Article
Materials Science, Multidisciplinary
Bei Ye, Aaron Oaks, Shenyang Hu, Benjamin Beeler, Jeff Rest, Zhi-Gang Mei, Abdellatif Yacout
Summary: A separate computational branch has been developed in the DART code to simulate the swelling behavior of U-10Mo monolithic fuel in high-power research and test reactors. The model incorporates various physical, thermal, and mechanical processes occurring in the fuel and uses lower length-scale computational methods to calculate materials properties. Calibrated with measured data, the fission gas behavior model is used to simulate and predict the fuel's swelling behavior under different operating conditions.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Nuclear Science & Technology
D. Salvato, C. A. Smith, B. Ye, Z-G Mei, A. M. Yacout, J. Van Eyken, B. D. Miller, D. D. Keiser, I. Y. Glagolenko, J. J. Giglio, A. B. Robinson, A. Leenaers, J. Wight, J. L. Henley
Summary: The effect of SEM imaging parameters on porosity analysis of U-Mo fuel was investigated, and it was found that SEM voltage has the greatest impact on porosity analysis. The increase in voltage from 5 kV to 30 kV resulted in a change in probing depth and improved the estimation of open porosity. This study has the potential to improve the consistency and uniformity of porosity analysis in different research laboratories.
NUCLEAR MATERIALS AND ENERGY
(2023)
Article
Materials Science, Multidisciplinary
Xiaobin Jian, Feng Yan, Xiangzhe Kong, Yong Li, Shurong Ding
Summary: In this study, a mechanistic model of fission gas swelling for U-10Mo fuels is developed and validated by comparing with experimental data. The results show that the model has satisfactory prediction ability, especially for higher burnup cases, and can describe the dominant dependences of external pressure and temperature. Parametric studies reveal the influences of temperature, external pressure, and grain-boundary diffusion enhancement factor on fission gas behaviors.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Rafael Caprani, Philippe Martin, Damien Prieur, Julien Martinez, Myrtille O. J. Y. Hunault, Florent Lebreton, Marie-Margaux Desagulier, Camille Aloin, Loic Picard, Meghan Alibert, Guillaume Gabriel, Patrice Signoret, Nicolas Clavier
Summary: A fabrication route for U1-yPuyO2 & PLUSMN; x SIMfuel (SIMMOx) has been developed to study the speciation of soluble FP (Ce, La, Nd, Sr, Y, Zr) inside spent MOx fuel and their interaction with the U1-yPuyO2 & PLUSMN; x matrix. The results show that the material is compliant with the homogeneity requirement of irradiated MOx and all FP successfully entered the (U,Pu)O2 solid solution.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Nuclear Science & Technology
A. I. Popescu, D. H. Barber, T. Daniels, H. Albasha, L. Sun, R. Murphy-Snow
Summary: SOURCE is a computer code that simulates the release of fission products from CANDU nuclear fuel. It considers the partitioning of fission products in different locations within the fuel and the physical and chemical state of the fuel and fission products. The code has been validated against experiments and used for sensitivity studies to assess its prediction capabilities.
NUCLEAR ENGINEERING AND DESIGN
(2023)
Article
Physics, Applied
J. Lee, Y. C. Chiu, M. A. Johar, C. Bayram
Summary: Cubic GaN epitaxy on large-area U-grooved silicon (100) dies is achieved by metalorganic chemical vapor deposition, exhibiting excellent structural and optical properties. Further improvement in internal quantum efficiency of cubic GaN can be achieved through selective etching.
