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        <title>Frontiers in Nuclear Engineering | Nuclear Materials section | New and Recent Articles</title>
        <link>https://www.frontiersin.org/journals/nuclear-engineering/sections/nuclear-materials</link>
        <description>RSS Feed for Nuclear Materials section in the Frontiers in Nuclear Engineering journal | New and Recent Articles</description>
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        <pubDate>2026-05-03T15:28:31.667+00:00</pubDate>
        <ttl>60</ttl>
        <item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2026.1823864</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2026.1823864</link>
        <title><![CDATA[Editorial: Analytical methods in nuclear forensics]]></title>
        <pubdate>2026-04-01T00:00:00Z</pubdate>
        <category>Editorial</category>
        <author>Kattathu J. Mathew</author><author>Amy E. Hixon</author><author>Matthew Higginson</author>
        <description></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2026.1757155</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2026.1757155</link>
        <title><![CDATA[Shielding design for IECF devices: ensuring safety through material analysis]]></title>
        <pubdate>2026-03-16T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Edward Martin</author><author>Thomas B. Scott</author><author>Mahmoud Bakr</author>
        <description><![CDATA[Inertial Electrostatic Confinement Fusion (IECF) devices have garnered increased recognition for their potential as compact, portable neutron sources for use in the generation of medical isotopes and various neutron-based interrogation techniques. This study investigates the shielding requirements for a laboratory enclosure that houses a deuterium-deuterium (DD) fueled IECF device. The simulation framework was first validated by reproducing the ambient dose equivalent, H*(10), conversion coefficient reference values, confirming the suitability of Geant4 for dose deposition measurements. Simulations were then used to evaluate neutron moderation and removal in various shielding materials, investigating five different concrete constituents and water, with the performance assessed relative to UK Ionising Radiations Regulations 2017 (IRR17) dose-rate limits. Neutron and gamma fluences were tallied in defined volumes of 40cm3, H*(10) measurements were then calculated using ICRP 74 conversion coefficients. The simulation results show that an IECF device can be operated safely at a neutron rate of 1×105ns−1 within public dose limits, recording <500nSvh−1 at a distance of 1.6m, with 10cm of all concretes tested other than Barite (Heavy) concrete, which measured 501nSvh−1. The results also show that a safe environment for radiation workers can be constructed, allowing the device to be operated at 1×107ns−1 if 30cm of water or ordinary concrete (OC1) is employed. The findings contribute to the understanding of optimal shielding configurations required to mitigate neutron radiation from IECF devices. Ensuring adherence to regulatory safety standards is paramount in the deployment of these fusion devices within populated areas. This research underscores the importance of selecting appropriate materials and thicknesses to achieve effective radiation protection, thereby facilitating the safe operation of IECF devices and contributing to advancements in medical isotope production and neutron-based interrogation technologies.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1683702</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1683702</link>
        <title><![CDATA[Assessment of structural materials in compact fusion reactor design]]></title>
        <pubdate>2025-11-07T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Davide Pettinari</author><author>Samuele Meschini </author><author>Raffaella Testoni </author>
        <description><![CDATA[The development of fusion energy systems demands structural components capable of withstanding extreme operational conditions, including intense neutron fluxes, high thermal and mechanical loads, and stringent requirements on neutron activation. Several structural materials have been proposed, such as nickel-based superalloys, reduced activation ferritic/martensitic steels, oxide-dispersion-strengthened alloys, SiC/SiC ceramic matrix composites, and vanadium-based alloys. While those materials have been extensively analysed for large tokamaks, no comparative studies exist on compact tokamaks. This work addresses this gap by considering an ARC-class tokamak as representative of compact design. The materials are evaluated based on the following criteria: power density deposition, absorption rate, TBR, energy multiplication factor within the breeding blanket, and displacement