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Advanced Characterization Techniques on Nuclear Fuels and Materials

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

Deadline for manuscript submissions: closed (20 December 2023) | Viewed by 8772

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Special Issue Editors

Dr. Yaqiao Wu

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Guest Editor

1. Micron School of Materials Science and Engineering, Boise State University, Boise, ID 83401, USA
2. Center for Advanced Energy Studies, 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

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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

Published Papers (6 papers)

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Research

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13 pages, 2899 KiB

Open AccessArticle

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

Viewed by 1088

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 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

Open AccessArticle

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

Viewed by 938

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 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

Open AccessArticle

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

Viewed by 1116

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 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

Open AccessArticle

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 921

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 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 ≤0.4T_m. 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

Open AccessArticle

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 935

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 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

Open AccessReview

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 4 | Viewed by 2773

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 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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