Article
Materials Science, Multidisciplinary
Yulia Mishchenko, Kyle D. Johnson, Daniel Jadernas, Janne Wallenius, Denise Adorno Lopes
Summary: The oxidation behavior of composite UN-AlN, UN-Cr 2 N/CrN, and UN-AlN-Cr 2 N/CrN pellets was investigated under different conditions, showing improved properties compared to pure UN pellets. The composites exhibited unique oxidation characteristics in air and steam environments.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
Cole Moczygemba, Jonathan George, Eduardo Montoya, Eunja Kim, Geronimo Robles, Elizabeth Sooby
Summary: The research and development of high uranium density fuels is crucial for enhancing reactor technology, safety, performance, and economics. This study explores the performance of U3Si2 fuel alloyed with Zr in high temperature steam oxidation environments, demonstrating delayed onset of oxidation and improved homogeneity.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Elizabeth S. Sooby, Brian A. Brigham, Geronimo Robles, Joshua T. White, Scarlett Widgeon Paisner, Erofili Kardoulaki, Brandon Williams
Summary: This study provides steam oxidation kinetic data for high purity uranium mononitride (UN) at high temperatures, and observes the variation in the oxidation kinetic response of UN to the presence of carbon contaminants. The results show that high purity, high density UN samples have a delayed reaction time with steam and the reaction is suppressed in a reducing steam atmosphere, producing predominantly ammonia and UO2.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Jason Schulthess, David Kamerman, Alexander Winston, Alex Pomo, Tammy Trowbridge, Xiaofei Pu, Nicolas Woolstenhulme, Devin Imholte, Colby Jensen, Daniel Wachs
Summary: New fuel systems are being developed to improve accident tolerance, fuel utilization, and economics of nuclear reactors. Experimental tests under transient irradiation conditions show that accident tolerant materials maintain their geometry but may develop cracks in the cladding due to fuel pellet-induced displacement.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Yulia Mishchenko, Kyle D. Johnson, Janne Wallenius, Denise Adorno Lopes
Summary: In this study, composite UN-AlN, UN-Cr, UN-CrN, and UN-AlN-CrN pellets were fabricated, obtaining an advanced microstructure with different modes of interaction between phases. The results allowed for the assessment of a methodology for fabricating UN composites with controlled microstructure, including the observation of the ternary U2CrN3 phase in pellets containing Cr and CrN dopants.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
J. Buckley, D. T. Goddard, T. J. Abram
Summary: The study focused on the effect of processing parameters on the achieved density of Spark plasma sintering (SPS)-fabricated uranium silicide (U3Si2) fuel forms. Results showed that varying density pellets can be produced by adjusting parameters, with dwell temperature and pressure having the most significant impact on density. Interaction studies revealed that graphite was a preferable die liner over molybdenum for U3Si2, and polyethylene glycol resulted in oxide impurities similar to U3Si2 manufactured via conventional methods.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Nuclear Science & Technology
Natsumi Mitsuboshi, Hiroshi Sagara
Summary: Small modular reactors (SMRs) have attracted significant research interest due to their various applications, and the flexibility and regulatory concerns are essential. The study found that U3Si2 fuel showed better reactivity and lower operating temperature compared to UO2 fuel, but demonstrated complexity during processing.
ANNALS OF NUCLEAR ENERGY
(2021)
Article
Energy & Fuels
A. Abdelghafar Galahom, Ehab M. Aboelyazid, S. A. EL-Fiki, Moustafa Aziz
Summary: Different fuel types were investigated to find an optimum fuel for the VVER-1200 reactor. Uranium carbide and uranium nitride were proposed as alternative fuels to UO2. The neutron properties of these fuel types were investigated using MCNPX 2.7. The results showed that UO2 and (UC)-C-12 achieved criticality till 35 GWD/MTHM, while (UN)-N-14 and (UN)-N-15 achieved criticality at 30 GWD/MTHM and 24 GWD/MTHM, respectively. Safety factors such as MTC, FTC, and beta(eff) were calculated to ensure the operation safety of the proposed fuel. The distribution of thermal neutron flux and normalized power was displayed to determine the optimum fuel compared to UO2.
