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Sakamoto, Yoshiaki; Hata, Kuniki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 67(12), P. 719, 2025/12
no abstracts in English
Kasahara, Shigeki; Hata, Kuniki; Iwata, Keiko; Chimi, Yasuhiro
JAEA-Review 2025-024, 243 Pages, 2025/11
JAEA-Review-2025-024.pdf:27.13MB
It has been reported that intergranular stress corrosion cracking (IGSCC) has occurred near weld fusion lines of low-carbon, namely unsensitized austenitic stainless steel for pipings and reactor internals used under the primary coolant environment of boiling water reactors in Japan since the early 2000s. It becomes one of the critical technical issues clarification of the mechanism and development of countermeasure techniques for IGSCC of low-carbon-containing stainless steel. From previous research, the hardness of stainless steel is increased due to the accumulation of local strain, after expansion and contraction during welding heat input, and the increment of hardness in such heat-affected zone is recognized one of the possible material factors caused by IGSCC. In particular, for boiling water reactor internal structures, it is essential to evaluate IGSCC taking into account neutron irradiation as well as strain accumulation for hardness increase, and it is desirable to accumulate multifaceted and systematic data that can be dedicated to evaluating the irradiation effect of weld heat-affected and hardened zones. In this study, we investigated and collected irradiation data on the crack growth rates and other material properties evaluated under simulated primary water conditions of boiling water reactor environments for neutron-irradiated low-carbon stainless steel weld heat-affected zones. Those were obtained through the ENI project of the Japan Nuclear Energy Safety Organization, including data that had not been made public until now.
Shimodaira, Masaki; Ha, Yoosung; Yamaguchi, Yoshihito; Hata, Kuniki; Katsuyama, Jinya
Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 8 Pages, 2025/07
In the structural integrity assessment of the reactor pressure vessels (RPVs), the stress intensity factor acting on the tip of a postulated crack compares with fracture toughness evaluated by the fracture toughness test. The plastic constraint of the postulated crack is lower than the fracture toughness specimen due to the shallow crack depth. Assessing the structural integrity of the RPV with a low-constraint specimen may give an overly conservative result. Recently, a rational fracture assessment method has been developed for RPVs based on the local approach (LA). The LA can estimate the fracture toughness distribution using the Weibull stress, an index of the fracture independent of the plastic constraint. To apply the LA for RPV, it must be confirmed to accurately estimate the Weibull stress and the fracture toughness distribution. In this study, we focused on Weibull parameters, such as shape parameter m and the scaling factor 
, which are used to calculate the Weibull stress and the fracture probability, respectively. The effect of neutron irradiation on these parameters was investigated by conducting fracture toughness tests and finite element analyses. As a result, the m value corresponding to the uncertainty of the Weibull stress was not affected by irradiation. In contrast, it was found that the value should be optimized to accurately estimate the fracture toughness distribution for irradiated steel, considering the change in the tensile property.
Uchida, Shunsuke; Hata, Kuniki; Hanawa, Satoshi
JAEA-Data/Code 2024-003, 119 Pages, 2025/01
JAEA-Data-Code-2024-003.pdf:11.29MB
JAEA-Data-Code-2024-003-appendix(CD-ROM).zip:0.28MB
The calculation code for determining corrosive circumstance in light water reactors, WRAC-JAEA, has been developed based on water radiolysis calculation codes for BWR. The code has involved several new calculation functions to apply it for PWR, i.e., (1) high temperature pH (pH
), (2) pH effects on water radiolysis, (3) electrochemical corrosion potential (ECP) based on the mixed potential theory, and (4) ECP based on the water radiolysis calculation results. Moderation of corrosive conditions in the primary cooling systems is one of the promising procedures to mitigate the loss of reliabilities of major components in the systems, especially in aging plants. However, water chemistry control for corrosive environment mitigation procedures are much different in BWRs and PWRs. In BWRs, intergranular stress corrosion cracking (IGSCC) of stainless steel is the dominant causes for determining plant reliability. It is difficult to increase pH and injected hydrogen amounts due to direct power cycle operation. So, precise control of hydrogen injection with supported by water radiolysis and ECP analyses has been carried out to keep material reliability. In PWRs, it is possible to maintain stable control of corrosive circumstances with higher pH and sufficiently large hydrogen concentration. Recently, it was pointed out that one of the major causes of primary water stress corrosion cracking (PWSCC) of nickel alloys was hydrogen. The optimal hydrogen concentration should be evaluated to mitigate ECP without increasing hydrogen concentration. For this, a combined water radiolysis and ECP analysis code is required to determine the suitable hydrogen concentration and ECP. WRAC-JAEA can contribute not only to evaluation of corrosive conditions each of BWR and PWR, but also to prepare for suitable countermeasures for both BWR and PWR by cross-talking the knowledge and experience with assistance of the code results.
