Hongliang Ming
Zhiming Zhang
Jian Chen
James J. Noël- 1Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research, Chinese Academy of Sciences, Shenyang, China
- 2Institute of Corrosion Science and Technology, Guangzhou, China
- 3Department of Chemistry, The University of Western Ontario, London, ON, Canada
- 4Surface Science Western, The University of Western Ontario, London, ON, Canada
Editorial on the Research Topic
Nuclear materials degradation
Nuclear energy is regarded as one of the most efficient ways to reduce global carbon emissions, especially in China, which has made a promise to the world. Safety is the prerequisite in the development of a nuclear power plant (NPP). However, during the operation of an NPP, degradation of materials occurs and severely threatens the safety of the operation. As a result, to ensure the long-term safety of an NPP, it is of great importance to understand the degradation behavior of key materials, which is essential for nuclear safety management and indispensable for life prediction.
The aim of this Research Topic is to provide a communication platform for researchers from all over the world who are working on nuclear material degradation. Nuclear materials include both metallic and non-metallic materials used in reactors (e.g., PWR, BWR, and fast reactors). The degradation here refers to corrosion, stress corrosion cracking, fretting wear, corrosion fatigue, galvanic corrosion, flow accelerated corrosion, and other types of failure modes (Ming et al., 2019; Ming et al., 2020; Okonkwo et al., 2021; Wu et al., 2021; Wu et al., 2022; Zhang et al., 2022; Zhang et al., 2023). Both experimental and simulating methods are suitable. Additionally, high-level nuclear waste disposal and other related issues are in the scope of this Research Topic. With the effort of the Guest Editor team, the Editorial Development Manager’s great support, and the participants’ valuable contributions, this Research Topic has successfully published five high-quality peer-reviewed papers.
The article entitled “The kinetics of hydrogen peroxide reduction on rare Earth doped UO2 and SIMFUEL” by Zhu et al. investigated the electrochemical reduction of hydrogen peroxide in sodium chloride solutions containing various anions (bicarbonate/carbonate and sulphate) on Gd-UO2, Dy-UO2, and a SIMFUEL (UO2 doped to simulate spent nuclear fuel). A faster reaction was found on the SIMFUEL surface, and bicarbonate/carbonate, but not sulphate, was found to suppress peroxide reduction. In addition, the noble metal particles present in the SIMFUEL appeared to play only a minor role in the reduction process.
The article entitled “Novel photocatalytic coating for corrosion mitigation in 304LSS of dry storage canisters” by Sathasivam et al. developed a multilayered titanium dioxide (TiO2) composite coating on 304 L stainless steel (a candidate canister material for storing the spent radioactive fuels of an NPP) and studied its corrosion behavior in aerated 3.5% NaCl solutions. They found that the Ce-doped TiO2 coating exhibited a better performance in terms of photocathodic protection than the coating without cerium doping.
A Review by Zhang et al. summarized the research progress on the corrosive environment large-scale evolution for nuclear waste containers. Typical corrosion environments, focusing on the temperature, saturation, oxygen content, and radiation obtained by numerical simulation under different deep geological conditions in various countries, are presented.
Shu et al. reviewed the degradation behavior of the fuel cladding material 20Cr25NiNb for the British Advanced Gas Reactor (AGR), focusing on the long-term in-core service degradation (oxidation, carbon deposition, high-temperature creep, thermal aging, and mechanical property degradation) and the intergranular stress corrosion cracking and intergranular attack behavior during the wet storage of spent fuel.
Zhang et al. reviewed the hydrogen solubility, hydrogen diffusion coefficient, hydrogen absorption, and embrittlement behavior of two types of potential alloys, carbon steel and titanium (or its alloys), which were used as the container material of nuclear waste for deep geological disposal.
We hope the five papers in this Research Topic can provide useful information, new insight, data support, and a safety assessment of nuclear materials for readers.
Author contributions
Writing—original draft preparation, HM; writing—review and editing, ZZ, JC, and JN. All authors contributed to the article and approved the submitted version.
Funding
This work was supported by the Youth Innovation Promotion Assessment CAS (2022187).
Conflict of interest
The authors declare that the research was conducted in the absence of any commercial or financial relationships that could be construed as a potential conflict of interest.
Publisher’s note
All claims expressed in this article are solely those of the authors and do not necessarily represent those of their affiliated organizations, or those of the publisher, the editors and the reviewers. Any product that may be evaluated in this article, or claim that may be made by its manufacturer, is not guaranteed or endorsed by the publisher.
References
Ming, H., Liu, X., Lai, J., Wang, J., Gao, L., and Han, E. H. (2020). Fretting wear between Alloy 690 and 405 stainless steel in high temperature pressurized water with different normal force and displacement. J. Nucl. Mat. 529, 151930.
Ming, H., Liu, X., Yan, H., Zhang, Z., Wang, J., Gao, L., et al. (2019). Understanding the microstructure evolution of Ni-based superalloy within two different fretting wear regimes in high temperature high pressure water. Scr. Mat. 170, 111–115.
Okonkwo, B. O., Ming, H., Wang, J., Han, E. H., Rahimi, E., Davoodi, A., et al. (2021). A new method to determine the synergistic effects of area ratio and microstructure on the galvanic corrosion of LAS A508/309 L/308 L SS dissimilar metals weld. J. Mat. Sci. Technol. 78, 38–50.
Wu, B., Ming, H., Meng, F., Li, Y., He, G., Wang, J., et al. (2022). Effects of surface grinding for scratched alloy 690TT tube in PWR nuclear power plant: Microstructure and stress corrosion cracking. J. Mat. Sci. Technol. 113, 229–245.
Wu, B., Ming, H., Zhang, Z., Meng, F., Li, Y., Wang, J., et al. (2021). Effect of surface scratch depth on microstructure change and stress corrosion cracking behavior of alloy 690TT steam generator tube. Corros. Sci. 192, 109792.
Zhang, Y., Ming, H., Lai, J., Gao, L., Wang, J., and Han, E. H. (2023). Fretting wear behaviour of Zr alloy cladding tube mated with Zr alloy dimple under mixed fretting regime in simulated primary water of PWR. J. Mat. Sci. Technol. 158, 43–52.
Keywords: nuclear material, corrosion, stress corrosion cracking, irradiation, materials for advanced reactors
Citation: Ming H, Zhang Z, Chen J and Noël JJ (2023) Editorial: Nuclear materials degradation. Front. Mater. 10:1218835. doi: 10.3389/fmats.2023.1218835
Received: 08 May 2023; Accepted: 19 May 2023;
Published: 30 May 2023.
Edited and reviewed by:
Guang-Ling Song, Southern University of Science and Technology, ChinaCopyright © 2023 Ming, Zhang, Chen and Noël. This is an open-access article distributed under the terms of the Creative Commons Attribution License (CC BY). The use, distribution or reproduction in other forums is permitted, provided the original author(s) and the copyright owner(s) are credited and that the original publication in this journal is cited, in accordance with accepted academic practice. No use, distribution or reproduction is permitted which does not comply with these terms.
*Correspondence: Hongliang Ming, hlming12s@imr.ac.cn