per atom. Numerical simulations were performed using the OpenMC Monte Carlo particle transport code to evaluate the neutronic behavior and activation characteristics of the selected structural materials. A simplified compact reactor model was developed using Constructive Solid Geometry (CSG) to enable consistent and reproducible comparisons. ODS steels and vanadium-based alloys emerged as the most promising candidates for application in compact, high-temperature fusion devices. ODS steels combine low activation with favorable performance across all evaluated metrics, offering a balanced tritium breeding capability alongside good resistance to radiation damage. Vanadium-based alloys, in turn, exhibit very low hydrogen and helium production, minimal power density deposition, facilitating heat removal from the structural material, and activation levels significantly lower than those of conventional austenitic steels. Across all materials, the simulations predict TBR values in the range of 0.90–1.25, energy multiplication factors of between 1.12 and 1.18, and first structural layer power densities of over 7 MW/m3. In the most favourable cases, the shutdown dose rates fall below natural background levels in less than 50 years.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1634367</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1634367</link>
        <title><![CDATA[Nuclear forensic analysis of thorium materials: recreation of a legacy processing method]]></title>
        <pubdate>2025-09-23T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Erin Holland</author><author>Matthew A. Higginson</author><author>Philip Kaye</author><author>Thomas B. Scott</author><author>Tomas Martin</author><author>Christopher R. D. Gilligan</author><author>Karen Kennedy</author><author>Samuel Cross</author><author>Christopher Brook</author>
        <description><![CDATA[Nuclear forensic science aims to correlate measurable parameters to the processing history of nuclear materials to support law enforcement investigations. Controlled studies on elemental fractionation with processing are valued on materials of known provenance to validate methods and signatures. There is need to understand how useful current applied techniques are when applied to thorium materials. In this study, we discuss the potential nuclear forensic signatures in thorium materials and report an academic study processing a monazite ore of known provenance through a historic industrial process to thorium dioxide. The measurements traced a variety of ‘fingerprint’ material properties and impurities through the processing route. It was shown that radiometric methods, relative rare earth element abundances, impurities, radiochronometry and microscopy were useful for characterising the material.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1655503</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1655503</link>
        <title><![CDATA[Optimizing laser powder directed energy deposition for Grade-91 and Grade-92 ferritic/martensitic steels for nuclear applications: linking process parameters to microstructure]]></title>
        <pubdate>2025-09-16T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Asif Mahmud</author><author>Subhashish Meher</author><author>Peter Renner</author><author>Ariel Rieffer</author><author>Chinthaka Silva</author><author>John Snitzer</author><author>Qianwen Zhang</author><author>Xiaoyuan Lou</author><author>Isabella Van Rooyen</author>
        <description><![CDATA[The nuclear industry is increasingly acknowledging the advantages of additive manufacturing (AM) due to its improved design flexibility and reduced manufacturing steps for producing complex engineering components. This study demonstrates the successful fabrication of nearly fully dense, nuclear-grade Grade-91 and, for the first time, Grade-92 Ferritic/Martensitic (F/M) steels via laser powder directed energy deposition (DED). Through rigorous process optimization, specifically tailoring laser power and scan speed, relative densities exceeding 99.8% were achieved in deposited 10 ×10×12 mm3 blocks, yielding exceptional build quality. The resulting microstructures exhibited a characteristic lath martensite morphology, indicative of the rapid solidification inherent to the DED process. While both alloys showed this general microstructure, the addition of tungsten (W), slightly higher carbon content, and higher geometrically necessary dislocation (GND) density in Grade-92 significantly influences mechanical properties, evidenced by a substantial increase in Vickers hardness (425 ± 12 HV) compared to Grade-91 (386 ± 14 HV). Estimated yield strengths, derived from hardness measurements, were 1063 MPa and 1195 MPa for Grade-91 and Grade-92, respectively. These findings suggest DED as a viable and promising route for manufacturing high-performance F/M steel components tailored for demanding nuclear applications, paving the way for improved reactor designs and enhanced operational efficiency.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1654123</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1654123</link>