APPLIED PHYSICS LETTERS
(2022)
Article
Materials Science, Multidisciplinary
Filippo Verdolin, Stephen Novascone, Davide Pizzocri, Giovanni Pastore, Tommaso Barani, Lelio Luzzi
Summary: This study presents a new model extension for fast reactor fuel, with the introduction of a lower bound to the number density of grain-face bubbles to improve predictive capabilities and validated against experimental data. The comparisons of BISON results with the experimental data show satisfactory results and an improvement compared to the standard version of the code.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
C. M. Barr, E. Lang, K. Burns, P. Price, B. D. Miller, D. D. Keiser, A. Aitkaliyeva, K. Hattar
Summary: Nanoscale microstructural characterization of a U-10Mo/Zr barrier layer monolithic fuel plate was conducted using advanced transmission electron microscopy techniques to evaluate the microstructural changes after high burn-up. The study investigated the evolution of gas bubble superlattice, grain restructuring, and Zr inter-action layer through detailed electron microscopy characterization. The results revealed that the ultra-fine grains in irradiated U-10Mo fuel undergo restructuring and are separated by high angle grain boundaries at a burn-up of 4.42 x 10 21 fissions/cm3. Advanced chemical analysis and multi-variable statistical analysis also showed spatial clustering of solid fission product precipitates. Additionally, a newly observed porous nanocrystalline Zr region in the barrier layer was characterized. This work provides important insights into the grain subdivision and restructuring process in neutron-irradiated U-10Mo fuel using advanced microscopy techniques.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Nuclear Science & Technology
Lokesh Verma, Francois Kremer, Karine Chevalier-Jabet
Summary: The release of radioactive fission products from defected fuels can increase coolant activity and lead to primary circuit contamination and radiation exposure. Artificial Neural Networks have been used to predict FP activities and characterize defected fuel rods. A physical model is presented, which considers diffusion release in the pellet region, diffusion and first-order kinetic release in the gap region, and first-order kinetic release in the coolant region. The model is validated and preliminary predictions from the ANN show 99.5 +/- 0.5% accuracy. Sensitivity analysis identifies influential parameters such as linear power, pellet centerline temperature, cladding external surface temperature, pellet radius, gap width, and time of defect onset.
PROGRESS IN NUCLEAR ENERGY
(2023)
Article
Nuclear Science & Technology
Yanbo Jiang, Yong Xin, Wenbo Liu, Zhipeng Sun, Ping Chen, Dan Sun, Mingyang Zhou, Xiao Liu, Di Yun
Summary: A phase-field model was developed to investigate the influence of recrystallization on bubble evolution during irradiation. The simulation results showed that recrystallization accelerated the formation and growth of bubbles.
NUCLEAR ENGINEERING AND TECHNOLOGY
(2022)
Article
Materials Science, Multidisciplinary
Lu Cai, Fei Xu, Fidelma Giulia Di Lemma, Jeffrey J. Giglio, Michael T. Benson, Daniel J. Murray, Cynthia A. Adkins, Joshua J. Kane, Min Xian, Luca Capriotti, Tiankai Yao
Summary: The U-10Zr based metallic nuclear fuel is a leading candidate for next-generation sodium-cooled fast reactors in the United States, but lack of mechanistic understanding of fuel performance poses challenges for commercial qualification. This paper proposes an image data-driven machine learning approach, coupled with domain knowledge, to provide unprecedented quantified insights into the morphology, size, density, and the connectivity of fission gas bubbles.
MATERIALS CHARACTERIZATION
(2022)
Article
Materials Science, Multidisciplinary
A. Germain, J. Sercombe, C. Riglet-Martial, C. Introini, L. Noirot, Y. Pontillon, Ph. Maugis, C. Gueneau
Summary: This paper presents simulations of the VERDON 1 and RT6 tests with high burnup UO2 fuel and a coupling between irradiated fuel thermochemistry and a fission gas release model. The simulations show good agreement with the release kinetics of various fission products and the released fractions of low-/non-volatile fission products and uranium, plutonium. However, the observed differences in fuel-clad melting temperatures are not reproduced.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Nuclear Science & Technology
P. R. Hania, D. A. Boomstra, O. Benes, P. Soucek, A. J. de Koning, I. Bobeldijk, S. de Groot, R. J. M. Konings, E. Capelli, M. Naji, C. Sciolla, P. J. Baas, V. Bhimanadam, N. B. Siccama, G. I. A. Lippens
Summary: The experiment involved irradiating four fluoride fuel salt samples in graphite crucibles in the HFR Petten for 508 Full Power Days under the name SALIENT-01. The goals of the experiment were to study fission product retention in the salt, size distributions of noble metal particles, and interactions between fuel salt and nuclear graphite. Discussions were held on the experiment design, irradiation history, and plans for post-irradiation examinations, along with addressing limitations and follow-up actions for future improvements.
