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Journal Articles

Verification of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessel

Lu, K.; Takamizawa, Hisashi; Li, Y.; Masaki, Koichi*; Takagoshi, Daiki*; Nagai, Masaki*; Nannichi, Takashi*; Murakami, Kenta*; Kanto, Yasuhiro*; Yashirodai, Kenji*; et al.

Mechanical Engineering Journal (Internet), 10(4), p.22-00484_1 - 22-00484_13, 2023/08

Journal Articles

Analysis of the effect of pre-crack curvature in Mini-C(T) specimen on fracture toughness evaluation

Shimodaira, Masaki; Ha, Yoosung; Takamizawa, Hisashi; Katsuyama, Jinya; Onizawa, Kunio

Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 11 Pages, 2023/07

In the current structural integrity assessment of the reactor pressure vessel, the accurate reference temperature (T$$_{o}$$) based on the Master Curve method is necessary. The T$$_{o}$$ can be estimated by using the Mini-C(T) fracture toughness specimen in accordance with ASTM E1921 and JEAC4216, which prescribe the pre-crack straightness criteria. A requirement in ASTM E1921 has been revised in a decade to increase the accuracy and reasonability, and the applicable crack curvature has been varied by applied codes. The pre-crack curvature of the Mini-C(T) specimen might have an impact on the T$$_{o}$$ because of the variation of the plastic constraint. In this work, the effect of the crack curvature on the fracture toughness (K$$_{Jc}$$) evaluation using the Mini-C(T) specimen was quantitatively evaluated by using the finite element analysis (FEA) including the Weibull stress analysis, to discuss the difference in a requirement of the crack straightness in ASTM E1921 and JEAC4216. FEAs showed a possibility that the upper limit curvature would decrease the plastic constraint, and consequently obtain higher K$$_{Jc}$$ in the Mini-C(T) specimen. Furthermore, if the upper limit curvature according to the ASTM E1921-21 was allowed, the T$$_{o}$$ would be estimated as non-conservative based on the Weibull stress analysis. In contrast, the difference in (T$$_{o}$$) between the crack with upper limit curvature according to JEAC4216 and the ideal straight crack was not significant.

Journal Articles

Development of stress intensity factor solution for surface crack at nozzle corner in reactor pressure vessel

Yamaguchi, Yoshihito; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.

Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 9 Pages, 2023/07

The stress intensity factor (SIF) for crack at nozzle corner is a key parameter in structural integrity assessment of nozzle in reactor pressure vessel (RPV). Although various SIF solutions for surface cracks at nozzle corners have been proposed, most of them are only focusing on the deepest point of the crack, and the information about geometric dimension of the nozzle corner is not clear. According to the previous fatigue test results regarding the surface crack at the nozzle corner, the amounts of crack growth at the surface points were larger than that at the deepest point of the crack. Such results imply that SIFs at the surface points may be higher than that at the deepest point. To increase the reliability of the structural integrity assessment, it is necessary to provide SIF solutions for both surface and deepest points. In this study, SIF solutions for two surface points and the deepest point of surface crack at nozzle corners are developed through finite element analyses and the solutions are provided corresponding to the geometric dimensions of nozzle corner and crack size.

Journal Articles

Fracture toughness evaluation of the heat-affected zone under the weld overlay cladding in reactor pressure vessel steel

Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya

Journal of Pressure Vessel Technology, 145(2), p.021501_1 - 021501_9, 2023/04

 Times Cited Count:1 Percentile:26.56(Engineering, Mechanical)

JAEA Reports

Guideline on structural integrity assessment for reactor pressure vessel in domestic light water reactor based on probabilistic fracture mechanics

Lu, K.; Katsuyama, Jinya; Takamizawa, Hisashi; Li, Y.