        <title><![CDATA[Applications of ultra-high resolution microcalorimeter gamma-ray spectrometry]]></title>
        <pubdate>2025-09-11T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Katherine A. Schreiber</author><author>Katrina E. Koehler</author><author>Mark P. Croce</author><author>Emily N. Stark</author><author>Daniel G. McNeel</author><author>Matthew H. Carpenter</author><author>David J. Mercer</author><author>E. Paige Abel</author><author>Brian Archambault</author><author>Leah Arrigo</author><author>Grey Batie</author><author>Daniel T. Becker</author><author>Douglas A. Bennett</author><author>Brian M. Bucher</author><author>Stefania Dede</author><author>Joseph Fowler</author><author>Johnathon D. Gard</author><author>David Glasgow</author><author>K. C. Goetz</author><author>Craig Gray</author><author>Catalin Harabagiu</author><author>Jianwei Hu</author><author>Mark W. Keller</author><author>J. A. Ben Mates</author><author>Christine Mathew</author><author>Galen C. O’Neil</author><author>Nathan J. Ortiz</author><author>Luca Pagani</author><author>Bruce D. Pierson</author><author>Daniel R. Schmidt</author><author>Rico U. Schoenemann</author><author>Edward Seabury</author><author>Daniel S. Swetz</author><author>Joel N. Ullom</author><author>Sophie L. Weidenbenner</author><author>Ammon N. Williams</author>
        <description><![CDATA[Ultra-high energy resolution microcalorimeter gamma-ray spectroscopy—with energy resolution 5 to 10 times better than observed in spectra obtained by commercial-off-the-shelf high purity germanium detectors—is an enabling technology for ultra-precise isotope identification and quantification. Microcalorimeter gamma spectroscopy complements measurements requiring high-accuracy mass spectrometry, a costly, destructive analysis technique, and may offer benefits over mass spectrometry in the future. Microcalorimeter detectors are fabricated from superconducting materials and operate at ultra-low temperatures (<0.1 K), properties which permit measurement of spectra with peak full width half maximum (FWHM) of less than 100 eV at 100 keV. The microcalorimeter collaboration between Los Alamos National Laboratory, National Institute of Standards and Technology, and University of Colorado, Boulder has deployed three microcalorimeter gamma-ray spectrometers to nuclear facilities and analytical laboratories so far. These are the Spectrometer Optimized for Facility Integrated Applications (SOFIA), a portable system that can be moved to any facility, and two instruments called the High Efficiency and Resolution Microcalorimeter Spectrometers (HERMES) intended for permanent installation at Idaho National Laboratory and Pacific Northwest National Laboratory. Each spectrometer was customized to satisfy requirements for their specific applications. This work describes samples examined by microcalorimeter gamma-ray spectrometers, including recently irradiated materials, nuclear material from various stages of the fuel cycle, and medical isotope products. It also highlights useful signatures from actinide and fission product gamma-rays that are otherwise infeasible to observe or use for analysis without costly chemical separations and mass spectrometric assay. Microcalorimeter technology provides additional spectral signatures to existing techniques to better constrain the origin and intended use of nuclear and radioactive materials.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1621780</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1621780</link>
        <title><![CDATA[Hyperspectral x-ray imaging mapping capabilities for nuclear forensics]]></title>
        <pubdate>2025-09-03T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Matthew H. Carpenter</author><author>Emily N. Stark</author><author>Daniel G. McNeel</author><author>Stefania Dede</author><author>Christopher J. Godt</author><author>Eli S. R. Kurtz</author><author>Daniel E. Jackson</author><author>Travis J. Tenner</author><author>Benjamin E. Naes</author><author>Kimberly N. Wurth</author><author>Gregory L. Wagner</author><author>Mark P. Croce</author>