NUCLEAR ENGINEERING AND DESIGN
(2021)
Article
Nuclear Science & Technology
A. Germain, J. Sercombe, C. Riglet-Martial, C. Introini, L. Noirot, Y. Pontillon, Ph. Maugis
Summary: This paper presents simulations of four tests performed on medium to high burnup fuel during the VERCORS and VERDON experimental programs. The impact of the burnup and fission product radial profiles on both the thermochemical equilibria within the pellet and the FP release kinetics is discussed. The conclusion is that the FP release from the fuel pellets is not significantly increased by the consideration of the burnup and FP radial profiles.
NUCLEAR ENGINEERING AND DESIGN
(2022)
Article
Materials Science, Multidisciplinary
Tiankai Yao, Xiang Liu, Yachun Wang, Fei Teng, Daniel J. Murray, Mitchell Meyer, Michael T. Benson, Luca Capriotti
Summary: This study revisited the performance of U-20Pu-10Zr metallic fuel irradiated in the Experimental Breeder Reactor-II using advanced electron microscopies and revealed critical phenomena such as Zr redistribution. These findings provide important guidance for fuel design and improving fuel performance.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Chemistry, Multidisciplinary
Kaitlin E. Johnson, Sukriti Gakhar, Subhash H. Risbud, Marjorie L. Longo
Article
Materials Science, Multidisciplinary
D. A. Lopes, V Kocevski, T. L. Wilson, E. E. Moore, T. M. Besmann
JOURNAL OF NUCLEAR MATERIALS
(2018)
Article
Materials Science, Ceramics
Tashiema L. Wilson, Emily E. Moore, Denise Adorno Lopes, Vancho Kocevski, Elizabeth Sooby Wood, Joshua T. White, Andrew T. Nelson, Jacob W. McMurray, Simon C. Middleburg, Peng Xu, Theodore M. Besmann
ADVANCES IN APPLIED CERAMICS
(2018)
Article
Materials Science, Multidisciplinary
D. A. Lopes, T. L. Wilson, V. Kocevski, E. E. Moore, T. M. Besmann, E. Sooby Wood, J. T. White, A. T. Nelson, S. C. Middleburgh, A. Claisse
JOURNAL OF NUCLEAR MATERIALS
(2019)
Article
Materials Science, Multidisciplinary
Cheng-Kai Chung, Xiaofeng Guo, Gaoxue Wang, Tashiema L. Wilson, Joshua T. White, Andrew T. Nelson, Anna Shelyug, Hakim Boukhalfa, Ping Yang, Enrique R. Batista, Artaches A. Migdisov, Robert C. Roback, Alexandra Navrotsky, Hongwu Xu
JOURNAL OF NUCLEAR MATERIALS
(2019)
Article
Materials Science, Multidisciplinary
A. P. Shivprasad, V Kocevski, T. L. Ulrich, J. R. Wermer, D. A. Andersson, J. T. White
Summary: In this study, the thermodynamics of hydrogen absorption reaction of U3Si2 were experimentally determined and correlated with crystallographic evolution using X-ray diffraction. The study found that the hydride phase of U3Si2 exhibited a maximum stoichiometry between U3Si2H1.8 and U3Si2H2, with a two-phase region and critical temperature. Enthalpy and entropy of the hydrogen absorption reaction increased with hydrogen content but were lower than values for other reactions. Density functional theory modeling results were consistent with experimental findings.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Najeb M. Abdul-Jabbar, Tashiema L. Ulrich, Joshua T. White
Summary: The use of uranium carbides as nuclear thermal propulsion fuels shows promise for deep-space exploration. Transition metals can be incorporated into uranium carbides to increase their melting point, addressing operational challenges. Thermodynamic models and experimental data are being used to refine phase relationships and compositions in the C-Ti-U system.
Article
Materials Science, Multidisciplinary
Jason L. Baker, Josh T. White, Aiping Chen, Tasheima Ulrich, Robert R. Roback, Hongwu Xu
Summary: This study investigated the low-temperature heat capacity of U3Si5 intermetallic compound, observing an enhancement in heat capacity at low temperatures attributed to potential spin-fluctuations. Several thermodynamic parameters were determined based on the experimental data.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
Johnathon C. Ard, Jacob A. Yingling, Kaitlin E. Johnson, Juliano Schorne-Pinto, Mina Aziziha, Clara M. Dixon, Matthew S. Christian, Jacob W. McMurray, Theodore M. Besmann
Summary: The Molten Salt Thermal Properties Database (MSTDB -TC) is a valuable tool for thermodynamic modeling in molten salt reactors. It contains a wide range of models and compounds, with ongoing expansion to include more relevant systems. The database has been used to compute phase equilibria and support corrosion modeling, demonstrating its practical applications.
