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Key Materials in Nuclear Reactors
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Key Materials in Nuclear Reactors

A special issue of Materials (ISSN 1996-1944). This special issue belongs to the section "Materials Physics".

Deadline for manuscript submissions: 20 May 2025 | Viewed by 6460

Special Issue Editors

Prof. Dr. Weizhong Han

E-Mail Website
Guest Editor
Center for Advancing Materials Performance from the Nanoscale, State Key Laboratory for Mechanical Behavior of Materials, Xi’an Jiaotong University, Xi’an 710049, China
Interests: radiation damage; mechanical properties; zirconium; tungsten; interface engineering; dislocation
Dr. Jing Hu

E-Mail Website
Guest Editor
China Nuclear Power Technology Research Institute, Shenzhen, China
Interests: structural and cladding materials for nuclear reactor; zirconium alloys; advanced microscopy; radiation damage; corrosion and hydrogen pickup; mechanical properties

Special Issue Information

Dear Colleagues,

Nuclear power continues to play a significant role in providing clean, reliable, and sustainable energy. However, the operation of nuclear reactors poses unique challenges related to material performance, safety, and long-term sustainability. There is a growing need to explore new materials that can withstand extreme environments, exhibit enhanced resistance to radiation damage, offer improved corrosion resistance, and provide superior mechanical properties. These advancements can result in the enhanced efficiency, safety, and longevity of nuclear reactors, paving the way for the future of nuclear energy.

This Special Issue on key materials in nuclear reactors will provide a platform on which to share cutting-edge research, exchange ideas, and foster a multidisciplinary dialogue that can lead to breakthroughs in material development, characterization techniques, and applications in the nuclear industry.

We cordially invite you to contribute to the Special Issue by presenting your research papers on nuclear materials; the topics of interest include, but are not limited to the following:

  1. Nuclear reactor materials
  2. Nuclear fuel cycle and materials
  3. Radiation effects on materials
  4. Nuclear waste management and disposal
  5. Structural integrity and material performance
  6. Radiation shielding materials
  7. Nuclear material characterization techniques
  8. Materials for fusion reactors

Thank you for considering our invitation, and we look forward to welcoming you to this Special Issue on key materials in nuclear reactors.

Prof. Dr. Weizhong Han
Dr. Jing Hu
Guest Editors

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Materials is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 2600 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • nuclear materials
  • radiation damage
  • structural integrity
  • corrosion and hydrogen pickup
  • fuel assemblies
  • core design

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Published Papers (6 papers)

