Search for Articles:
Title / Keyword
Author / Affiliation / Email
Journal
Article Type
 
 
Advanced Search
 
Section
Special Issue
Volume
Issue
Number
Page
 
4.4
2.9
Journals
Metals
Special Issues
Irradiation Response and Microstructure Characterization of Metallic Materials
share announcement

Irradiation Response and Microstructure Characterization of Metallic Materials

A special issue of Metals (ISSN 2075-4701). This special issue belongs to the section "Structural Integrity of Metals".

Deadline for manuscript submissions: 30 June 2024 | Viewed by 3401

Special Issue Editors

Prof. Dr. Chenxu Wang

E-Mail Website
Guest Editor
State Key Laboratory of Nuclear Physics and Technology, Center for Applied Physics and Technology, Peking University, Beijing 100871, China
Interests: radiation effects; structural materials; synergistic damage
Dr. Tengfei Yang

E-Mail Website
Guest Editor
College of Materials Science and Engineering, Hunan University, Changsha 410082, China
Interests: irradiation damage; nuclear materials; high-entropy alloys; defect behavior
Prof. Dr. Ning Gao

E-Mail Website
Guest Editor
Institute of Frontier and Interdisciplinary Science and Key Laboratory of Particle Physics and Particle Irradiation (MOE), Shandong University, Qingdao 266237, China
Interests: radiation effects; computer simulations; ion implantation experiments
Special Issues, Collections and Topics in MDPI journals

Special Issue Information

Dear Colleagues,

Many metallic materials are a combination of high strength and ductility, high electrical and thermal conductivities, and easy to process. Thus, metals are the most widely used structural materials in nuclear reactors, many important components consist of metallic materials, such as reactor pressure vessel (RPV), cladding, breeding blanket, etc. Neutron irradiation produces numerous defects in metallic materials, the migration and aggregation of defects result in various deleterious effects, including void swelling, hardening, element segregation, creep rupture, etc. The performances of metallic materials under irradiations play an important role in the safety and economy of nuclear reactors. In this Special Issue, we will focus on the irradiation responses of metallic materials from both theoretical and experimental investigations, and underlying mechanisms influencing defect behaviors in different metallic materials, as well as the advanced characterization approaches of irradiation effects. We also cover reviews related with current status and concept of advanced nuclear materials and general characteristics of defect behaviors in metallic materials.

This Special Issue on irradiation responses and microstructure characterization of metallic materials aims to provide a communication and discussion platform covering a broad range of up-to-date findings and progress related to irradiation-induced variations of microstructures, defect behaviors, and advanced characterization approaches of irradiation effects. Scientists working from various disciplines are invited to contribute to this cause.

The keywords of this Special Issue broadly cover examples of the great number of subtopics in this field. The volume is especially open to any innovative contributions involving advanced characterization approaches of the topics and/or subtopics.

Prof. Dr. Chenxu Wang
Dr. Tengfei Yang
Prof. Dr. Ning Gao
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. Metals is an international peer-reviewed open access monthly 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

  • metallic materials
  • irradiation damage
  • defect behaviors
  • microstructures
  • void swelling
  • irradiation-induced hardening
  • multiscale simulation

Published Papers (3 papers)

Download All Papers
Order results
Result details
Show export options expand_more Show export options expand_less
Select all
Export citation of selected articles as:

Research

11 pages, 3354 KiB  
Article
Effect of Hf Dopant on Resistance to CO Toxicity on ZrCo(110) Surface for H Adsorption
by Xianggang Kong, Rongjian Pan, Dmitrii O. Kharchenko and Lu Wu
Metals 2023, 13(12), 1973; https://doi.org/10.3390/met13121973 - 04 Dec 2023
Viewed by 750
Abstract
Co-adsorption of multi-components in ZrCo-based hydrogen storage materials can lead to a number of synergistic effects, such as the modification of adsorption sites, and further worsen the hydrogen storage capability. In this work, we explore the co-adsorption of H and CO on the [...] Read more.
Co-adsorption of multi-components in ZrCo-based hydrogen storage materials can lead to a number of synergistic effects, such as the modification of adsorption sites, and further worsen the hydrogen storage capability. In this work, we explore the co-adsorption of H and CO on the ZrCo(110) surface and find that the molecular CO can be adsorbed on the clean alloy surface and thus decrease the hydrogen storage ability of the alloy. Moreover, CO occupies the adsorption site of H and therefore prevents the adsorption and diffusion into the interior of the lattice. Fortunately, the Hf dopant reduces the number of adsorption sites of the CO molecule and inhibits the formation of carbides to a certain extent. In addition, the partial density of states (PDOS) result shows that there is almost no interaction between the s orbital of H and the s orbital of Co on the pure surface of pre-adsorbed CO, while on the Hf-doped surface of pre-adsorbed CO, the s orbital of H overlapped greatly with the s orbital of Co, indicating that Hf doping inhibits CO toxicity in the interaction between H and the surface. Hence, the doping of Hf has the effect of giving resistance to CO toxicity and is conducive to the adsorption of H. Full article
(This article belongs to the Special Issue Irradiation Response and Microstructure Characterization of Metallic Materials)
►▼ Show Figures