        <description><![CDATA[Nuclear forensics relies on the integration of complementary signatures to constrain the origins and history of materials. Outcomes benefit from the timeliness and precision of the disparate methods that form typical analysis chains. Sample forms are often either minute in quantity or contain signatures like morphology or composition heterogeneity encoded on a microscale, so many analysis techniques focus on resolving signatures on ever-smaller length scales. The new hyperspectral x-ray imaging (HXI) instrument developed at Los Alamos National Laboratory seeks to improve the information available from scanning electron microscopy (SEM) x-ray spectrum analysis through superior spectral energy resolution vs. typical energy dispersive spectroscopy (EDS) systems in common use in nuclear forensics and other microanalysis fields. Based on arrays of transition-edge sensor (TES) microcalorimeter detectors, this instrument achieves a typical energy resolution of 7 eV full-width at half-maximum (FWHM) at 2 keV, opening new possibilities in trace element detection/analysis and chemical state determination through spectral shape shifts. We present here some of the first applications of the HXI instrument to actinide samples and discuss potential maturation of this nascent technology for future analysis pipelines.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1603437</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1603437</link>
        <title><![CDATA[Role of additive manufacturing in developing functionally graded materials for nuclear applications]]></title>
        <pubdate>2025-08-21T00:00:00Z</pubdate>
        <category>Review</category>
        <author>Amal Sasi</author><author>Madhulika Srivastava</author><author>K. Dash</author>
        <description><![CDATA[The global acceptance of additive manufacturing has evolved with time and has proven to provide promising solutions to varied critical requirements of the nuclear industry. The components of a nuclear reactor, when built using additive manufacturing techniques, offer high microstructural control, making them versatile for a range of properties. These properties can be made easily achievable and tailorable by using functionally graded materials. The nuclear components with a wide range of properties are essential, as the environment inside and outside the reactor varies drastically. This study reviews the current progress in additive manufacturing techniques used for manufacturing functionally graded materials for nuclear applications, highlighting the gradient design methodologies and processing techniques. Additive Manufacturing techniques such as selective laser melting uses multiple powder feeders, and mechanical pre-mixing of powders along with controlled process parameters for effectively fabricating functionally graded materials. These materials possess superior mechanical properties (such as microhardness ranging up to 890 H00.5 and compressive strength up to 2040 MPa for FeCrCoNiMo0.5W0.75), thermal conductivity and thermal properties compared to monolithic counterparts. A comparative analysis of the manufacturing capabilities of the additive manufacturing techniques, along with the usage of advanced computational techniques such as AI in optimising process parameters for desirable strength and low defect generation, is also presented. The study emphasises on the need for strategies such as process parameters optimisation and data-driven design to fully utilise the potential of additively manufactured functionally graded materials in the nuclear sector.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1639874</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1639874</link>
        <title><![CDATA[Hydrodynamic characterization of the redox chemistry of crown-encapsulated uranyl complexes]]></title>
        <pubdate>2025-07-24T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Alexander C. Ervin</author><author>James D. Blakemore</author>
        <description><![CDATA[The redox properties of actinide-containing species strongly influence their reactivity, speciation, and interfacial behavior, but the experimental quantification of the electrochemical characteristics of molecular actinide complexes in nonaqueous media has not received the attention it deserves. Here, results from hydrodynamic methods and electrochemical simulations of U(VI)/U(V) redox are reported, including quantification of heterogeneous electron-transfer kinetics and estimation of chemical reversibility of U(VI)/U(V) interconversion at electrodes in acetonitrile-based electrolyte. The complexes investigated are recently reported U(VI) and U(V) complexes in which the uranyl ion (UO2n+) is encapsulated in a macrocyclic 18-crown-6-like moiety templated by a Pt(II) center. These complexes feature the most positive value UVI/UV reduction potential yet reported and are thus particularly relevant to study of facile U(V) generation from U(VI) precursors as well as uranium electroanalysis. Rotating disk electrode (RDE) studies have been used to quantify the diffusion coefficients of the U(VI) and