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Research

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17 pages, 2957 KiB  
Article
The Influence of Rhenium Content on Helium Desorption Behavior in Tungsten–Rhenium Alloy
by Yongli Liu, Yamin Song, Ye Dong, Te Zhu, Peng Zhang, Lu Wu, Xingzhong Cao and Baoyi Wang
Materials 2024, 17(11), 2732; https://doi.org/10.3390/ma17112732 - 4 Jun 2024
Viewed by 796
Abstract
To investigate the influence of different rhenium contents on the helium desorption behavior in tungsten–rhenium alloys, pure tungsten and tungsten–rhenium alloys were irradiated with helium under the same conditions. All irradiated samples were characterized using TDS and DBS techniques. The results indicate that [...] Read more.
To investigate the influence of different rhenium contents on the helium desorption behavior in tungsten–rhenium alloys, pure tungsten and tungsten–rhenium alloys were irradiated with helium under the same conditions. All irradiated samples were characterized using TDS and DBS techniques. The results indicate that the addition of rhenium can reduce the total helium desorption quantity in tungsten–rhenium alloys and slightly accelerate the reduction in the concentration of vacancy-type defects accompanying helium dissociation. The desorption activation energy of helium is approximately 2 eV at the low-temperature peak (~785 K) and about 4 eV at the high-temperature peak (~1475 K). An increase in rhenium content causes the desorption peak to shift towards higher temperatures (>1473 K), which is attributed to the formation of the stable complex structures between rhenium and vacancies. Besides, the migration of He-vacancy complexes towards traps and dynamic annealing processes both lead to the recovery of vacancy-type defects, resulting in a decrease in the positron annihilation S parameters. Full article
(This article belongs to the Special Issue Key Materials in Nuclear Reactors)
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16 pages, 5504 KiB  
Article
Adsorption Behavior of Co2+, Ni2+, Sr2+, Cs+, and I− by Corrosion Products α-FeOOH from Typical Metal Tanks
by Yingzhe Du, Lili Li, Yukun Yuan, Yufaning Yin, Genggeng Dai, Yaqing Ren, Shiying Li and Peng Lin
Materials 2024, 17(11), 2706; https://doi.org/10.3390/ma17112706 - 3 Jun 2024
Viewed by 530
Abstract
Throughout the nuclear power production process, the disposal of radioactive waste has consistently raised concerns about environmental safety. When the metal tanks used for waste disposal are corroded, radionuclides seep into the groundwater environment and eventually into the biosphere, causing significant damage to [...] Read more.
Throughout the nuclear power production process, the disposal of radioactive waste has consistently raised concerns about environmental safety. When the metal tanks used for waste disposal are corroded, radionuclides seep into the groundwater environment and eventually into the biosphere, causing significant damage to the environment. Hence, investigating the adsorption behavior of radionuclides on the corrosion products of metal tanks used for waste disposal is an essential component of safety and evaluation protocols at disposal sites. In order to understand the adsorption behavior of important radionuclides 60Co, 59Ni, 90Sr, 135Cs and 129I on α-FeOOH, the influences of different pH values, contact time, temperature and ion concentration on the adsorption rate were studied. The adsorption mechanism was also discussed. It was revealed that the adsorption of key nuclides onto α-FeOOH is significantly influenced by both pH and temperature. This change in surface charge corresponds to alterations in the morphology of nuclide ions within the system, subsequently impacting the adsorption efficiency. Sodium ions (Na+) and chlorate ions (ClO3−) compete for coordination with nuclide ions, thereby exerting an additional influence on the adsorption process. The XPS analysis results demonstrate the formation of an internal coordination bond (Ni–O bond) between Ni2+ and iron oxide, which is adsorbed onto α-FeOOH. Full article
(This article belongs to the Special Issue Key Materials in Nuclear Reactors)
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14 pages, 4278 KiB  
Article
Study on the Thermophysical Properties of 80% 10B Enrichment of B4C
by Zhipeng Lv, Haixiang Hu, Jin Cao, Shaofang Lin, Changzheng Li, Lihong Nie, Xuanpu Zhou, Qisen Ren, Qingyang Lv and Jing Hu
Materials 2023, 16(22), 7212; https://doi.org/10.3390/ma16227212 - 17 Nov 2023
Viewed by 989
Abstract
In this paper, a specific type of Boron Carbide (B4C) with a high enrichment of 80 ± 0.3 at% 10B was prepared as an absorbing material for control rods in nuclear reactors. The enrichment of 10B was achieved using [...] Read more.
In this paper, a specific type of Boron Carbide (B4C) with a high enrichment of 80 ± 0.3 at% 10B was prepared as an absorbing material for control rods in nuclear reactors. The enrichment of 10B was achieved using a chemical exchange method, followed by obtaining boron carbide powder through a carbothermal reduction method. Finally, B4C with a high enrichment of 68.3~74.2% theoretical density was obtained using a hot-pressed sintering process. This study focused on investigating the basic out-of-pile thermophysical properties of the high enrichment B4C compared to natural B4C reference pellets under non-irradiated conditions. These properties included the thermal expansion coefficient, thermal conductivity, emissivity, elastic limit, elastic modulus, and Poisson’s ratio. The research results indicate that the enriched B4C pellet exhibits good thermal stability and meets the technical requirements for mechanical capability. It was observed that porosity plays a significant role in determining the out-of-pile mechanical capability of B4C, with higher porosity samples having a lower thermal conductivity, elastic–plastic limit, and elastic modulus. In short, all the technical indexes studied meet the requirements of nuclear-grade Boron Carbide pellets for Pressurized Water Reactors. Full article
(This article belongs to the Special Issue Key Materials in Nuclear Reactors)
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11 pages, 4019 KiB  
Article
Study on the Stability of Cu-Ni Cluster Components and the Effect of Strain on Its Structure
by Xiaochuan Zeng, Cuizhu He, Xuejun Li and Qiaodan Hu
Materials 2023, 16(21), 6952; https://doi.org/10.3390/ma16216952 - 30 Oct 2023
Viewed by 1050
Abstract
Solute clusters are one of the important mechanisms of irradiation embrittlement of ferritic steels. It is of great significance to study the stability of solute clusters in ferritic steels and their effects on the mechanical properties of the materials. Molecular dynamics was used [...] Read more.
Solute clusters are one of the important mechanisms of irradiation embrittlement of ferritic steels. It is of great significance to study the stability of solute clusters in ferritic steels and their effects on the mechanical properties of the materials. Molecular dynamics was used to study the binding energy, defect energy, and interaction energy of 2 nm-diameter Cu-Ni clusters in the ferritic lattice, which have six categories of Cu-Ni clusters, such as the pure Cu cluster, the core–shell structural cluster with one layer to four layers of Ni atoms and the pure Ni cluster. It was found that Cu-Ni clusters have lower energy advantages than pure Ni clusters. Through shear strain simulation of the three clusters, the structure of 2 nm diameter clusters does not undergo phase transformation. The number of slip systems and the length of dislocation lines in the cluster system are positively correlated with the magnitude of the critical stress of material plastic deformation. Full article
(This article belongs to the Special Issue Key Materials in Nuclear Reactors)
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Review

