Materials: 20th International Symposium (ASTM STP 1405), p.79 - 96, 2001/07
no abstracts in English
Tsukada, Takashi; Ebine, Noriya
Nihon AEM Gakkai-Shi, 9(2), p.171 - 177, 2001/06
no abstracts in English