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Advanced Characterization Techniques on Nuclear Fuels and Materials
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Advanced Characterization Techniques on Nuclear Fuels and Materials

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

Deadline for manuscript submissions: closed (20 April 2025) | Viewed by 16929

Special Issue Editors

Prof. Dr. Yaqiao Wu

E-Mail Website
Guest Editor
Center for Advanced Energy Studies, Boise State University, 997 MK Simpson Blvd., Idaho Falls, ID 83401, USA
Interests: nanoscale structural and chemistry characterization achieved by the combination of transmission electron microscopy (TEM) and atom probe tomography (APT) techniques; material design, synthesis, property analysis; nanoscale structure and chemistry characterization; establishing critical connections between structure, chemistry and behavior at nanometric to atomic level through combinational TEM and APT techniques in a large range of various material systems from nanostructured magnetic materials, semiconductors, ceramics, quasicrystals, carbon nanotubes, and nuclear materials
Dr. Ching-Heng Shiau

E-Mail Website
Guest Editor
1. Advanced Materials Laboratory, Boise State University, Boise, ID 83401, USA
2. Center for Advanced Energy Studies, Idaho Falls, ID 83401, USA
Interests: nuclear materials for extreme environment; radiation damage; electron microscopy; mechanical properties; advanced manufacturing

Special Issue Information

Dear Colleages,

The advancement of characterization techniques drives the fundamental understanding of materials science. Innovative techniques can reveal information that cannot be observed or lost in conventional methods. The application of modern techniques provides unique microstructural and chemical information down to the atomic scale. For nuclear application materials, the extreme environment in a nuclear reactor possesses significant challenges for characterization due to the combination of stress, corrosion, high temperature, and radiation. Therefore, research on nuclear fuels and structure materials requires advanced characterization techniques to investigate the material properties and verify the material performance in a nuclear reactor. This Special Issue highlights the world-leading capabilities and application of state-of-the-art characterization and micro-scale mechanical testing techniques to explore nuclear fuels and materials, including but not limited to advanced transmission electron microscopy, atom probe tomography, focused ion beam, nanoindenter and in situ testing techniques.

Contributing papers are solicited in the following areas:

  • Nuclear fuels and structure materials;
  • Microstructure characterization on non-irradiated and irradiated materials (ion-, proton- and neutron-irradiated);
  • In situ irradiation and post-irradiation examination (PIE) utilizing advanced characterization techniques, including but not limited to transmission electron microscopy (TEM), scanning transmission electron microscopy (STEM), high-resolution TEM (HRTEM), high-resolution STEM (HRSTEM), monochromated, probe-corrected, image-corrected (S)TEM, energy-dispersive X-ray spectroscopy (EDS), electron energy loss spectroscopy (EELS), energy-filtered TEM (EFTEM), precession electron diffraction (PED), atom probe tomography (APT), scanning electron microscopy (SEM), focused ion beam (FIB), electron backscattered diffraction (EBSD), X-ray powder diffraction (XRD), etc.;
  • In situ mechanical testing combined with TEM and/or SEM techniques, e.g., picoindenter;
  • Ex situ mechanical testing techniques, e.g., nanoindentation, microhardness testing.

Dr. Yaqiao Wu
Dr. Ching-Heng Shiau
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 fuels
  • nuclear structure materials
  • irradiation
  • post-irradiation examination
  • advanced microstructure characterization techniques
  • (scanning) transmission electron microscopy ((S)TEM)
  • atom probe tomography
  • in situ/ex situ mechanical testing

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Published Papers (8 papers)