JAEA-Research 2022-012, 39 Pages, 2023/02

JAEA-Research-2022-012.pdf:1.72MB

For reactor pressure vessels (RPVs) in the light water reactors, the fracture toughness decreases due to the neutron irradiation embrittlement with operating years. In Japan, to prevent RPVs from a nil-ductile fracture, deterministic fracture mechanics methods in accordance with the codes provided by the Japan Electric Association are performed for assessing the structural integrity of RPVs under the pressurized thermal shock (PTS) events by taking the neutron irradiation embrittlement into account. On the other hand, in recent years, probabilistic methodologies for PTS evaluation are introduced into regulations in the United States and some European countries. For example, in the United States, a PTS screening criterion related to the reference temperature based on the probabilistic method is stipulated. If the screening criterion is not satisfied, it is allowable to perform the evaluation based on the probabilistic method by calculating numerical index such as through-wall crack frequency (TWCF). In addition, the reduction of non-destructive examination extent or extension of examination intervals for RPV welds have been discussed based on the probabilistic method. Here, the probabilistic method is a structural integrity assessment method based on probabilistic fracture mechanics (PFM) which is rational in calculating the failure probability of components by considering uncertainties of various factors related to the aged degradation due to the long-term operation. Based on these backgrounds, we developed a PFM analysis code PASCAL and released a guideline on structural integrity assessment based on PFM by reflecting the latest knowledge and expertise in 2017. Here, the main analysis target was the RPV of pressurized water rector considering neutron irradiation embrittlement and PTS events in the structural integrity assessment of RPVs. The objective of the guideline is that persons who have knowledge on the fracture mechanics can carry out the PFM analyses and

JAEA Reports

User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.5 for reactor pressure vessels

Takamizawa, Hisashi; Lu, K.; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.

JAEA-Data/Code 2022-006, 221 Pages, 2023/02

JAEA-Data-Code-2022-006.pdf:4.79MB

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

Journal Articles

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

Lu, K.; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.

International Journal of Pressure Vessels and Piping, 199, p.104706_1 - 104706_13, 2022/10

 Times Cited Count:5 Percentile:63.98(Engineering, Multidisciplinary)

Journal Articles

Constraint effect on fracture behavior of underclad crack in reactor pressure vessel

Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi

Journal of Pressure Vessel Technology, 144(1), p.011304_1 - 011304_7, 2022/02

 Times Cited Count:1 Percentile:12.41(Engineering, Mechanical)

In the structural integrity assessment of a reactor pressure vessel (RPV), the fracture toughness (K$$_{Jc}$$) should be higher than the stress intensity factor at the crack tip of an under-clad crack (UCC), which is prescribed in JEAC4206-2016. However, differences in crack depth and existence of cladding between the postulated crack and fracture toughness test specimens would be affected to the plastic constraint state and K$$_{Jc}$$ evaluation. In this study, we performed fracture toughness tests and finite element analyses (FEAs) to investigate the effect of cladding on K$$_{Jc}$$ evaluation. FEA showed that the cladding decreased the plastic constraint in the UCC rather than the surface crack. Moreover, it was also found that the apparent K$$_{Jc}$$ for the UCC was higher than that for the surface crack from tests and the local approach.

Journal Articles

EXAFS studies for atomic structural change induced by ion irradiation of a reactor pressure vessel steel

Iwata, Keiko; Takamizawa, Hisashi; Ha, Yoosung; Shimodaira, Masaki; Okamoto, Yoshihiro; Honda, Mitsunori; Katsuyama, Jinya; Nishiyama, Yutaka

Nuclear Instruments and Methods in Physics Research B, 511, p.143 - 152, 2022/01

 Times Cited Count:2 Percentile:39.49(Instruments & Instrumentation)

Journal Articles

The Role of silicon on solute clustering and embrittlement in highly neutron-irradiated pressurized water reactor surveillance test specimens

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:37.09(Materials Science, Multidisciplinary)

Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 $$times$$ 10$$^{20}$$ n/cm$$^{2}$$ 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 ($$r$$) and number density ($$N_{d}$$) decreased and increased, respectively, with increasing nominal Si content. The shift in the reference temperature for nil-ductility transition ($$Delta$$RT$$_{NDT}$$) showed a good correlation with the square root of volume fraction ($$V_{f}$$) multiplied by r ($$sqrt{V_{f}times {r}}$$). 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 $$Delta$$RT$$_{NDT}$$ indicated that increasing of nominal Si content reduces the degree of embrittlement.

