Article
Materials Science, Multidisciplinary
Afiqa Mohamad, Bowen Gong, Tiankai Yao, Adrian R. Wagner, Michael T. Benson, Jie Lian
Summary: This work reports an innovative approach to develop advanced U3Si2 fuels with state-of-the-art materials properties by synergizing multiple effects of microstructure control, oxide protection, and phase transformation-induced mechanically toughening. The incorporation of 3Y-TZP additives enhances fracture toughness and oxidation resistance of U3Si2 fuel, representing a major step towards realizing the potential of high density U3Si2 as the leading fuel concept for nuclear energy systems.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Chemistry, Physical
Afiqa Mohamad, Tiankai Yao, Bowen Gong, Jason Harp, Adrian R. Wagner, Andrew T. Nelson, Jie Lian
Summary: Al-doped U3Si2 composite fuels with controlled microstructure displayed greatly improved oxidation resistance and thermal-mechanical properties. Minimal addition of Al effectively increased the onset oxidation temperature and thermal annealing further enhanced it. Additionally, the Al-doped composite fuels exhibited higher hardness and fracture toughness compared to undoped U3Si2.
JOURNAL OF ALLOYS AND COMPOUNDS
(2021)
Article
Materials Science, Ceramics
Bowen Gong, Erofili Kardoulaki, Kun Yang, Andre Broussard, Dong Zhao, Kathryn Metzger, Joshua T. White, Michael R. Sivack, Kenneth J. Mcclellan, Edward J. Lahoda, Jie Lian
Summary: The synthesis of UN-U3Si2 composites using spark plasma sintering has effectively reduced micro-crack formation, leading to high-density fuel pellets with uniform distribution of phases. These composites exhibit enhanced strength, fracture toughness, and excellent thermal conductivity, showing potential for improving fuel properties compared to monolithic UN.
CERAMICS INTERNATIONAL
(2022)
Article
Materials Science, Multidisciplinary
D. Frazer, B. Maiorov, U. Carvajal-Nunez, J. Evans, E. Kardoulaki, J. Dunwoody, T. A. Saleh, J. T. White
Summary: Understanding the mechanical interaction between fuel and cladding in a nuclear reactor is crucial for preventing cladding failures and the release of radioactive material into the coolant. By measuring the mechanical properties of various materials, better models can be developed to improve understanding of pellet-clad mechanical interactions.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
William A. Hanson, Fabiola Cappia, Joshua T. White, Kenneth J. McClellan, Jason M. Harp
Summary: This study presents post-irradiation examination data on UN-U3Si5 and U3Si5 fuels with Kanthal AF (R) cladding at low burnup levels. The results show good irradiation performance for both the silicide and nitride-silicide composite pellets. Limited axial cracking was observed only in UN-U3Si5 pellets, and microcracking was isolated to the U3Si5 phase, indicating it was not caused by irradiation. Fission gas release was minimal, and no fission gas bubbles were observed in the optical metallography, suggesting acceptable swelling and fission gas behavior.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Energy & Fuels
Hao Chen, Yi Zhou, Quan Li, Zhenhai Liu, Feipeng Qi, Chao Ma, Bo Zhao, Yongzhong Huang
Summary: A multi-objective optimization framework based on genetic algorithm and Gaussian process was established for the optimal design of U3Si2-FeCrAl fuel element. The framework enables achieving the best performance of U3Si2-FeCrAl fuel element under different design parameters.
INTERNATIONAL JOURNAL OF ENERGY RESEARCH
(2022)
Article
Materials Science, Multidisciplinary
Andrea Fazi, Mohammad Sattari, Michal Strach, Torben Boll, Krystyna Stiller, Hans-Olof Andr, Denise Adorno Lopes, Mattias Thuvander
Summary: The accident tolerant fuel (ATF) concept aims to improve safety measures and fuel cladding performance in light water reactors. This study tested two versions of a nitride coating on zirconium claddings under boiling water reactor (BWR) operating conditions and found that both coatings demonstrated improved oxidation protection.
