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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

https://doi.org/10.1016/j.jnucmat.2021.153203

Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 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 () and number density ( $N_{d}$ ) decreased and increased, respectively, with increasing nominal Si content. The shift in the reference temperature for nil-ductility transition (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 RT $_{NDT}$ indicated that increasing of nominal Si content reduces the degree of embrittlement.

Journal Articles

Radiolysis effects, which should be taken into account for safe decommissioning of Fukushima-Daiichi Nuclear Power Station

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 H $_{2}$ O $_{2}$ , 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.

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

JAEA Reports

Database for corrosion under irradiation conditions (Contract research)

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.

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

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

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 1.210 $^{20}$ 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 1.010 $^{20}$ 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

Corrosion monitoring in humidity-controlled environment simulating gamma ray irradiation

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

https://doi.org/10.3323/jcorr.69.107

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 $^{circ}$ 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.

Journal Articles

A Simulation of radiolysis of chloride solutions containing ferrous ion

Hata, Kuniki; Inoue, Hiroyuki*

Journal of Nuclear Science and Technology, 56(9-10), p.842 - 850, 2019/09

https://doi.org/10.1080/00223131.2019.1585986

Times Cited Count：1 Percentile：23.13(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 $^{2+}$ were carried out. The results showed that the generation of radiolytic products (H $_{2}$ O $_{2}$ , O $_{2}$ and H $_{2}$ ) increased mainly by the addition of Fe $^{2+}$ , and a drop in the pH was caused by the hydrolysis of Fe $^{3+}$ . This pH drop enhanced the reactivity of Cl $^{-}$ with $^{.}$ OH, which induced additional generation of H $_{2}$ O $_{2}$ and O $_{2}$ . These results show that low concentrations of Cl $^{-}$ (1 10 $^{-3}$ mol/dm $^{3}$ = 35ppm) in the presence of Fe $^{2+}$ could influence the generation of H $_{2}$ O $_{2}$ and O $_{2}$ during water radiolysis. However, it is considered that these effects of Fe $^{2+}$ and low concentration of Cl $^{-}$ on water radiolysis are less important for corrosion of steels due to the low concentrations of H $_{2}$ O $_{2}$ and O $_{2}$ 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 H $_{2}$ O $_{2}$ and O $_{2}$ .

Journal Articles

Preliminary verification of water radiolysis and ECP calculation models by in-pile ECP measurements

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

Journal Articles

Influence of applied load on oxidation in the vicinity of crack tips of stainless steel under high temperature water

Kasahara, Shigeki; Chimi, Yasuhiro; Hata, Kuniki; Hanawa, Satoshi

Zairyo To Kankyo, 68(9), p.240 - 247, 2019/09

https://doi.org/10.3323/jcorr.68.240

In order to study environment assisted cracking mechanism of stainless steel under BWR primary coolant condition, effects of applied load on oxidation in the vicinity of crack tips of CT specimens were evaluated. Loaded CT specimens were immersed in an aqueous condition at 290 $^{circ}$ C as a simulated BWR coolant condition, and microstructural observation on oxide near the tips of pre-cracks was carried out. Oxide inner layers, which consisted of fine grain magnetite containing Fe and Cr were formed, and oxide outer layers consisting of large grains of Fe $_{3}$ O $_{4}$ were observed to cover the inner layers. FEM analysis of stress and strain in the loaded CT specimen suggests that both of dislocations due to localized plastic deformation and elastic strain could play important roles to accelerate inner oxide formation in the vicinity of the crack tip of the specimens.

Journal Articles

Empirical equations of crack growth rates based on data fitting of neutron irradiated stainless steel under high temperature water simulating boiling water reactor core conditions

Kasahara, Shigeki; Chimi, Yasuhiro; Hata, Kuniki; Fukuya, Koji*; Fujii, Katsuhiko*

Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), p.1345 - 1355, 2019/08

This paper describes empirical equation development of crack growth rates (CGR) in consideration of IASCC of neutron irradiated stainless steel to contribute to structural integrity assessment of BWR reactor internals. Empirical equations of CGR (da/dt) were developed based on a formula of da/dt = MK $^{n}$ , assuming that "M" and "n" tend to be saturated with increasing neutron fluence. To obtain the empirical equations for normal water chemistry (NWC) and hydrogen water chemistry (HWC) of BWR, a data fitting with least square method was applied to the datasets consisting of F, K and CGR from post irradiation examinations of neutron irradiated stainless steel under simulated NWC and HWC conditions from open literature. As a result, calculated results by the equation for NWC showed good agreement with measured CGR data, meanwhile those for HWC did not. The above difference was seemed to be attributed that CGR data obtained under HWC conditions were scattered extensively.

Journal Articles

Empirical equations for tensile properties and stress-strain curves of neutron irradiated stainless steels in LWR conditions

Fukuya, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro; Hata, Kuniki

Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), p.523 - 531, 2019/08

For structural integrity assessment on reactor internals of light water reactors, empirical equations of tensile properties as a function of neutron dose, and trend curves of stress-strain relations of neutron-irradiated austenitic stainless steels was proposed by fitting to recently developed database. The data in the database were obtained from reports of national projects in Japan and open literature, which was summarized in the form of data sheets. The empirical equations for tensile properties were formulated by using a saturation-type formulae. The equations were for CW 316 and SA 304/316 stainless steels in the temperature range of 280-350 $^{circ}$ C and the dose range up to 80 dpa. Stress-strain relation curves were reproduced based on the Swift model. Obtained calculated results by the empirical equations and stress-strain relations were reasonably well fitted to experimental data. The effects of composition and cold-working, etc. on tensile properties were discussed.