Journal Articles

Bayesian analysis of Japanese pressurized water reactor surveillance data for irradiation embrittlement prediction

Takamizawa, Hisashi; Nishiyama, Yutaka

Journal of Pressure Vessel Technology, 143(5), p.051502_1 - 051502_8, 2021/10

 Times Cited Count:4 Percentile:31.00(Engineering, Mechanical)

no abstracts in English

Journal Articles

Evaluation of brittle crack arrest toughness for highly-irradiated reactor pressure vessel steels

Iwata, Keiko; Hata, Kuniki; Tobita, Toru; Hirota, Takatoshi*; Takamizawa, Hisashi; Chimi, Yasuhiro; Nishiyama, Yutaka

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07

Journal Articles

Fracture toughness in postulated crack area of PTS evaluation in highly-neutron irradiated RPV steel

Ha, Yoosung; Shimodaira, Masaki; Takamizawa, Hisashi; Tobita, Toru; Katsuyama, Jinya; Nishiyama, Yutaka

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 6 Pages, 2021/07

Journal Articles

Assessment of residual stress for thick butt-welded plate of a reactor pressure vessel steel

Ha, Yoosung; Okano, Shigetaka*; Takamizawa, Hisashi; Katsuyama, Jinya; Mochizuki, Masahito*

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 6 Pages, 2021/07

Journal Articles

Grain-boundary phosphorus segregation in highly neutron-irradiated reactor pressure vessel steels and its effect on irradiation embrittlement

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:16 Percentile:90.82(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 1.2$$times$$10$$^{20}$$ n/cm$$^{2}$$ (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 1.0$$times$$10$$^{20}$$ n/cm$$^{2}$$ (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.

Journal Articles

Atomistic modeling of hardening in spinodally-decomposed Fe-Cr binary alloys

Suzudo, Tomoaki; Takamizawa, Hisashi; Nishiyama, Yutaka; Caro, A.*; Toyama, Takeshi*; Nagai, Yasuyoshi*

Journal of Nuclear Materials, 540, p.152306_1 - 152306_10, 2020/11

 Times Cited Count:11 Percentile:76.86(Materials Science, Multidisciplinary)

Spinodal decomposition in thermally aged Fe-Cr alloys leads to significant hardening, which is the direct cause of the so-called 475C-embrittlement. To illustrate how spinodal decomposition induces hardening by atomistic interactions, we conducted a series of numerical simulations as well as reference experiments. The numerical results indicated that the hardness scales linearly with the short-range order (SRO) parameter, while the experimental result reproduced this relationship within statistical error. Both seemingly suggest that neighboring Cr-Cr atomic pairs essentially cause hardening, because SRO is by definition uniquely dependent on the appearance probability of such pairs. A further numerical investigation supported this notion, as it suggests that the dominant cause of hardening is the pinning effect of dislocations passing over such Cr-Cr pairs.

Journal Articles

Constraint effect on fracture mechanics evaluation for an under-clad crack in a reactor pressure vessel steel

Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08

In JEAC 4206 which prescribes the methodology for assessing the structural integrity of reactor pressure vessels (RPVs), an under-clad crack (UCC) at the inner surface of RPV is postulated, and it is required that the fracture toughness of RPV steels is higher than stress intensity factor for at the crack tip during the pressurized thermal shock event. In the present study, to investigate the effect of cladding on the fracture toughness, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for an RPV steel containing an UCC or a surface crack, and the constraint effect for UCC was also discussed. As the result, we found that the fracture toughness for UCC was considerably higher than that for surface crack. On the other hand, the FEAs showed that the cladding decreased the constraint effect for UCC.

Journal Articles

Bayesian uncertainty evaluation of Charpy ductile-to-brittle transition temperature for reactor pressure vessel steels

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

Journal Articles

Ion-induced irradiation hardening of the weld heat-affected zone in low alloy steel

Ha, Yoosung; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi; Nishiyama, Yutaka

Nuclear Instruments and Methods in Physics Research B, 461, p.276 - 282, 2019/12

 Times Cited Count:4 Percentile:37.95(Instruments & Instrumentation)

Journal Articles

Susceptibility to neutron irradiation embrittlement of heat-affected zone of reactor pressure vessel steels

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

75 (Records 1-20 displayed on this page)