JOURNAL OF NUCLEAR MATERIALS
(2023)
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
Materials Science, Ceramics
Sang-Chae Jeon, Dong-Joo Kim, Dong Seok Kim, Jae-Hwan Yang, Kyoung-Seok Moon
Summary: This study investigates the diffusion kinetics of cesium in ceramic microcell UO2 fuels and suggests a new evaluation method for UO2 fuels with trapping agents based on kinetic data. The addition of aluminum was found to lower the degree of cesium diffusivity, possibly due to the formation of a solid Cs-Al-Si-O compound.
JOURNAL OF THE EUROPEAN CERAMIC SOCIETY
(2021)
Article
Nuclear Science & Technology
F. N. Kryukov, A. Belyaeva, M. Skupov, L. M. Zabudko, Yu S. Mochalov
Summary: The investigations on helium-bonded fuel pins with mixed uranium-plutonium nitride and lead sub-layers have been completed after irradiation, revealing the patterns of fuel swelling, cladding corrosion, and mechanical properties change. The results showed differences in fuel swelling rates between helium and lead-bonded pins, as well as the influence of impurities on cladding carburization and oxidation. Additionally, the mechanical properties of the cladding materials showed significant strength and ductility along the fuel pins, indicating the possibility of prolonging irradiation to higher fuel burn-up.
NUCLEAR ENGINEERING AND DESIGN
(2021)
Article
Materials Science, Multidisciplinary
Keyou S. Mao, Caleb P. Massey, Yukinori Yamamoto, King A. Unocic, Maxim N. Gussev, Dalong Zhang, Samuel A. Briggs, Omer Karakoc, Andrew T. Nelson, Kevin G. Field, Philip D. Edmondson
Summary: Post-neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys, revealing a compositional dependency of the Cr and Al content on the ratio of sessile and glissile dislocation loops. The number density of alpha' precipitates is inversely correlated to the starting Cr concentration. Higher irradiation doses result in higher densities of dislocation loops and alpha' precipitates. Compared with first-generation FeCrAl alloys, these advanced alloys exhibit lower Cr concentration in alpha' precipitation, leading to improved radiation resistance.
Article
Physics, Applied
Bowen Gong, Kun Yang, Dong Zhao, Andrew T. Nelson, Jie Lian
Summary: This study systematically investigates the oxidation behavior of microcrystalline and nanocrystalline U3Si2 and elucidates the oxidation mechanisms and kinetic properties. The results provide a quantitative method for evaluating the oxidation resistance of nuclear fuels.
JOURNAL OF APPLIED PHYSICS
(2022)
Article
Materials Science, Multidisciplinary
Yuan Yuan, Yanqing Qin, Kai Xu, Yiming Zhang, Qing Huang, Keke Chang, Shiyu Du
Summary: The solubility of metal M in U3Si2 is closely related to the differences in radius and electronegativity between Si and M, as well as the number of intermediate phases in the binary U-M and Si-M systems. Aluminum has the largest solubility in U3Si2 among the elements studied.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Nuclear Science & Technology
G. Beausoleil, L. Capriotti, B. Curnutt, R. Fielding, S. Hayes, D. Wachs
Summary: The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) completed its first irradiation cycle at Idaho National Laboratory. The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The results of nondestructive post irradiation examination revealed the integrity of the experiments and provided insights into the behavior of the fuel rods. Further destructive examination is needed to fully understand the effects of accelerated irradiation on U-Zr metallic fuel behavior.
NUCLEAR ENGINEERING AND TECHNOLOGY
(2022)
Article
Materials Science, Multidisciplinary
Liang Yin, T. B. Jurewicz, M. Larsen, M. Drobnjak, C. C. Graff, D. R. Lutz, R. B. Rebak
Summary: Iron-Chromium-Aluminum (FeCrAl) alloys are candidate materials for light water reactor accident tolerant fuel (ATF) rod cladding due to their unparalleled resistance to attack by superheated steam. Tests show that these alloys exhibit excellent corrosion resistance in 300 degrees C water, mainly due to the development of a protective oxide layer rich in Cr and Al on the surface.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Chemistry, Physical
Afiqa Mohamad, Tiankai Yao, Bowen Gong, Jason Harp, Adrian R. Wagner, Andrew T. Nelson, Jie Lian
Summary: Al-doped U3Si2 composite fuels with controlled microstructure displayed greatly improved oxidation resistance and thermal-mechanical properties. Minimal addition of Al effectively increased the onset oxidation temperature and thermal annealing further enhanced it. Additionally, the Al-doped composite fuels exhibited higher hardness and fracture toughness compared to undoped U3Si2.
