Development of a new corrosion mitigation technology using nanobubbles toward corrosion mitigation in PCV system under the influence of //-rays radiolysis (Contract research); FY2021 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tohoku University*

JAEA-Review 2022-069, 114 Pages, 2023/03

JAEA-Review-2022-069.pdf:5.91MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F), Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2020, this report summarizes the research results of the "Development of a new corrosion mitigation technology using nanobubbles toward corrosion mitigation in PCV system under the influence of //-rays radiolysis" conducted in FY2021. In this work, in order to ensure the long-term reliability of steel structures that ensure important confinement functions in the debris removal process, such as existing PCVs and newly constructed negative pressure maintenance systems and piping, corrosion phenomena in wet environments where - and -ray emitting nuclides come into contact with steel are clarified for the first time. At the same time, we will develop a new corrosion prevention technology that has excellent basic applicability to PCVs

Introduction of Cu to the inside of the crevice by chelation and its effect on crevice corrosion of Type 316L stainless steel

Aoyama, Takahito; Kato, Chiaki

Corrosion Science, 210(2), p.110850_1 - 110850_10, 2023/01

https://doi.org/10.1016/j.corsci.2022.110850

The chelated complex of Cu: [Cu(EDTA)] was used to introduce Cu from outside to the inside of crevice. The introduced Cu was expected to act as an inhibitor for the crevice corrosion on stainless steels. Crevice corrosion tests, confirmed the introduction of Cu to the inside of the crevice via electromigration of [Cu(EDTA)] was confirmed. Migrated [Cu(EDTA)] reacted with H and inhibited decrease in pH inside the crevice, where Cu was separated from [Cu(EDTA)] and suppressed active dissolution of the stainless steel.

Development of a new corrosion mitigation technology using nanobubbles toward corrosion mitigation in PCV system under the influence of //-rays radiolysis (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tohoku University*

JAEA-Review 2022-002, 85 Pages, 2022/06

JAEA-Review-2022-002.pdf:3.39MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2020, this report summarizes the research results of the "Development of a new corrosion mitigation technology using nanobubbles toward corrosion mitigation in PCV system under the influence of //-rays radiolysis" conducted in FY2020. In this work, in order to ensure the long-term reliability of steel structures that ensure important confinement functions in the debris removal process, such as existing PCVs and newly constructed negative pressure maintenance systems and piping, corrosion phenomena in wet environments where - and -ray emitting nuclides come into contact with steel are clarified for the first time. At the same time, we will develop a new corrosion prevention technology that has excellent basic applicability to PCVs and has

Calculation of appropriate oxygen concentration for corrosion inhibition of 9Cr-1Mo steel in molten lead bismuth eutectic

Komatsu, Atsushi

JAEA-Research 2021-019, 24 Pages, 2022/05

JAEA-Research-2021-019.pdf:1.53MB

In order to reduce the corrosion rate of materials in molten lead bismuth eutectic (LBE), it is important to adjust the oxygen concentration, and past reports show that the oxygen concentration is often adjusted to about 10 to 10wt%. However, it is not clearly stated what concentration is optimal, and there are some reports of severe corrosion even within this concentration range. In this study, a corrosion model considering diffusion in oxide and LBE was developed for 9Cr-1Mo steel, and the corrosion control method estimated from the corrosion model were investigated. We also tried to calculate the optimum oxygen concentration to prevent the flow blockage at the low temperature of loop environment while reducing the corrosion of 9Cr-1Mo steel in molten LBE. As a result, it was expected that the corrosion mode of 9Cr-1Mo steel in LBE could be classified into three types, dense film formation, precipitation film formation, and film dissolution, depending on the ratio of oxide film thickness to diffusion layer thickness, iron concentration in LBE, and temperature. In order to inhibit corrosion, it is important to adjust the oxygen concentration so that the conditions for dense film formation can be maintained. For this purpose, it was expected that a pre-oxidized film of more than 10m should be applied before immersion in LBE. The oxygen concentration of about 10 to 10wt% is the appropriate oxygen concentration when the oxide film has grown to some extent, and a higher oxygen concentration was expected to be required when the film is thin.

Effects of a corrosion inhibitor on the corrosion of steels under thin solution layers

Momma, Yuichiro*; Sakairi, Masatoshi*; Ueno, Fumiyoshi; Otani, Kyohei

Zairyo To Kankyo, 71(5), p.133 - 137, 2022/05

https://doi.org/10.3323/jcorr.71.133

The effect of the corrosion inhibitor on the corrosion of steel under a thin solution layer was investigated. As a result of forming a thin solution layer with a thickness of 1.0-0.2 mm on the specimen, adding a mixed solution of sodium molybdate and aluminum lactate as a corrosion inhibitor, and performing electrochemical measurement, the corrosion inhibitor suppresses the anodic reaction. And in the thin solution layer, it was suggested that the morphology of the protective layer structure by the corrosion inhibitor changed according to the amount of liquid as compared with the bulk immersion.

Polarization characteristics and evaluation of corrosion rate of stainless steel in nitric acid solution containing Np

Irisawa, Eriko; Kato, Chiaki; Yamashita, Naoki; Sano, Naruto

Zairyo To Kankyo, 71(3), p.70 - 74, 2022/03

In order to evaluate the corrosion of stainless steels used in spent nuclear fuel reprocessing facilities, the immersion corrosion tests and polarization measurements were performed using R-SUS304ULC stainless steel in nitric acid solution containing a kind of radionuclides, Np. At temperatures above 328 K, the corrosion potential was higher than that in nitric acid solution and was near the transpassive region. From the comparison between the corrosion amount calculated by the immersion corrosion tests and the polarization resistance, the values of =0.018-0.025 V were obtained as a conversion factor, and the possibility of calculating the corrosion amount from the electrochemical measurement was examined.

Optimization of dissolved hydrogen concentration for mitigating corrosive conditions of pressurised water reactor primary coolant under irradiation, 2; Evaluation of electrochemical corrosion potential

Hata, Kuniki; Hanawa, Satoshi; Chimi, Yasuhiro; Uchida, Shunsuke; Lister, D. H.*

Journal of Nuclear Science and Technology, 14 Pages, 2022/00

https://doi.org/10.1080/00223131.2022.2127955

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.

Effect of gamma-ray irradiation on corrosion of stainless steel contacted with Zeolite particle

Kato, Chiaki; Yamagishi, Isao; Sato, Tomonori; Yamamoto, Masahiro*

Zairyo To Kankyo, 70(12), p.441 - 447, 2021/12

https://doi.org/10.3323/jcorr.70.441

Zeolite particles have been used in a Cs adsorption vessel for purification of contaminated water in Fukushima Dai-ich Nuclear Power Station (1F). The used Cs adsorption vessels were kept in storage space on 1F site. The risk of localized corrosion of stainless steel used in the vessel was worried. To evaluate the risk of localized corrosion, using specially designed electrochemical testing apparatus was used under gamma-ray irradiation test. And, real size mock-up test conducted. The results showed the potential change caused by creation of HO by water radiolysis decreased by zeolite particles and the enrichment of chloride ion concentration in the vessel do not propagate during dry up procedure of Cs adsorption vessel. These data indicate the risk of localized corrosion of Cs adsorption vessel may stay at considerably low level.

Experiments of self-wastage phenomena elucidation in steam generator tube of sodium-cooled fast reactor

Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12

https://doi.org/10.3327/taesj.J19.012

Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.

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