Article
Chemistry, Physical
Hanns Gietl, Takaaki Koyanagi, Xunxiang Hu, Makoto Fukuda, Akira Hasegawa, Yutai Katoh
Summary: This study provides experimental evidence of radiation-enhanced recrystallization in tungsten and undoped tungsten-rhenium alloys under fusion-relevant environments. It is found that potassium or lanthanum doping in tungsten alloys improves resistance to radiation-enhanced grain growth. The study also highlights the importance of considering radiation-enhanced recrystallization in the design and application of tungsten plasma-facing components in future nuclear fusion reactors.
JOURNAL OF ALLOYS AND COMPOUNDS
(2022)
Article
Physics, Multidisciplinary
Zhang Guo-Shuai, Yin Chao, Wang Zhao-Fan, Chen Ze, Mao Shi-Feng, Ye Min-You
Summary: Tungsten is a candidate material for divertor targets in future fusion reactors. Neutron irradiation can induce recrystallization of tungsten, leading to potential brittle failure of the divertor. This study introduces an irradiation-enhanced grain boundary migration factor into the recrystallization model and finds that both irradiation and thermal activation diffusion coefficients play important roles in the recrystallization process. The thermal activation effect eventually dominates the recrystallization process as the temperature increases.
ACTA PHYSICA SINICA
(2023)
Article
Chemistry, Physical
Byeong Seo Kong, Ji Ho Shin, Taejeong An, Changheui Jang, Hyoung Chan Kim
Summary: In this study, the tensile deformation behavior of double-forged and recrystallized commercial-grade tungsten was investigated under different strain rates. The results showed that the change in the dominant dynamic recrystallization mechanism has a significant effect on the ductility of the material, and the initial microstructure also plays a crucial role in the strain-rate dependence of mechanical behavior.
Article
Materials Science, Multidisciplinary
D. Terentyev, C. Yin, A. Dubinko, C. C. Chang, J. H. You
Summary: In this study, neutron irradiation and subsequent hardness measurements were performed on tungsten grades at temperatures reaching up to 1200 degrees C. The results showed that even at one third of the melting point, neutron exposure can increase hardness by 40% to 70%, depending on the grade. Transmission electron microscopy analysis revealed that the hardening at 1200 degrees C is mainly attributed to the presence of dislocation loops and a high density of voids.
INTERNATIONAL JOURNAL OF REFRACTORY METALS & HARD MATERIALS
(2021)
Article
Nuclear Science & Technology
J. Riesch, A. Feichtmayer, J. W. Coenen, B. Curzadd, H. Gietl, T. Hoeschen, A. Manhard, T. Schwarz-Selinger, R. Neu
Summary: Advanced materials, such as tungsten fibre-reinforced composites, are used to address issues with baseline materials in fusion reactors. High-energy ions are used to simulate the effects of neutron irradiation on the mechanical properties of these materials. Fine tungsten wire provides a means to study irradiation damage and can be used to predict the properties of bulk composite materials.
NUCLEAR MATERIALS AND ENERGY
(2022)
Article
Materials Science, Multidisciplinary
Masafumi Akiyoshi, Lauren M. Garrison, Josina W. Geringer, Hsin Wang, Akira Hasegawa, Shuhei Nogami, Yutai Katoh
Summary: The Japan-US PHENIX project irradiated tungsten materials in the HFIR reactor using Gd shielding to absorb thermal neutrons. The thermal diffusivity of irradiated specimens was compared at different temperatures and annealing conditions, showing limited contribution of crystalline defects to the degradation of thermal diffusivity. The results indicated that the thermal diffusivity of irradiated specimens approached that of unirradiated specimens at elevated temperatures.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Physics, Multidisciplinary
ChihCheng Chang, Dmitry Terentyev, Aleksandr Zinovev, Wouter Van Renterghem, Chao Yin, Patricia Verleysen, Thomas Pardoen, Monika Vilemova, Jiri Matejicek
Summary: The study evaluated the microstructure and hardness of four tungsten grades before and after neutron irradiation, investigating the influence of microstructure on neutron damage accumulation, characterizing through various experimental techniques.
Article
Materials Science, Multidisciplinary
Dmitry Terentyev, Chih-Cheng Chang, Chao Yin, A. Zinovev, Xin-Fu He
Summary: This contribution presents the results of recent neutron irradiation experiments on pure tungsten in the material test reactor BR2 in Belgium. Various irradiation conditions were applied to assess the effects on mechanical properties, revealing a considerable shift in the ductile to brittle transition temperature in the high-temperature region even at low damage doses.