Hata, Kuniki; Hanawa, Satoshi; Chimi, Yasuhiro; Uchida, Shunsuke
Journal of Nuclear Science and Technology, 61(4), p.448 - 458, 2024/04
Times Cited Count：1 Percentile：14.77(Nuclear Science & Technology)no abstracts in English
Sato, Tomonori; Hata, Kuniki; Kato, Chiaki; Igarashi, Takahiro
Zairyo To Kankyo, 73(4), p.102 - 109, 2024/04
To evaluate the effects of dissolved oxygen concentration to water quality within SCC crack and the distribution of water quality in the depth direction under irradiation, immersion tests of stainless steel specimens given a gap and water radiolysis calculations for the water quality in the crevice gap were performed. As a result, it was confirmed that Fe
O
was formed in the entire area within the crevice regardless of the dissolved oxygen concentration. It was also estimated that under irradiation, the oxidant species produced directly by radiolysis in the crack are consumed by the oxide growth, and anion enrichment occurs in the crack even in the irradiation conditions.
Hata, Kuniki; Hanawa, Satoshi; Chimi, Yasuhiro; Uchida, Shunsuke; Lister, D. H.*
Journal of Nuclear Science and Technology, 60(8), p.867 - 880, 2023/08
Times Cited Count：2 Percentile：21.81(Nuclear Science & Technology)One of the major subjects for evaluating the corrosive conditions in the PWR primary coolant was to determine the optimal hydrogen concentration for mitigating PWSCC without any adverse effects on major structural materials. As suitable procedures for evaluating the corrosive conditions in PWR primary coolant, a couple of procedures, i.e., water radiolysis and ECP analyses, were proposed. The previous article showed the radiolysis calculation in the PWR primary coolant, which was followed by an ECP study here. The ECP analysis, a couple of a mixed potential model and an oxide layer growth model, was developed originally for BWR conditions, which was extended to PWR conditions with adding Li
(Na) and H effects on the anodic polarization curves. As a result of comparison of the calculated results with INCA in-pile-loop experiment data as well as other experimental data, it was confirmed that the ECPs calculated with the coupled analyses agreed with the measured within 
100mV discrepancies.
Hata, Kuniki; Uchida, Shunsuke; Hanawa, Satoshi; Chimi, Yasuhiro; Sato, Tomonori
Proceedings of 21st International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), 14 Pages, 2023/08
Hata, Kuniki; Kimura, Atsushi*; Taguchi, Mitsumasa*; Sato, Tomonori; Kato, Chiaki; Watanabe, Yutaka*
Zairyo To Kankyo, 72(4), p.126 - 130, 2023/04
Gamma-radiolysis experiments with gas-liquid coexistent samples were carried out to investigate effects of gas-phase radiolysis on corrosive environment for materials in solutions under irradiation. After gamma-ray irradiation, hydrogen peroxide, nitrate ion, nitrite ion were detected in the liquid phase. The production yields of nitrate ion and nitrite ion increased with increasing gas-phase volume and oxygen concentration. This result indicated that chemical reactions including oxygen and nitrogen in the gas phase were required for the production of nitrate ion and nitrite ion. To magnify the effects of gas-phase radiolysis in the gas-liquid coexistent samples, absorption dose rate in the liquid phase was reduced by one-hundredth using lead shield. The concentration of hydrogen peroxide and the pH in the shielded liquid phase were similar to those in the irradiated pure water, which did not contact with gas phase. This result indicated that the effects of nitrate ion and nitrite ion dissolved in the liquid phase on water radiolysis were not important in the current experimental system, in which the effects of gas-phase radiolysis were increased by 100-times.
Hata, Kuniki; Sato, Tomonori
Hoshasen Kagaku (Internet), (114), p.33 - 38, 2022/10
no abstracts in English
Sato, Tomonori; Hata, Kuniki; Kaji, Yoshiyuki; Taguchi, Mitsumasa*; Seito, Hajime*; Inoue, Hiroyuki*; Tada, Eiji*; Abe, Hiroshi*; Akiyama, Eiji*; Suzuki, Shunichi*
Isotope News, (782), p.40 - 44, 2022/08
The stagnant water in the reactor building at Fukushima Daiichi Nuclear Power Station (1F) is exposed to the radiation from fuel debris and radioactive species. This water contains much amounts of impurities from the seawater which was injected in the emergency cooling. The impurities will affect the radiolysis and corrosive conditions in the water under irradiation. So, the water radiolysis data, corrosion data of steels under irradiations, and the evaluated potential impacts of corrosion in the decommissioning process of 1F are arranged as the database for corrosion under irradiation. This paper introduces the outline of this database.