U(V) complexes, and standard heterogeneous electron transfer rate constants (k0) for the redox have been determined using a conventional Koutecký-Levich analysis. Rotating ring-disk electrode (RRDE) studies have been used to directly interrogate the chemical reversibility of U(VI)-U(V) interconversion, confirming that reduction of the U(VI) complex at an Au disk is associated with formation of the U(V) analogue that can be readily re-oxidized at a Pt ring under hydrodynamic (rotating) conditions. Because measurements of the type reported here are generally associated with current flows that are larger than those found in corresponding quiescent (unstirred) conditions, our findings suggest that hydrodynamic methods could be advantageous for design of electroanalytical approaches to detection of actinide species and study of their redox properties.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1619584</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1619584</link>
        <title><![CDATA[Editorial: Structures and properties of fluorite-related systems for nuclear applications]]></title>
        <pubdate>2025-06-05T00:00:00Z</pubdate>
        <category>Editorial</category>
        <author>Gianguido Baldinozzi</author><author>Thierry Wiss</author>
        <description></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1569103</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1569103</link>
        <title><![CDATA[Uranium fluoride micromaterials: a new frontier in nuclear engineering]]></title>
        <pubdate>2025-03-25T00:00:00Z</pubdate>
        <category>Perspective</category>
        <author>Harry Jang</author><author>Frederic Poineau</author>
        <description><![CDATA[This perspective explores recent advancements in the synthesis and application of uranium fluoride micromaterials, emphasizing their role in the nuclear industry. Uranium micromaterials, including oxides, fluorides, and carbides, are crucial for applications ranging from high-temperature gas-cooled reactors to nuclear forensics and medical isotope production. The perspective highlights a novel chemical transformation process for synthesizing uranium fluoride micromaterials, in which uranium oxides are fluorinated in an autoclave using HF gas (generated from the decomposition of silver bifluoride) or ammonium bifluoride while preserving their original morphologies. This transformation produces various uranium fluoride microstructures, including UF4, UO2F2, and (NH4)3UO2F5, in the form of microrods, microplates, and microspheres. The perspective discusses challenges in maintaining controlled morphologies during fluorination and explores future directions, such as the synthesis of actinide fluoride micromaterials and the development of uranium chloride and other uranium compounds. The continued advancement of these materials holds significant potential for innovations in nuclear fuel cycles, actinide material chemistry, and nuclear forensics.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1495360</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1495360</link>
        <title><![CDATA[Impact of alpha-damage and helium production on the heat capacity of actinide oxides]]></title>
        <pubdate>2025-02-24T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Thierry Wiss</author><author>Rudy J. M. Konings</author><author>Dragos Staicu</author><author>Alessandro Benedetti</author><author>Jean-Yves Colle</author><author>Vincenzo V. Rondinella</author><author>Emilio Maugeri</author><author>Zeynep Talip</author><author>Arne Janssen</author><author>Oliver Dieste</author><author>Luana Cognini</author><author>Emanuele De Bona</author><author>Gianguido Baldinozzi</author><author>Christine Guéneau</author>
        <description><![CDATA[The heat capacity of alpha-damaged uranium, plutonium, and americium mixed dioxide (Uu, Puv, Amw)O2±x samples was measured during thermal annealing. The excess of heat released was assessed and the recovery stages associated with various defects described by integrating results from transmission electron microscopy, helium desorption spectroscopy, thermal diffusivity, and XRD annealing studies. It is shown that different defect-annealing stages could be singled out. It could also be evidenced that the excess of energy stored in defects tends to saturate after rather low damage levels, but that, with increasing radiogenic helium production, another contribution of stored energy appears which can be attributed to the formation of He-defect complexes that cannot be annihilated until higher temperatures are reached.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2025.1534820</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2025.1534820</link>
        <title><![CDATA[Effect of microstructure and neutron irradiation defects on deuterium retention in SiC]]></title>