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28 pages, 11309 KiB  
Review
Preparation, Deformation Behavior and Irradiation Damage of Refractory Metal Single Crystals for Nuclear Applications: A Review
by Benqi Jiao, Weizhong Han, Wen Zhang, Zhongwu Hu and Jianfeng Li
Materials 2024, 17(14), 3417; https://doi.org/10.3390/ma17143417 - 10 Jul 2024
Viewed by 824
Abstract
Refractory metal single crystals have been applied in key high-temperature structural components of advanced nuclear reactor power systems, due to their excellent high-temperature properties and outstanding compatibility with nuclear fuels. Although electron beam floating zone melting and plasma arc melting techniques can prepare [...] Read more.
Refractory metal single crystals have been applied in key high-temperature structural components of advanced nuclear reactor power systems, due to their excellent high-temperature properties and outstanding compatibility with nuclear fuels. Although electron beam floating zone melting and plasma arc melting techniques can prepare large-size oriented refractory metals and their alloy single crystals, both have difficulty producing perfect defect-free single crystals because of the high-temperature gradient. The mechanical properties of refractory metal single crystals under different loads all exhibit strong temperature and crystal orientation dependence. Slip and twinning are the two basic deformation mechanisms of refractory metal single crystals, in which low temperatures or high strain rates are more likely to induce twinning. Recrystallization is always induced by the combined action of deformation and annealing, exhibiting a strong crystal orientation dependence. The irradiation hardening and neutron embrittlement appear after exposure to irradiation damage and degrade the material properties, attributed to vacancies, dislocation loops, precipitates, and other irradiation defects, hindering dislocation motion. This paper reviews the research progress of refractory metal single crystals from three aspects, preparation technology, deformation behavior, and irradiation damage, and highlights key directions for future research. Finally, future research directions are prospected to provide a reference for the design and development of refractory metal single crystals for nuclear applications. Full article
(This article belongs to the Special Issue Key Materials in Nuclear Reactors)
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19 pages, 10450 KiB  
Review
A Review of Cavitation Erosion on Pumps and Valves in Nuclear Power Plants
by Guiyan Gao, Shusheng Guo and Derui Li
Materials 2024, 17(5), 1007; https://doi.org/10.3390/ma17051007 - 22 Feb 2024
Cited by 1 | Viewed by 1565
Abstract
The cavitation erosion failure of pumps or valves induces the low efficiency and reduced service life of nuclear reactors. This paper reviews works regarding the cavitation erosion of pumps and valves in the nuclear power industry and academic research field. The cavitation erosion [...] Read more.
The cavitation erosion failure of pumps or valves induces the low efficiency and reduced service life of nuclear reactors. This paper reviews works regarding the cavitation erosion of pumps and valves in the nuclear power industry and academic research field. The cavitation erosion mechanisms of materials of pumps and valves are related to the microstructure and mechanical properties of the surface layer. The cavitation erosion resistance of austenitic stainless steel can be ten times higher than that of ferritic steel. The cavitation erosion of materials is related to the hardness, toughness, and martensitic transformation capacity. Erosion wear and erosion–corrosion research is also reviewed. Erosion wear is mainly influenced by the hardness of the material surface. Erosion–corrosion behavior is closely connected with the element composition. Measures for improving the cavitation erosion of pumps and valves are summarized in this paper. The cavitation erosion resistance of metallic materials can be enhanced by adding elements and coatings. Adhesion, inclusion content, and residual stress impact the cavitation erosion of materials with coatings. Full article
(This article belongs to the Special Issue Key Materials in Nuclear Reactors)
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