Journal Articles

Radiolysis of mixed solutions of Cl $^{-}$ and Br $^{-}$ and its effect on corrosion of a low-alloy steel

Hata, Kuniki; Inoue, Hiroyuki*; Kojima, Takao*; Kasahara, Shigeki; Hanawa, Satoshi; Ueno, Fumiyoshi; Tsukada, Takashi; Iwase, Akihiro*

Proceedings of Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia 2017 (AWC 2017) (USB Flash Drive), p.304 - 314, 2017/09

A model simulation of radiolysis of mixed solutions of NaCl and NaBr was carried out. The simulation result agreed well with the experimental result, and Br $^{-}$ played an important role in determining the amounts of products from water radiolysis. The simulation result also showed that, in highly pure NaCl solutions, the steady-state concentration of a radolytic product, H $_{2}$ O $_{2}$ , was mainly controlled by three reactions (Cl $^{-}$ + $^{.}$ OH ClOH $^{.-}$ , ClOH $^{.-}$ Cl $^{-}$ + $^{.}$ OH, and ClOH $^{.-}$ + H $^{+}$ Cl $^{.}$ + H $_{2}$ O), which indicated that accurate evaluation of the rate constants of these reactions was very important in improving the radiolysis simulation of solutions containing Cl $^{-}$ . An immersion test using a low-alloy steel, SQV2A, in the mixed solutions was also carried out under irradiation. The corrosion rate increased or decreased depending on the pH or the concentrations of the halide ions in a similar way to the change in concentration of H $_{2}$ O $_{2}$ produced from water radiolysis, which is affected by the presence of Cl $^{-}$ and Br $^{-}$ . However, at high pH values (12), the corrosion rate was almost zero even though the concentration of H $_{2}$ O $_{2}$ was high. This could be attributed to enhancement of the passivity of test specimens at higher pH values.

Journal Articles

Reactivity of an antioxidant, edaravone, with reactive oxygen species and its chemical repair properties against oxidative damage on DNA

Hata, Kuniki; Lin, M.*; Yokoya, Akinari*; Fujii, Kentaro*; Yamashita, Shinichi*; Muroya, Yusa*; Katsumura, Yosuke*

Hoshasen Kagaku (Internet), (103), p.29 - 34, 2017/04

Reactivity of edaravone (3-methyl-1-phenyl-2-pyrazolin-5-one), which is known to show high antioxidative properties, with oxidative species, such as hydroxyl radical ( ${}^{bullet}$ OH) and azide radical (N $^{bullet}$ $_{3}$ ), was investigated by a pulse radiolysis experiment, and generation behavior of edaravone radicals produced through these reactions were observed. It was shown that OH-adducts are produced by the reaction with ${}^{bullet}$ OH in contrast to the other oxidative radicals, which react with edaravone by an electron transfer reaction. Chemical repair properties of edaravone against DNA lesions produced by reactions of DNA with oxidative species were also investigated by a pulse radiolysis experiment with deoxyguanosine monophosphate (dGMP) and a -radiolysis experiment with plasmid DNA solutions. It was observed that edaravone reduced dGMP radicals just after produced in a dilute aqueous solution and inhibited some base lesions on plasmid DNA more effectively than single strand breaks. These results show that edaravone may protect living system from oxidative stress, such as radiation, by not only scavenging oxidative species but also reducing precursors of DNA leisons.

Journal Articles

Report on the stay in Norway as a secondee of OECD Halden Reactor Project

Hata, Kuniki

Hoshasen Kagaku (Internet), (103), P. 65, 2017/04

no abstracts in English

Journal Articles

Evaluation of ECP measured in in-pile environment

Hanawa, Satoshi; Uchida, Shunsuke; Hata, Kuniki; Chimi, Yasuhiro; Kasahara, Shigeki*; Nishiyama, Yutaka

Proceedings of 20th Nuclear Plant Chemistry International Conference (NPC 2016) (USB Flash Drive), 11 Pages, 2016/10

ECP is the exclusive index to evaluate corrosion condition directly at the points of interest in the mixing of neutron and -ray environment. ECP can be calculated through the combination of water radiolysis and ECP model. A water radiolysis model have been applied to experiments performed in in-pile loops in the experimental reactors and applicability was confirmed. An ECP model based on the Butler-Volmer equation was also prepared. ECP of stainless steel was measured under well controlled water chemistry condition in in-pile loop in the Halden reactor, and the model was applied to evaluate ECP measured in the Halden reactor. The measured data were well explained by the water radiolysis calculation and ECP model. Accumulation of in-pile ECP data are expected for further validation of the models.

Journal Articles

Study of irradiation effect on ECP using in-pile loops in the JMTR

Hanawa, Satoshi; Uchida, Shunsuke; Hata, Kuniki; Chimi, Yasuhiro; Kasahara, Shigeki*; Nishiyama, Yutaka

Proceedings of 20th Nuclear Plant Chemistry International Conference (NPC 2016) (USB Flash Drive), 10 Pages, 2016/10

The authors proposed and ECP evaluation model introducing irradiation-induced diffusion in the oxide layer to simulate neutron irradiation effect, and predicted with this model that ECP is started to depress from the neutron flux of about ten to the fourteenth per square meter. As the JMTR has in-pile loops applicable to water chemistry experiments, degree of irradiation effect on ECP appears in the in-pile loop was estimated by the model. Under oxygen injected condition, ECP in a capsule becomes constant along the vertical direction due to the presence of high amount of oxygen and hydrogen peroxide in a capsule. However, if neutron irradiation depress ECP, ECP in a capsule along vertical direction wouldn't become constant, and the degree to the decrement is detectable by experiments.

Journal Articles