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

https://doi.org/10.1115/1.4051704

In the structural integrity assessment of a reactor pressure vessel (RPV), the fracture toughness (K) 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 evaluation. In this study, we performed fracture toughness tests and finite element analyses (FEAs) to investigate the effect of cladding on K 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 for the UCC was higher than that for the surface crack from tests and the local approach.

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

https://doi.org/10.1115/1.4050317

no abstracts in English

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

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

https://doi.org/10.1016/j.jnucmat.2020.152564

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.

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

https://doi.org/10.1115/PVP2020-21441

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.

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

https://doi.org/10.1016/j.nimb.2019.10.018

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

https://doi.org/10.1115/1.4040642

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.

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. 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. values determined from Mini-C(T) specimens agreed very well with the correlation between Charpy 41J transition temperature and of commercially manufactured RPV steels.

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

https://doi.org/10.1016/j.jnucmat.2016.07.035

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.