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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

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

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

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:0 Percentile:0(Engineering, Mechanical)

no abstracts in English

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

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

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:1 Percentile:81.22(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:1 Percentile:39.17(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:0 Percentile:0.02(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

Journal Articles

Applicability of miniature compact tension specimens for fracture toughness evaluation of highly neutron irradiated reactor pressure vessel steels

Ha, Yoosung; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Nishiyama, Yutaka

Journal of Pressure Vessel Technology, 140(5), p.051402_1 - 051402_6, 2018/10

 Times Cited Count:1 Percentile:11.36(Engineering, Mechanical)

Journal Articles

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

Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Hanawa, Satoshi; Nishiyama, Yutaka

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 6 Pages, 2018/07

JAEA Reports

Confirmation tests for Warm Pre-stress (WPS) effect in reactor pressure vessel steel (Contract research)

Chimi, Yasuhiro; Iwata, Keiko; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Yoshimoto, Kentaro*; Murakami, Takeshi*; Hanawa, Satoshi; Nishiyama, Yutaka

JAEA-Research 2017-018, 122 Pages, 2018/03

JAEA-Research-2017-018.pdf:44.03MB

Warm pre-stress (WPS) effect is a phenomenon that after applying a load at a high temperature fracture does not occur in unloading during cooling, and then the fracture toughness in reloading at a lower temperature increases effectively. Engineering evaluation models to predict an apparent fracture toughness in reloading are established using experimental data with linear elasticity. However, there is a lack of data on the WPS effect for the effects of specimen size and surface crack in elastic-plastic regime. In this study, fracture toughness tests were performed after applying load-temperature histories which simulate pressurized thermal shock transients to confirm the WPS effect. The experimental results of an apparent fracture toughness tend to be lower than the predictive results using the engineering evaluation models in the case of a high degree of plastic deformation in preloading. Considering the plastic component of preloading can refine the engineering evaluation models.

Journal Articles

Microstructure analysis using X-ray absorption on heat-affected zone of reactor pressure vessel steel

Iwata, Keiko; Takamizawa, Hisashi; Ha, Yoosung; Okamoto, Yoshihiro; Shimoyama, Iwao; Honda, Mitsunori; Hanawa, Satoshi; Nishiyama, Yutaka

Photon Factory Activity Report 2017, 2 Pages, 2018/00

no abstracts in English

Journal Articles

Fracture toughness evaluation of neutron-irradiated reactor pressure vessel steel using miniature-C(T) specimens

Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Nishiyama, Yutaka

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 5 Pages, 2017/07

The applicability of miniature-C(T) (Mini-C(T)) specimens to fracture toughness evaluation was investigated for neutron-irradiated reactor pressure vessel (RPV) steel. $$T_{o}$$ value determined from irradiated Mini-C(T) specimens was in good agreement with that determined from the irradiated pre-cracked Charpy-type (PCCv) specimens. Also, the scatter of the 1T-equivalent fracture toughness values obtained from the irradiated Mini-C(T) specimens was not significantly different from that obtained from the irradiated PCCv. $$T_{o}$$ values determined from Mini-C(T) specimens agreed very well with the correlation between Charpy 41J transition temperature and $$T_{o}$$ of commercially manufactured RPV steels.

Journal Articles

Predoping effects of boron and phosphorous on arsenic diffusion along grain boundaries in polycrystalline silicon investigated by atom probe tomography

Takamizawa, Hisashi; Shimizu, Yasuo*; Inoue, Koji*; Nozawa, Yasuko*; Toyama, Takeshi*; Yano, Fumiko*; Inoue, Masao*; Nishida, Akio*; Nagai, Yasuyoshi*

Applied Physics Express, 9(10), p.106601_1 - 106601_4, 2016/10

 Times Cited Count:0 Percentile:0(Physics, Applied)

Journal Articles

Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

Takamizawa, Hisashi; Ito, Hiroto; Nishiyama, Yutaka

Journal of Nuclear Materials, 479, p.533 - 541, 2016/10

 Times Cited Count:1 Percentile:14.27(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.

Journal Articles

Bayesian nonparametric analysis of crack growth rates in irradiated austenitic stainless steels in simulated BWR environments

Chimi, Yasuhiro; Takamizawa, Hisashi; Kasahara, Shigeki*; Iwata, Keiko; Nishiyama, Yutaka

Nuclear Engineering and Design, 307, p.411 - 417, 2016/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material ($$sigma$$$$_{rm YS-irr}$$), stress intensity factor (${it K}$), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.

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