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Research

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13 pages, 3190 KiB  
Article
Assessing the Practical Constraints and Capabilities of Accelerator-Based Focused Ion Thermal Analysis
by Rijul R. Chauhan, Artur Santos Paixao, Benjamin E. Mejia Diaz, Frank A. Garner and Lin Shao
Materials 2025, 18(7), 1514; https://doi.org/10.3390/ma18071514 - 27 Mar 2025
Cited by 1 | Viewed by 308
Abstract
This study investigates the capabilities of accelerator-based Focused Ion Thermal Analysis (FITA), a remote nondestructive method developed for characterizing thermal properties using a proton beam as a localized heat source. Employing infrared (IR) imaging, FITA captures the evolution of temperature in material samples [...] Read more.
This study investigates the capabilities of accelerator-based Focused Ion Thermal Analysis (FITA), a remote nondestructive method developed for characterizing thermal properties using a proton beam as a localized heat source. Employing infrared (IR) imaging, FITA captures the evolution of temperature in material samples after the beam is deactivated, enabling precise extraction of thermal properties. However, the performance of FITA is inherently influenced by the IR camera’s resolution and frame rate, which imposes constraints on the types of materials that can be effectively analyzed. Here, a comprehensive series of finite element analysis (FEA) simulations were performed to evaluate the applicability of FITA for a wide range of materials. These simulations assess how variations in IR camera specifications impact the effectiveness of FITA in analyzing different materials. Our findings show that the current method can characterize a wide range of materials, including the majority of nuclear materials typically used in the nuclear industry. Full article
(This article belongs to the Special Issue Advanced Characterization Techniques on Nuclear Fuels and Materials)
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15 pages, 5550 KiB  
Article
Microstructure of Neutron-Irradiated Al3Hf-Al Thermal Neutron Absorber Materials
by Donna Post Guillen, Janelle Wharry, Yu Lu, Michael Wu, Jeremy Sharapov and Matthew Anderson
Materials 2025, 18(4), 833; https://doi.org/10.3390/ma18040833 - 14 Feb 2025
Cited by 1 | Viewed by 953
Abstract
A thermal neutron-absorbing metal matrix composite (MMC) comprised of Al3Hf particles in an aluminum matrix was developed to filter out thermal neutrons and create a fast flux environment for material testing in a mixed-spectrum nuclear reactor. Intermetallic Al3Hf particles [...] Read more.
A thermal neutron-absorbing metal matrix composite (MMC) comprised of Al3Hf particles in an aluminum matrix was developed to filter out thermal neutrons and create a fast flux environment for material testing in a mixed-spectrum nuclear reactor. Intermetallic Al3Hf particles capture thermal neutrons and are embedded in a highly conductive aluminum matrix that provides conductive cooling of the heat generated due to thermal neutron capture by the hafnium. These Al3Hf-Al MMCs were fabricated using powder metallurgy via hot pressing. The specimens were neutron-irradiated to between 1.12 and 5.38 dpa and temperatures ranging from 286 °C to 400 °C. The post-irradiation examination included microstructure characterization using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy. This study reports the microstructural observations of four irradiated samples and one unirradiated control sample. All the samples showed the presence of oxide at the particle–matrix interface. The irradiated specimens revealed needle-like structures that extended from the surface of the Al3Hf particles into the Al matrix. An automated segmentation tool was implemented based on a YOLO11 computer vision-based approach to identify dislocation lines and loops in TEM images of the irradiated Al-Al3Hf MMCs. This work provides insight into the microstructural stability of Al3Hf-Al MMCs under irradiation, supporting their consideration as a novel neutron absorber that enables advanced spectral tailoring. Full article
(This article belongs to the Special Issue Advanced Characterization Techniques on Nuclear Fuels and Materials)
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13 pages, 2899 KiB  
Article
Real-Time Observation of Nanoscale Kink Band Mediated Plasticity in Ion-Irradiated Graphite: An In Situ TEM Study
by Melonie P. Thomas, Ryan Schoell, Nahid Sultan Al-Mamun, Winson Kuo, John Watt, William Windes, Khalid Hattar and Aman Haque
Materials 2024, 17(4), 895; https://doi.org/10.3390/ma17040895 - 15 Feb 2024
Cited by 5 | Viewed by 1920
Abstract
Graphite IG-110 is a synthetic polycrystalline material used as a neutron moderator in reactors. Graphite is inherently brittle and is known to exhibit a further increase in brittleness due to radiation damage at room temperature. To understand the irradiation effects on pre-existing defects [...] Read more.
Graphite IG-110 is a synthetic polycrystalline material used as a neutron moderator in reactors. Graphite is inherently brittle and is known to exhibit a further increase in brittleness due to radiation damage at room temperature. To understand the irradiation effects on pre-existing defects and their overall influence on external load, micropillar compression tests were performed using in situ nanoindentation in the Transmission Electron Microscopy (TEM) for both pristine and ion-irradiated samples. While pristine specimens showed brittle and subsequent catastrophic failure, the 2.8 MeV Au2+ ion (fluence of 4.378 × 1014 cm−2) irradiated specimens sustained extensive plasticity at room temperature without failure. In situ TEM characterization showed nucleation of nanoscale kink band structures at numerous sites, where the localized plasticity appeared to close the defects and cracks while allowing large average strain. We propose that compressive mechanical stress due to dimensional change during ion irradiation transforms buckled basal layers in graphite into kink bands. The externally applied load during the micropillar tests proliferates the nucleation and motion of kink bands to accommodate the large plastic strain. The inherent non-uniformity of graphite microstructure promotes such strain localization, making kink bands the predominant mechanism behind unprecedented toughness in an otherwise brittle material. Full article
(This article belongs to the Special Issue Advanced Characterization Techniques on Nuclear Fuels and Materials)
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15 pages, 4314 KiB  
Article
Investigating Hydrogen in Zirconium Alloys by Means of Neutron Imaging
by Sarah Weick and Mirco Grosse
Materials 2024, 17(4), 781; https://doi.org/10.3390/ma17040781 - 6 Feb 2024
Cited by 2 | Viewed by 1759
Abstract