ENERGY SOURCES PART A-RECOVERY UTILIZATION AND ENVIRONMENTAL EFFECTS
(2023)
Article
Materials Science, Multidisciplinary
D. A. Andersson, C. Matthews, Y. Zhang, B. Beeler
Summary: DFT calculations were used to study the thermodynamic and kinetic properties of point defects in the ,B phase of uranium. The fastest diffusion rate was determined to be through a vacancy mechanism in the z crystallographic direction, with good agreement between predicted uranium self diffusivity and experimental data. The study provides valuable information for fuel performance models regarding swelling and gas evolution.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
M. W. D. Cooper, K. A. Gamble, L. Capolungo, C. Matthews, D. A. Andersson, B. Beeler, C. R. Stanek, K. Metzger
Summary: U3Si2, an advanced fuel candidate with high fissile density and thermal properties, has data gaps in thermophysical and thermomechanical properties. This study used DFT and MD simulations to predict point defect concentrations under irradiation, informing a creep model based on diffusional creep and dislocation creep, which compares well with experimental data and has been implemented in a fuel performance code. Demonstrations show negligible creep in U3Si2 due to its high thermal conductivity at low reactor temperatures.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
Qusai Mistarihi, James Buckley, Joel Turner, Tim Abram
Summary: The fabrication and thermal conductivity of mixed-phase UN-UB2 composite pellets were investigated. The results showed that the fabricated composites had a higher thermal conductivity than UN at low temperatures, but the increase was less significant at higher temperatures and composites with low UB2 content had lower thermal conductivity.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Materials Science, Multidisciplinary
Jianguo Yu, Cole D. Blakely, Jason D. Hales, Hongbin Zhang
Summary: This study compares the coping time of FeCrAl and Zircaloy claddings under Short-Term Station Blackout (ST-SBO) accidents and identifies a new cladding failure mechanism for high temperature and low stress, as well as investigates the origin of low stress at burst.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
J. A. Yingling, K. A. Gamble, Elwyn Roberts, R. Austin Freeman, Travis W. Knight
Summary: U3Si2-SiC concept fuel shows promise as a replacement for UO2-Zr4 fuels during steady-state operation, but the failure of the mSiC layer generally occurs before significant thermal creep in U3Si2 during low power operation.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Chemistry, Analytical
Abhijit Saha, Khushboo Kumari, Sadhan Bijoy Deb, Manoj Kumar Saxena
Summary: Quality assurance of nuclear materials is crucial for the safe operation of nuclear reactors. This study developed a novel three-step matrix separation procedure for quantifying trace impurities in U3Si2-Al using ICP-MS, reducing liquid waste generation significantly. Validation of the analytical methodology showed analyte recoveries >= 95% and RSDs within 8%.
JOURNAL OF ANALYTICAL ATOMIC SPECTROMETRY
(2021)
Article
Materials Science, Multidisciplinary
Douglas E. Burkes, Ian J. Schwerdt, Tanja K. Huber, Harald Breitkreutz, Christian Reiter, Winfried Petry, Jason L. Schulthess, Andrew M. Casella, Amanda J. Casella, Edgar C. Buck, Karl N. Pool, Paul J. MacFarlan, Matthew K. Edwards, Frances N. Smith
Summary: Measurements of physical and thermal properties were conducted on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier clad in Al alloy 6061. The study provides insights into the behavior of U-Mo fuel under high performance research reactor irradiation conditions, showing a decrease in density with increasing fission density and an increase in thermal conductivity with temperature. The findings suggest that fission gas swelling may be a key factor influencing the deviation between model calculations and measurements.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
Kyle D. Johnson, Janne Wallenius, Mikael Jolkkonen, Antoine Claisse
JOURNAL OF NUCLEAR MATERIALS
(2016)
Article
Nuclear Science & Technology
Kyle Johnson, Valter Strom, Janne Wallenius, Denise Adorno Lopes
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
(2017)
Article
Materials Science, Multidisciplinary
Alicia M. Raftery, Joao Gustavo Pereira da Silva, Darrin D. Byler, David A. Andersson, Blas P. Uberuaga, Christopher R. Stanek, Kenneth J. McClellan
JOURNAL OF NUCLEAR MATERIALS
(2017)
Article
Nuclear Science & Technology
Denise Adorno Lopes, Selim Uygur, Kyle Johnson
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
(2017)
Article
Materials Science, Multidisciplinary
Kyle D. Johnson, Denise Adorno Lopes
JOURNAL OF NUCLEAR MATERIALS
(2018)
Article
Materials Science, Multidisciplinary
Christian M. Petrie, Joseph R. Burns, Alicia M. Raftery, Andrew T. Nelson, Kurt A. Terrani
JOURNAL OF NUCLEAR MATERIALS
(2019)
Article
Nuclear Science & Technology
Nathan Capps, Yong Yan, Alicia Raftery, Zachary Burns, Tyler Smith, Kurt Terrani, Ken Yueh, Michelle Bales, Kory Linton
NUCLEAR ENGINEERING AND DESIGN
(2020)
Article
Nuclear Science & Technology
Alicia M. Raftery, Rachel L. Seibert, Daniel R. Brown, Michael P. Trammell, Andrew T. Nelson, Kurt A. Terrani
Summary: Ceramic-metallic nuclear fuels, such as uranium nitride microspheres encased in a molybdenum matrix, were fabricated using advanced manufacturing techniques. Detailed microstructural analysis and thermal conductivity measurements of the sintered fuel pellets were conducted to evaluate the effects of temperature and pressure processing conditions.
NUCLEAR TECHNOLOGY
(2021)
Article
Materials Science, Multidisciplinary
Yulia Mishchenko, Kyle D. Johnson, Daniel Jadernas, Janne Wallenius, Denise Adorno Lopes
Summary: The oxidation behavior of composite UN-AlN, UN-Cr 2 N/CrN, and UN-AlN-Cr 2 N/CrN pellets was investigated under different conditions, showing improved properties compared to pure UN pellets. The composites exhibited unique oxidation characteristics in air and steam environments.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
David J. Sprouster, Erofili Kardoulaki, Randy Weidner, Alicia M. Raftery, Mohamed Elbakhshwan, Reeju Pokharel, Helmut M. Reiche, Darrin D. Byler, Sanjit K. Ghose, Eric Dooryhee, Kenneth J. McClellan, Lynne E. Ecker
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)