Figure 1

11 pages, 1999 KiB  
Article
First-Principles Study of Oxygen in ω-Zr
by Yonghao Chen, Zhixiao Liu, Dong Wang and Yi Zhao
Metals 2023, 13(6), 1042; https://doi.org/10.3390/met13061042 - 30 May 2023
Viewed by 1080
Abstract
Zirconium alloys, which are widely used as cladding materials in nuclear reactors, are prone to react with oxygen (O). Furthermore, the ω-Zr in zirconium alloys can significantly increase the strength and hardness of these alloys, but there is a lack of reports on [...] Read more.
Zirconium alloys, which are widely used as cladding materials in nuclear reactors, are prone to react with oxygen (O). Furthermore, the ω-Zr in zirconium alloys can significantly increase the strength and hardness of these alloys, but there is a lack of reports on the behavior of oxygen in ω-Zr in the current literature. To investigate their interactions, we have studied the behavior of O in ω-Zr using the first-principles approach. In this work, we examined the effects of vacancy and alloying elements (Nb, Sn) on the behavior of O in ω-Zr. The results show that O with a formation energy of −5.96 eV preferentially occupies an octahedral interstitial position in ω-Zr. A vacancy reduces the formation energy of O in a tetrahedral interstitial position in ω-Zr. Nb and Sn decrease the formation energy of O in the octahedral interstitial position by 6.16 eV and 5.08 eV. Vacancy effectively reduces the diffusion barrier of O around it, which facilitates the diffusion of O in ω-Zr. Nb and Sn preferentially occupy the 1b and 2d substitution sites in ω-Zr, respectively. Nb makes the diffusion barrier of O in ω-Zr lower and promotes the diffusion of O in ω-Zr. Moreover, Sn makes the diffusion of O around Sn difficult. It was further found that O is less prone to form clusters in ω-Zr and tends to independently occupy interstitial positions in ω-Zr. In particular, a single vacancy would make the binding energy between O atoms to be further reduced. Full article
(This article belongs to the Special Issue Irradiation Response and Microstructure Characterization of Metallic Materials)
►▼ Show Figures

Figure 1

10 pages, 3998 KiB  
Article
The Effect of Black-Dot Defects on FeCrAl Radiation Hardening
by Jian Sun, Miaosen Yu, Zhixian Wei, Hui Dai, Wenxue Ma, Yibin Dong, Yong Liu, Ning Gao and Xuelin Wang
Metals 2023, 13(3), 458; https://doi.org/10.3390/met13030458 - 22 Feb 2023
Viewed by 1111
Abstract
FeCrAl is regarded as one of the most promising cladding materials for accident-tolerant fuel at nuclear fission reactors due to its comprehensive properties of inherent corrosion resistance, excellent irradiation resistance, high-temperature oxidation resistance, and stress corrosion cracking resistance. In this work, the irradiation [...] Read more.
FeCrAl is regarded as one of the most promising cladding materials for accident-tolerant fuel at nuclear fission reactors due to its comprehensive properties of inherent corrosion resistance, excellent irradiation resistance, high-temperature oxidation resistance, and stress corrosion cracking resistance. In this work, the irradiation response of FeCrAl irradiated by 2.4 MeV He2+ ions with a fluence of 1.1 × 1016 cm−2 at room temperature was studied using X-ray diffraction, transmission electron microscopy, and nanoindentation. The characterization results of structural and mechanical properties showed that only black-dot defects exist in irradiated FeCrAl samples, and that the hardness of the irradiated samples was 11.5% higher than that of the unirradiated samples. Similar to other types of radiation defects, black-dot defects acted as fixed defect obstacles and hindered the movement of slip dislocations moving under the applied load, resulting in a significant increase in the hardness of FeCrAl. Importantly, this work points out that irradiation-induced black-dot defects can significantly affect the mechanical properties of materials, and that their contribution to radiation hardening cannot be ignored. Full article
(This article belongs to the Special Issue Irradiation Response and Microstructure Characterization of Metallic Materials)
►▼ Show Figures

Figure 1

Show export options expand_more Show export options expand_less
Select all
Export citation of selected articles as:
 
Displaying articles 1-3
Back to TopTop