Neutrons interact with the magnetic moment of the atomic shell of an atom, as is common for X-rays, but mainly they interact directly with the nucleus. Therefore, the atomic number and the related number of electrons does not play a role in the [...] Read more.
Neutrons interact with the magnetic moment of the atomic shell of an atom, as is common for X-rays, but mainly they interact directly with the nucleus. Therefore, the atomic number and the related number of electrons does not play a role in the strength of an interaction. Instead, hydrogen that is nearly invisible for X-rays has a higher attenuation for neutrons than most of the metals, e.g., zirconium, and thus would be visible through dark contrast in neutron images. Consequently, neutron imaging is a precise, non-destructive method to quantify the amount of hydrogen in materials with low attenuation. Because nuclear fuel cladding tubes of light water reactors are made of zirconium (98%), the hydrogen amount and distribution in metallic claddings can be investigated. Even hydrogen concentrations smaller than 10 wt.ppm can be determined locally with a spatial resolution of less than 10 μm (with a high-resolution neutron microscope). All in all, neutron imaging is a very fast and precise method for several applications. This article explains the basics of neutron imaging and provides samples of investigation possibilities, e.g., for hydrogen in zirconium alloy cladding tubes or in situ investigations of hydrogen diffusion in metals. Full article
(This article belongs to the Special Issue Advanced Characterization Techniques on Nuclear Fuels and Materials)
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14 pages, 5358 KiB  
Article
Investigation of Deformation Behavior of Additively Manufactured AISI 316L Stainless Steel with In Situ Micro-Compression Testing
by Fei Teng, Ching-Heng Shiau, Cheng Sun, Robert C. O’Brien and Michael D. McMurtrey
Materials 2023, 16(17), 5980; https://doi.org/10.3390/ma16175980 - 31 Aug 2023
Cited by 2 | Viewed by 1911
Abstract
Additive manufacturing techniques are being used more and more to perform the precise fabrication of engineering components with complex geometries. The heterogeneity of additively manufactured microstructures deteriorates the mechanical integrity of products. In this paper, we printed AISI 316L stainless steel using the [...] Read more.
Additive manufacturing techniques are being used more and more to perform the precise fabrication of engineering components with complex geometries. The heterogeneity of additively manufactured microstructures deteriorates the mechanical integrity of products. In this paper, we printed AISI 316L stainless steel using the additive manufacturing technique of laser metal deposition. Both single-phase and dual-phase substructures were formed in the grain interiors. Electron backscatter diffraction and energy-dispersive X-ray spectroscopy indicate that Si, Mo, S, Cr were enriched, while Fe was depleted along the substructure boundaries. In situ micro-compression testing was performed at room temperature along the [001] orientation. The dual-phase substructures exhibited lower yield strength and higher Young’s modulus compared with single-phase substructures. Our research provides a fundamental understanding of the relationship between the microstructure and mechanical properties of additively manufactured metallic materials. The results suggest that the uneven heat treatment in the printing process could have negative impacts on the mechanical properties due to elemental segregation. Full article
(This article belongs to the Special Issue Advanced Characterization Techniques on Nuclear Fuels and Materials)
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29 pages, 15606 KiB  
Article
Thermomechanical Properties of Neutron Irradiated Al3Hf-Al Thermal Neutron Absorber Materials
by Donna Post Guillen, Mychailo B. Toloczko, Ramprashad Prabhakaran, Yuanyuan Zhu, Yu Lu and Yaqiao Wu
Materials 2023, 16(16), 5518; https://doi.org/10.3390/ma16165518 - 8 Aug 2023
Viewed by 1473
Abstract
A thermal neutron absorber material composed of Al3Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al3Hf powders at volume [...] Read more.
A thermal neutron absorber material composed of Al3Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al3Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4Tm. Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al3Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al3Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al3Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of −6.3% for the annealed irradiated 36.5 vol% specimen. Full article
(This article belongs to the Special Issue Advanced Characterization Techniques on Nuclear Fuels and Materials)
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15 pages, 4863 KiB  
Article
Nanocluster Evolution in D9 Austenitic Steel under Neutron and Proton Irradiation
by Suraj Venkateshwaran Mullurkara, Akshara Bejawada, Amrita Sen, Cheng Sun, Mukesh Bachhav and Janelle P. Wharry
Materials 2023, 16(13), 4852; https://doi.org/10.3390/ma16134852 - 6 Jul 2023
Viewed by 1676
Abstract
Austenitic stainless steel D9 is a candidate for Generation IV nuclear reactor structural materials due to its enhanced irradiation tolerance and high-temperature creep strength compared to conventional 300-series stainless steels. But, like other austenitic steels, D9 is susceptible to irradiation-induced clustering of Ni [...] Read more.
Austenitic stainless steel D9 is a candidate for Generation IV nuclear reactor structural materials due to its enhanced irradiation tolerance and high-temperature creep strength compared to conventional 300-series stainless steels. But, like other austenitic steels, D9 is susceptible to irradiation-induced clustering of Ni and Si, the mechanism for which is not well understood. This study utilizes atom probe tomography (APT) to characterize the chemistry and morphology of Ni–Si nanoclusters in D9 following neutron or proton irradiation to doses ranging from 5–9 displacements per atom (dpa) and temperatures ranging from 430–683 °C. Nanoclusters form only after neutron irradiation and exhibit classical coarsening with increasing dose and temperature. The nanoclusters have Ni3Si stoichiometry in a Ni core–Si shell structure. This core–shell structure provides insight into a potentially unique nucleation and growth mechanism—nanocluster cores may nucleate through local, spinodal-like compositional fluctuations in Ni, with subsequent growth driven by rapid Si diffusion. This study underscores how APT can shed light on an unusual irradiation-induced nanocluster nucleation mechanism active in the ubiquitous class of austenitic stainless steels. Full article
(This article belongs to the Special Issue Advanced Characterization Techniques on Nuclear Fuels and Materials)
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Review