Hata, Kuniki; Uchida, Shunsuke; Hanawa, Satoshi; Chimi, Yasuhiro
Proceedings of International Symposium on Contribution of Materials Investigations and Operating Experience to LWRs' Safety, Performance and Reliability (Fontevraud 10) (Internet), 11 Pages, 2022/00
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：9 Percentile：64.41(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.
Hata, Kuniki
Zairyo To Kankyo, 70(12), p.468 - 473, 2021/12
In order to estimate corrosive environment in the contaminated water at Fukushima Daiichi Nuclear Power Station, effects of oxidants, such as HO, which were generated from water radiolysis, should be taken into account due to the irradiation field in the reactor building. The process of water radiolysis and the amounts of these oxidants can change depending on the conditions of water and types of radiation. After the accident, a variety of factors, which can affect water radiolysis, such as seawater constituents, surface of oxides, and 
-radionuclides, had been discussed. In this paper, these effects on radiolysis are reviewed for the better understanding of the corrosive environment in the contaminated water.
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
Sato, Tomonori; Hata, Kuniki; Kaji, Yoshiyuki; Ueno, Fumiyoshi; Inoue, Hiroyuki*; Taguchi, Mitsumasa*; Seito, Hajime*; Tada, Eiji*; Abe, Hiroshi*; Akiyama, Eiji*; et al.
JAEA-Review 2021-001, 123 Pages, 2021/06
JAEA-Review-2021-001.pdf:10.33MB
In the implement of the decommissioning of Fukushima Daiichi Nuclear Power Station (1F), there are many problems to be solved. Specially, the mitigation of the aging degradation by the corrosion of the structural materials is important to implement the decommissioning safely and continuously. However, there are limited data for the environmental factors of corrosion in 1F, and the condition of 1F is continuously changing. So, the literature data for the water radiolysis and the corrosion under irradiation are listed as the database of corrosion under irradiation in this report. And the new obtained radiolysis and corrosion data, which have not been reported in the literature and will be required in the decommissioning of 1F, are reported.
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：18 Percentile：86.07(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.
Omori, Atsushi*; Akiyama, Eiji*; Abe, Hiroshi*; Hata, Kuniki; Sato, Tomonori; Kaji, Yoshiyuki; Inoue, Hiroyuki*; Taguchi, Mitsumasa*; Seito, Hajime*; Tada, Eiji*; et al.
Zairyo To Kankyo, 69(4), p.107 - 111, 2020/04
To evaluate the effect of oxidants, which are formed by radiolysis of water under gamma ray irradiation, on the corrosion of a carbon steel in humid environment, ozone was introduced as a model oxidant in to humidity-controlled air at 50
C in a thermo-hygrostat chamber. Corrosion monitoring was performed by using an Atmospheric Corrosion Monitor-type (ACM) sensor consisting of a carbon steel anode and an Ag cathode. The output current of the ACM sensor was increased with the increase in relative humidity and it was obviously increased with the increase in the introduced ozone concentration at each relative humidity. The results indicate that ozone accelerates the corrosion of the carbon steel. The effect of ozone on the corrosion acceleration is attributed to the fast reduction reaction and fast dissolution reaction in to water compared to that of oxygen.
Hata, Kuniki; Inoue, Hiroyuki*
Journal of Nuclear Science and Technology, 56(9-10), p.842 - 850, 2019/09
Times Cited Count：2 Percentile：15.62(Nuclear Science & Technology)To investigate the effect of dissolved species from steels on the radiolysis processes of Cl
, radiolysis simulations of solutions containing both Cl and Fe
were carried out. The results showed that the generation of radiolytic products (HO, O and H) increased mainly by the addition of Fe, and a drop in the pH was caused by the hydrolysis of Fe
. This pH drop enhanced the reactivity of Cl with 
OH, which induced additional generation of HO and O. These results show that low concentrations of Cl (1 10
mol/dm
= 35ppm) in the presence of Fe could influence the generation of HO and O during water radiolysis. However, it is considered that these effects of Fe and low concentration of Cl on water radiolysis are less important for corrosion of steels due to the low concentrations of HO and O generated. The other process, such as dissolution of iron enhanced by FeOOH, might predominantly induce corrosion under the conditions of solutions with low concentrations of HO and O.
Hanawa, Satoshi; Hata, Kuniki; Chimi, Yasuhiro; Kasahara, Shigeki
Proceedings of 21st International Conference on Water Chemistry in Nuclear Reactor Systems (Internet), 12 Pages, 2019/09