        <pubdate>2025-02-10T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Alex Leide</author><author>Weicheng Zhong</author><author>Isabel Fernandez-Victorio</author><author>Duc Nguyen-Manh</author><author>Takaaki Koyanagi</author>
        <description><![CDATA[Retention of hydrogen isotopes is a critical concern for operating fusion reactors as retained tritium both activates components and removes scarce fuel from the fuel cycle. Radiation-induced displacement damage in SiC influences the retention of hydrogen isotopes compared to pristine SiC. Deuterium retention in neutron irradiated high purity SiC has been compared to different microstructures of non-irradiated high purity SiC using thermal desorption spectroscopy after gas charging and low energy ion implantation. Experimental results show lower deuterium retention in single crystal SiC than in polycrystal SiC indicating that grain boundaries are key trapping features in unirradiated SiC. Deuterium is released at lower temperatures in neutron irradiated polycrystal SiC compared to pristine polycrystal SiC, suggesting weaker trapping by radiation-induced defects compared to grain boundary trapping sites in the pristine materials. Low energy ion implantation caused a high deuterium release temperature, highlighting the sensitivity of deuterium release behaviour to radiation defect characteristics. First principles calculations have been conducted to identify energetically favourable trapping sites in SiC at the HABcVSi and HTSiVC complexes, and migration barriers between interstitial sites. This helps interpret experimental results and derive effective diffusivity of hydrogen isotopes in SiC in the presence of vacancies.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2024.1487828</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2024.1487828</link>
        <title><![CDATA[Liquidus curve of uranium–plutonium mixed oxide (MOX) system]]></title>
        <pubdate>2025-01-07T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Leonid Burakovsky</author><author>Dean L. Preston</author><author>Andrew A. Green</author>
        <description><![CDATA[Mixed oxides of uranium and plutonium (MOX) are currently considered reference fuels for the new generation of fast breeder reactors such as ASTRID. The key factor determining the performance and safety of fuel such as MOX is its operational limits in applied practice, which are closely related to the material’s structure and thermodynamic stability. They are, in turn, closely related to the ambient (zero pressure) melting point (Tm); thus, Tm is an important engineering parameter. However, the current knowledge of Tm of MOX is limited and controversial, as several reported studies do not converge on the unique behavior of Tm as a function of x. In this study, we present a theoretical model for the melting curve (liquidus) of a mixture and apply it to MOX considered a mixture of pure UO2 and PuO2. The model uses the known melting curves of pure constituents as an input and predicts the melting curve of their mixture. It has only one free parameter, which must be determined independently. In the case of MOX, Tm of MOX as a function of x as given by the model has a local minimum at x≈0.64, which disagrees slightly with our previous ab initio molecular dynamics studies that place this minimum at x=0.7.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2024.1544499</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2024.1544499</link>
        <title><![CDATA[Editorial: Applications of spectroscopy and chemometrics in nuclear materials analysis]]></title>
        <pubdate>2025-01-07T00:00:00Z</pubdate>
        <category>Editorial</category>
        <author>Luke R. Sadergaski</author><author>Robert J. Lascola</author><author>Kyle C. Hartig</author>
        <description></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2024.1486694</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2024.1486694</link>
        <title><![CDATA[Microstructure, electrical resistivity, and tensile properties of neutron-irradiated Cu–Cr–Nb–Zr]]></title>
        <pubdate>2024-11-12T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Alice Perrin</author><author>Dai Hamaguchi</author><author>Josina W. Geringer</author><author>Steve Zinkle</author><author>Ying Yang</author><author>Steve Skutnik</author><author>Jon Poplawsky</author><author>Yutai Katoh</author>