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28 pages, 9497 KiB  
Review
Recent Progress in Gd-Containing Materials for Neutron Shielding Applications: A Review
by Kangbao Wang, Litao Ma, Chen Yang, Zeyu Bian, Dongdong Zhang, Shuai Cui, Mingliang Wang, Zhe Chen and Xianfeng Li
Materials 2023, 16(12), 4305; https://doi.org/10.3390/ma16124305 - 10 Jun 2023
Cited by 20 | Viewed by 5513
Abstract
With the rising demand for nuclear energy, the storage/transportation of radioactive nuclear by-products are critical safety issues for humans and the environment. These by-products are closely related to various nuclear radiations. In particular, neutron radiation requires specific protection by neutron shielding materials due [...] Read more.
With the rising demand for nuclear energy, the storage/transportation of radioactive nuclear by-products are critical safety issues for humans and the environment. These by-products are closely related to various nuclear radiations. In particular, neutron radiation requires specific protection by neutron shielding materials due to its high penetrating ability to cause irradiation damage. Herein, a basic overview of neutron shielding is presented. Since gadolinium (Gd) has the largest thermal neutron capture cross-section among various neutron absorbing elements, it is an ideal neutron absorber for shielding applications. In the last two decades, there have been many newly developed Gd-containing (i.e., inorganic nonmetallic-based, polymer-based, and metallic-based) shielding materials developed to attenuate and absorb the incident neutrons. On this basis, we present a comprehensive review of the design, processing methods, microstructure characteristics, mechanical properties, and neutron shielding performance of these materials in each category. Furthermore, current challenges for the development and application of shielding materials are discussed. Finally, the potential research directions are highlighted in this rapidly developing field. Full article
(This article belongs to the Special Issue Advanced Characterization Techniques on Nuclear Fuels and Materials)
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