JOURNAL OF NUCLEAR MATERIALS
(2022)
News Item
Materials Science, Multidisciplinary
Kiyo T. Fujimoto, Tashiema L. Ulrich
Article
Materials Science, Multidisciplinary
Jason L. Baker, Gaoxue Wang, Tashiema Ulrich, Josh T. White, Enrique R. Batista, Ping Yang, Robert C. Roback, Changyong Park, Hongwu Xu
JOURNAL OF NUCLEAR MATERIALS
(2020)
Article
Materials Science, Multidisciplinary
Tashiema L. Ulrich, Sven C. Vogel, Joshua T. White, David A. Andersson, Elizabeth Sooby Wood, Theodore M. Besmann
Proceedings Paper
Materials Science, Ceramics
Kaitlin E. Johnson, Marjorie L. Longo, Subhash H. Risbud
PROCEEDINGS OF THE 42ND INTERNATIONAL CONFERENCE ON ADVANCED CERAMICS AND COMPOSITES: CERAMIC ENGINEERING AND SCIENCE PROCEEDINGS, VOL 39, ISSUE 3
(2019)
Article
Materials Science, Multidisciplinary
Liuming Wei, Jingwen Li, Yonggang Li, Qirong Zheng, Fan Cheng, Chuanguo Zhang, Jingyu Li, Gaofeng Zhao, Zhi Zeng
Summary: This study investigates the influence of He-V complexes on H behaviors on different W surfaces using DFT calculations. The results show that H dissolution is most difficult but H trapping is easiest on the W (110) surface, while the opposite is true on the W (111) surface. Moreover, the presence of He-V complexes increases the difficulty of H diffusion from bulk to surface and desorption.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Yan Meng, Song Zeng, Chen Chen, Chaowen Zhu, Huahai Shen, Xiaosong Zhou, Xiaochun Han
Summary: The characteristics of magnetron sputtered Cr coatings vary with different temperature, bias voltage, and pressure. Coatings with random orientation, good crystallinity, and small grain size exhibit favorable oxidation behavior, while coatings with strong (200) texture, poor crystallinity, and large grains have many intrinsic defects that are detrimental to the protection property of the Cr coatings.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Xinyuan Xu, Zefeng Yu, Wei-Ying Chen, Aiping Chen, Arthur Motta, Xing Wang
Summary: This study presents an automated approach for characterizing grain morphology in TEM images recorded during ion irradiation. By combining a machine learning model and a computer vision algorithm, comparable results to human analysis can be achieved with significantly reduced analysis time. Researchers can train their own models following the procedures described in this study to automate grain morphology analysis of their own TEM images.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Shihao Wu, Dong Wang, Yapei Zhang, Koji Okamoto, Marco Pellegrini, Wenxi Tian, Suizheng Qiu, G. H. Su
Summary: The oxidation and degradation mechanisms of Cr coating on Zr alloy cladding under high temperature steam atmosphere are summarized, and a mathematical analysis model is established to predict the changes in coating thickness. The model is applied in the analysis of structure evolution under different conditions.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
V. Diaz-Mena, J. de Prado, M. Roldan, I. Izaguirre, M. Sanchez, M. Rieth, A. Urena
Summary: The brazeability of a cupronickel alloy was evaluated as a filler alloy for high-temperature joining of tungsten to steel. The study investigated the brazing conditions and the impact of the selected filler on the joint quality using numerical software. The results showed different metallurgical interactions and diffusion phenomena between the filler alloy and the base materials at different temperatures. The study emphasized the importance of selecting a suitable filler to mitigate residual stresses in the joints.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Pengcheng Zhu, Yajie Zhao, Yan-Ru Lin, Jean Henry, Steven J. Zinkle
Summary: This study investigates the effect of heavy-ion irradiation on radiation hardening in high-purity binary alloy Fe18Cr. Nanoindentation testing and high-quality TEM imaging were conducted to extract hardness and microstructure information. The strength factor was accurately calculated based on the detailed TEM characterization of irradiated microstructures, and a refined hardening superposition method was applied to quantify the mechanical properties of ion-irradiated materials.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Bin Wu, Haixia Ning, Hanzhen Zhu, Jianjun Chen, Kang Wang, Daiyu Zhang, Fu Wang, Qilong Liao