        <description><![CDATA[High strength, high conductivity copper alloys that can resist creep at high temperatures are one of the primary candidates for efficient heat exchangers in fusion reactors. Cu–Cr–Nb–Zr (CCNZ) alloys, which were designed to improve the strength and creep life of ITER Cu–Cr–Zr (CCZ) reference alloys, have been found to have comparable electrical conductivity and tensile properties to CCZ alloys. The measured creep rupture times for these improved alloys is about ten times higher than the ITER reference alloys at 90–125 MPa at 500 °C. However, the effects of neutron irradiation on these alloys, and the ensuing material properties, have not been studied; thus, their utility in a fusion reactor environment is not well understood. This study characterizes the room temperature mechanical and electrical properties of a neutron-irradiated CCNZ alloy and compares them to a neutron-irradiated ITER reference heat sink CCZ alloy. Tensile specimens were neutron irradiated in the High Flux Isotope Reactor (HFIR) to 5 dpa between 250 °C and 325 °C. Post-irradiation characterization included electrical resistivity measurements, hardness, and tensile tests. Microstructural evaluation used scanning electron microscopy, energy dispersive x-ray spectroscopy, and atom probe tomography to characterize the irradiation-produced changes in the microstructure and investigate the mechanistic processes leading to post-irradiation properties. Transmutation calculations were validated with composition measurements from atom probe data and used to calculate contributions to the increased electrical resistivity measured after irradiation. Comparisons with CCZ alloys in the same irradiation heat found that the post-irradiated CCNZ and CCZ alloys had comparable electrical resistivity. Although CCNZ alloys suffered more irradiation hardening than CCZ, the overall tensile behavior deviated very little from non-irradiated values in the temperature range studied.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2024.1465080</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2024.1465080</link>
        <title><![CDATA[Insights into the UO2+x/U4O9 phase characterization in oxidized UO2 pellets as a function of hyper-stoichiometry]]></title>
        <pubdate>2024-10-15T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>C. Gaillard</author><author>H. Lotz</author><author>L. Sarrasin</author><author>Y. Pipon</author><author>R. Ducher</author><author>N. Moncoffre</author>
        <description><![CDATA[We present new insights into the study of the UO2+x/U4O9 equilibrium in UO2 as a function of the hyper-stoichiometry (x) by coupling HERFD-XANES at the uranium M4-edge with micro-Raman spectroscopy mapping. XANES allowed the measurement of uranium speciation in the samples, while Raman spectroscopy was used to individually characterize the composition and localization of the different oxide phases. UO2 pellets were oxidized under dry conditions at temperatures above the UO2+x/U4O9 phase transition to reach hyper-stoichiometries in the range of 0.01 ≤ x ≤ 0.1. Combining both techniques, we could determine the proportions of U4O9 and UO2+x. We show that at a low O/U ratio, U4O9 is present as small clusters inside UO2 grains. As the O/U increases, we found evidence of the formation of a network of U4O9 crystallized inside the UO2+x grains. The variation of the UO2+x phase hyper-stoichiometry (x) was evaluated as a function of the sample oxidation.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2024.1446635</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2024.1446635</link>
        <title><![CDATA[Recovery of neutron-irradiated VVER-440 RPV base metal and weld exposed to isothermal annealing at 343°C up to 2,000 h]]></title>
        <pubdate>2024-08-16T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Eberhard Altstadt</author><author>Frank Bergner</author><author>Jann-Erik Brandenburg</author><author>Paul Chekhonin</author><author>Jakub Dykas</author><author>Mario Houska</author><author>Andreas Ulbricht</author>
        <description><![CDATA[Neutron irradiation causes embrittlement of reactor pressure vessel (RPV) steels. Post-irradiation annealing is capable of partly or fully restoring the unembrittled condition. While annealing at high temperatures (e.g., 475°C) was successfully applied to extend the lifetime of operating VVER-440 reactors, the benefit of annealing at lower temperatures (e.g., 343°C–the maximum to which the primary cooling water can be heated) is a matter of debate. In this study, neutron-irradiated VVER-440 RPV base metal and weld were exposed to isothermal annealing at 343°C up to 2,000 h. Given the limited amount of material, the degree of recovery was estimated in terms of Vickers hardness, the ductile-brittle transition temperature derived from small punch tests, and the master curve reference temperature derived from fracture mechanics tests of mini samples. For the base metal, small-angle neutron scattering was applied to underpin the findings at the nm-scale. We have found significant partial recovery in both materials after annealing for 300 h or longer. The variations of the degree of recovery are critically discussed and put into the context of wet annealing.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2024.1379996</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2024.1379996</link>