JOURNAL OF ALLOYS AND COMPOUNDS
(2021)
Article
Materials Science, Multidisciplinary
Jason M. Harp, Robert N. Morris, Christian M. Petrie, Joseph R. Burns, Kurt A. Terrani
Summary: An overview of postirradiation examination results for uranium nitride kernels and coated particles irradiated in the High Flux Isotope Reactor shows low irradiation temperatures and relatively small burnup, with no significant irradiation induced changes observed.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
Michael T. Benson, Jason M. Harp, Yi Xie, Tiankai Yao, Kevin R. Tolman, Karen E. Wright, James A. King, Ayman I. Hawari, Qingsheng Cai
Summary: Palladium is being studied as a fuel additive to potentially immobilize lanthanide fission products. The study shows that palladium can react with lanthanide fission products to prevent or decrease fuel-cladding chemical interaction (FCCI). Results indicate that lanthanide compounds in irradiated metallic fuel can be reliably simulated in out-of-pile experiments.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
Fidelma G. Di Lemma, Tammy M. Trowbridge, Luca Capriotti, Jason M. Harp, Michael T. Benson, Robert D. Mariani
Summary: This study investigates the microstructural and elemental characteristics of irradiated metallic fuels with palladium as an additive, which can prevent lanthanide migration and attack on the cladding, but may also form compounds with zirconium leading to unexpected fuel-performance issues. Further research is needed to fully understand the efficacy of using palladium as an additive in metallic fuels.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
D. Frazer, F. Cappia, J. M. Harp, P. G. Medvedev, K. J. McClellan, S. L. Voit, J. Giglio, D. Jadernas, P. Hosemann
Summary: The Advanced Fuels Campaign conducted irradiation tests on minor actinide-bearing mixed oxide fuel to investigate the transmutation of long-lived transuranic actinide isotopes in spent nuclear fuel. The tests showed changes in the microstructure of the fuel and enrichment of Cr in the fuel clad chemical interaction layer.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Bowen Gong, Dong Zhao, Andre Broussard, Jason Harp, Andrew T. Nelson, Jie Lian
Summary: This paper presents a systematic study of the high-temperature creep behavior of dense U3Si2 pellets using spark plasma sintering (SPS) under vacuum conditions. The results indicate a grain boundary sliding creep mechanism and are in agreement with literature. Finite element modeling shows excellent agreement with experimental data, confirming the validity of the experiments. The study demonstrates the potential of SPS for high-temperature mechanical testing of nuclear fuels under vacuum conditions.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Yi Xie, Sven C. Vogel, Michael T. Benson, Jason M. Harp
Summary: The phase evolution of as-cast U-35 wt% Zr and U-50 wt% Zr alloys during thermal cycling was investigated using in-situ neutron diffraction. The results show that different alloys exhibit different phase structures during the thermal cycling process.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Amani Cheniour, Giovanni Pastore, Jason M. Harp, Christian M. Petrie, Nathan A. Capps
Summary: There is a recent push to accelerate nuclear fuel qualification by combining advanced modeling and simulation with accelerated separate effects irradiation testing. An assessment and sensitivity study was conducted on calculated fission gas release from UO2 MiniFuel disks. Qualitatively, the existing fission gas behavior model in BISON reproduces the effects of temperature and burnup as expected, but shows a less satisfactory predictive capability when compared quantitatively with FGR data from irradiated UO2 disks under thermal annealing representative of LOCA conditions.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Yachun Wang, David M. Frazer, Fabiola Cappia, Fei Teng, Daniel J. Murray, Tiankai Yao, Colin D. Judge, Jason M. Harp, Luca Capriotti
Summary: This study investigated the microscale mechanical properties of fuel cladding chemical interaction (FCCI) through advanced electron microscopy characterizations and small-scale mechanical testing techniques. The results showed significant hardening and embrittlement in the FCCI region. Meanwhile, irradiation-induced microstructural and microchemical evolution led to mechanical softening in the unreacted matrix. This study is of importance for understanding the mechanical properties of cladding under high-temperature irradiation.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Yachun Wang, Brandon D. Miller, Jason M. Harp, Daniele Salvato, Luca Capriotti, Tiankai Yao