Summary: This study discusses the effects of ZrO2 and B2O3 on the phase composition and properties of SAP-based glass-ceramics. The results show that ZrO2 addition improves the formation of NZP phase while restricting the crystallization of AlPO4 phases. The correct ratios of ZrO2 and B2O3 allow only the formation of NZP phase within the SAP glass.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Hwasung Yeom, Greg Johnson, Benjamin Maier, Tyler Dabney, Kumar Sridharan
Summary: Cr-Nb bilayer coatings were developed using cold spray deposition to improve the limiting operational temperature of Cr-coated Zr-alloy system. The coatings exhibited outstanding oxidation resistance at high temperatures and formed continuous intermetallic compound layers at the interfaces.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Padhraic L. Mulligan, Andrew T. Nelson, Chad M. Parish, Patrick A. Champlin, Xiang Chen, Daniel Morrall, Jason M. Harp
Summary: Environmental barrier coatings are being developed to reduce oxidation and embrittlement in Zr-based materials. Chromium nitride is a candidate for this application, but understanding its impact on irradiation-induced creep and microstructure is critical.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Dexuan Yan, Xinlei Cao, Ke Shen
Summary: This study investigated the purification mechanism of polycrystalline graphite by comparing IG-11 graphite with IG-110 nuclear grade graphite. The analysis revealed that metallic impurities in IG-11 were primarily segregated within graphite porosities, while IG-110 demonstrated a significant reduction in impurities. This research contributes to the development of innovative graphite purification techniques for greater purity and stronger oxidation resistance.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Wei Xu, Wei Peng, Lei Shi, Qi Sun
Summary: This paper investigates the oxidation and shape evolution of matrix graphite in high temperature gas-cooled reactors during air-ingress accidents. A reaction kinetics model is established and computational fluid dynamics with a dynamic mesh method is used to simulate the oxidation process. The results show that the geometric shape of graphite changes significantly with increasing flow rate, and the graphite pebbles tend to form a structure with a narrow front and wide tail.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Allison Harward, Casey Elliott, Michael Shaltry, Krista Carlson, Tae-Sic Yoo, Guy Fredrickson, Michael Patterson, Michael F. Simpson
Summary: This paper investigates the hygroscopic properties of eutectic LiCl-KCl absorbed into zeolite-4A. The study finds that water absorption and corrosion worsen with increasing salt loading. It also suggests that the salt can be stored in a non-inert atmosphere for a certain period of time.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Takaaki Koyanagi, Xunxiang Hu, Christian M. Petrie, Gyanender Singh, Caen Ang, Christian P. Deck, Weon-Ju Kim, Daejong Kim, James Braun, Yutai Katoh
Summary: This study provides critical experimental data on the effects of irradiation on the hermeticity of SiC composite cladding, finding that irradiation can cause a decrease in hermeticity and cracking, and coating the outer surface can mitigate the cracking issue.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
S. Krat, A. Prishvitsyn, Yu. Gasparyan
Summary: This study proposes a probabilistic and diffusion-based model to describe the co-deposition of multiple hydrogen isotopes with slowly grown metal layers. The model calculates the relative concentrations of different hydrogen isotopes in the co-deposited metal layers. It is found that if hydrogen isotopes have different detrapping energies, only the isotope with the highest detrapping energy shows a monotonic decrease in concentration with deposition temperature. Furthermore, the study evaluates the uncertainty of tritium concentration in the co-deposited layer based on the uncertainty in detrapping energy of tritium and deuterium, predicting a >10% tritium concentration uncertainty for a 0.01 eV difference.
JOURNAL OF NUCLEAR MATERIALS
(2024)
Article
Materials Science, Multidisciplinary
Tijo Vazhappilly, Arup Kumar Pathak
Summary: This study investigates the effect of Ce atom substitution in UO2 on its thermophysical properties using density functional theory. The results show that the Ce substitution levels and the oxidation state of Ce/U atoms strongly influence the band structure and specific heat capacity of the UO2 lattice. These findings provide important insights into the fuel properties of UO2 under reactor conditions.
JOURNAL OF NUCLEAR MATERIALS
(2024)