        <title><![CDATA[Concept validation of separations for thorium-based radionuclide generator systems for medical application]]></title>
        <pubdate>2024-08-08T00:00:00Z</pubdate>
        <category>Brief Research Report</category>
        <author>Bianca Schacherl</author><author>Kiara Maurer</author><author>Martin Schäfer</author><author>Yvonne Remde</author><author>Frank Geyer</author><author>Annika Fried</author><author>Steffen Alexander Happel</author><author>Martina Benešová-Schäfer</author>
        <description><![CDATA[Targeted alpha therapy (TαT) represents an emerging and cutting-edge treatment option for patients dealing with highly challenging metastatic cancer diseases. Critically, the limited supply of alpha-particle-emitting radionuclides, so-called alpha in vivo nanogenerators, hampers wider utilization of TαT in clinical settings. This could effectively be circumvented by alternative production routes, including straightforward purification and reformulation strategies. Radionuclide generators offering great potential in simple and robust elution strategies can be provided that still adhere to high radioisotopic, radionuclidic, and radiochemical purity criteria. This study takes a first step towards novel separation strategies by providing additional sources of alpha in vivo nanogenerators for TαT through experiments with various metal surrogates. With different systems, 232Th/natBa was used as a radionuclide generator analogue to 227Th/223Ra, and 232Th/natBa/natLa was used as a triplet analogue to 229Th/225Ra/225Ac. Three selective resins (UTEVA, TEVA, DGA-N) were evaluated for the 232Th/natBa system. Two perturbations of the best-performing resin were further evaluated using a larger diameter column and 1 week of equilibration. For the 232Th/natBa/natLa separation system, a combined column with two selective resins (TK200, TK101) was employed and evaluated. The results thus obtained pave the way for alternative separation strategies in radioactive proof-of-concept validation in the near future.]]></description>
      </item><item>
        <guid isPermaLink="true">https://www.frontiersin.org/articles/10.3389/fnuen.2024.1411840</guid>
        <link>https://www.frontiersin.org/articles/10.3389/fnuen.2024.1411840</link>
        <title><![CDATA[Leveraging design of experiments to build chemometric models for the quantification of uranium (VI) and HNO3 by Raman spectroscopy]]></title>
        <pubdate>2024-08-08T00:00:00Z</pubdate>
        <category>Original Research</category>
        <author>Luke R. Sadergaski</author><author>Jeffrey D. Einkauf</author><author>Laetitia H. Delmau</author><author>Jonathan D. Burns</author>
        <description><![CDATA[Partial least squares regression (PLSR) and support vector regression (SVR) models were optimized for the quantification of U(VI) (10–320 g L−1) and HNO3 (0.6–6 M) by Raman spectroscopy with optimized calibration sets chosen by optimal design of experiments. The designed approach effectively minimized the number of samples in the calibration set for PLSR and SVR by selecting sample concentrations with a quadratic process model, despite complex confounding and covarying spectral features in the spectra. The top PLS2 model resulted in percent root mean square errors of prediction for U(VI), HNO3, and NO3− of 3.7%, 3.6%, and 2.9%, respectively. PLS1 models performed similarly despite modeling an analyte with a majority linear response (i.e., uranyl symmetric stretch) and another with more covarying vibrational modes (i.e., HNO3). Partial least squares (PLS) model loadings and regression coefficients were evaluated to better understand the relationship between weaker Raman bands and covarying spectral features. Support vector machine models outperformed PLS1 models, resulting in percent root mean square error of prediction values for U(VI) and HNO3 of 1.5% and 3.1%, respectively. The optimal nonlinear SVR model was trained using a similar number of samples (11) compared with the PLSR model, even though PLS is a linear modeling approach. The generic D-optimal design presented in this work provides a robust statistical framework for selecting training set samples in disparate two-factor systems. This approach reinforces Raman spectroscopy for the quantification of species relevant to the nuclear fuel cycle and provides a robust chemometric modeling approach to bolster online monitoring in challenging process environments.]]></description>
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