Summary: Fuel cladding chemical interaction (FCCI) is an important phenomenon to establish design basis for U-10Zr metallic fuel performance. This study investigated the FCCI region in HT9 cladded U-10Zr fuel irradiated to 5.7% FIMA burnup using transmission electron microscopy. Four distinct layers were identified in the FCCI region. The characterization results enhance our understanding of FCCI phenomenon and facilitate the development of microstructure-informed FCCI modeling for metallic fuel.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
William A. Hanson, Fabiola Cappia, Joshua T. White, Kenneth J. McClellan, Jason M. Harp
Summary: This study presents post-irradiation examination data on UN-U3Si5 and U3Si5 fuels with Kanthal AF (R) cladding at low burnup levels. The results show good irradiation performance for both the silicide and nitride-silicide composite pellets. Limited axial cracking was observed only in UN-U3Si5 pellets, and microcracking was isolated to the U3Si5 phase, indicating it was not caused by irradiation. Fission gas release was minimal, and no fission gas bubbles were observed in the optical metallography, suggesting acceptable swelling and fission gas behavior.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Materials Science, Multidisciplinary
Nathan Capps, Larry Aagesen, David Andersson, Oliver Baldwin, W. Cade Brinkley, Michael W. D. Cooper, Jason Harp, Stephen Novascone, Pierre-Clement A. Simon, Christopher Matthews, Brian D. Wirth
Summary: In response to the nuclear industry's desire for extended burnup, the US NRC released a RIL to analyze FFRD in light-water reactors. The unclear element is the significance of tFGR and its effects on fuel performance during loss-of-coolant accidents. Recent research shows that FGR increases with burnup, occurring at lower temperatures due to microcracking and embrittlement. This manuscript summarizes tFGR data, discusses dependencies, and proposes a roadmap for developing a mechanistic tFGR model applicable to various fuel conditions.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Materials Science, Multidisciplinary
Peter Doyle, Juri Stuckert, Mirco Grosse, Martin Steinbrueck, Andrew T. Nelson, Jason Harp, Kurt Terrani
Summary: The QUENCH-19 experiment was a full-bundle test of accident-tolerant fuel cladding, specifically the B136Y3 alloy. Compared to the ZIRLO cladding in QUENCH-15, the B136Y3 cladding in QUENCH-19 released significantly less H2 and reached a lower maximum temperature. No breakaway oxidation was observed in QUENCH-19. The experiment suggests that Fe-CrAl cladding can chemically survive loss-of-coolant accidents followed by rapid ECCS quench if the correct geometry and core design are present.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Materials Science, Multidisciplinary
Shipeng Shu, Yinbin Miao, Bei Ye, Kun Mo, Laura Jamison, Sumit Bhattacharya, Aaron Oaks, Abdellatif M. Yacout, Jason Harp, L. Amulya Nimmagadda, Sanjiv Sinha
Summary: This study investigates the thermal conductivity of U3Si2 that has been amorphized by ion irradiation. The results show that the thermal conductivity of amorphous U3Si2 is significantly lower than that of crystalline U3Si2, which is consistent with previous research findings.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Materials Science, Multidisciplinary
Keyou S. Mao, Tyler J. Gerczak, Jason M. Harp, Casey S. McKinney, Timothy G. Lach, Omer Karakoc, Andrew T. Nelson, Kurt A. Terrani, Chad M. Parish, Philip D. Edmondson
Summary: Characterizing oxide nuclear fuels is challenging due to complex fission products and extreme operating conditions. In this study, a machine learning-enhanced approach is used to accelerate the characterization of spent nuclear fuels and improve the identification of nanophase fission products and bubbles. The application of this approach to irradiated light-water reactor fuels reveals relationships between fission product precipitates and gases, as well as insights into fission versus decay pathways. An algorithm is provided to quantify the chemical segregation of fission products and enable processing of large amounts of microscopy data.
COMMUNICATIONS MATERIALS
(2022)
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)