Article
Physics, Fluids & Plasmas
Dmitry Terentyev, Petra Jenus, Elisa Sal, Aleksandr Zinovev, Chih-Cheng Chang, Carmen Garcia-Rosales, Matej Kocen, Sasa Novak, W. Van Renterghem
Summary: Development of refractory metals for plasma-facing armour material remains a priority in fusion research. Key indicators for material selection include resistance to high temperature recrystallization, material strength for thermal fatigue cracking, and tolerance to neutron irradiation. In this study, the effect of neutron irradiation on mechanical properties and microstructure of several tungsten grades is investigated.
Article
Physics, Multidisciplinary
Wolfgang Pantleon
Summary: Plasma-facing components of future fusion reactors will use tungsten-based materials as armor. Annealed pure tungsten is brittle at room temperature, while plastically deformed tungsten is more ductile at ambient temperatures. Understanding restoration mechanisms and quantifying temperature-dependent recrystallization kinetics are essential for assessing materials performance and informed materials selection.
Article
Materials Science, Multidisciplinary
H. Wang, Z. M. Xie, L. C. Zhang, L. Han, R. Liu, Q. F. Fang, X. P. Wang, C. S. Liu, Xuebang Wu
Summary: The damage behaviors of ITER grade tungsten under repetitive thermal loads have been investigated. The results show that there are negligible effects on the microstructures, surface morphologies and mechanical properties with low absorbed power densities. However, serious damages occur when the heat loads reach a certain level, leading to severe surface roughness and grains coarsening, which significantly degrade the material's mechanical properties.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Physics, Fluids & Plasmas
Chao Yin, Dmitry Terentyev, Andrii Dubinko, Tao Zhang, Marius Wirtz, Steffen Antusch, Roumen H. Petrov, Thomas Pardoen
Summary: Six tungsten grades were irradiated at different temperatures and doses to investigate the irradiation-induced hardening through Vickers hardness tests. Most materials showed an increase in Vickers hardness as the irradiation temperature increased from 600 to 1000 degrees Celsius, except for ZrC-reinforced tungsten. The hardening in TiC-strengthened material progressively increased with irradiation temperature, not solely explained by the accumulation of neutron irradiation defects.
Article
Materials Science, Multidisciplinary
A. Dubinko, D. Terentyev, C. Yin, W. Van Renterghem, B. Rossaert, M. Rieth, E. E. Zhurkin, A. Zinovev, C. C. Chang, S. Van Dyck, G. Bonny
Summary: The study found that the main contribution to hardening at high irradiation temperatures comes from voids, while dislocation loops only provide a significant contribution at 600 degrees Celsius irradiation. Excellent agreement between model predictions and experimentally measured hardness increase was observed for single crystal and ITER specification grade, while hardening induced in cold rolled plate was overestimated by the model.
INTERNATIONAL JOURNAL OF REFRACTORY METALS & HARD MATERIALS
(2021)
Article
Materials Science, Multidisciplinary
Chao Yin, Giovanni Bonny, Dmitry Terentyev
Summary: In this study, the hardness of single crystal tungsten after high temperature neutron irradiation was investigated, revealing anisotropy in hardness and proposing a method to deduce the true hardness and its angle-dependent variation.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Nuclear Science & Technology
Shigeru Saito, Kazuhiro Suzuki, Hiroki Obata, Yong Dai
Summary: This study conducted a post-irradiation examination of pure tungsten (W) and tantalum (Ta) specimens irradiated at a neutron source. Tungsten is a potential candidate as a solid spallation-target material due to its favorable properties, but it also suffers from disadvantages such as poor corrosion resistance and irradiation embrittlement. To improve these properties, cladding technologies using Ta for W alloys have been developed. The experimental results showed the effect of irradiation on the mechanical properties of the materials, with poly-crystal W exhibiting almost no ductility after irradiation and single-crystal W showing some elongation.
NUCLEAR MATERIALS AND ENERGY
(2023)
Article
Materials Science, Multidisciplinary
Jing Wang, Yuji Hatano, Takeshi Toyama, Tomoaki Suzudo, Tatsuya Hinoki, Vladimir Kh. Alimov, Thomas Schwarz-Selinger
Summary: The addition of 0.3 at.% Cr in the tungsten matrix leads to a significant reduction in the retention of hydrogen isotopes, especially at high temperatures. This is attributed to the suppression of vacancy-type defects formation by the addition of chromium.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Nuclear Science & Technology
Yina Du, Tatsuya Hinoki
Summary: SiC fiber reinforced tungsten composites were prepared by hot press process, and the effect of tungsten foil thickness on the properties of the composites was investigated. The results showed that the composites with a thickness of 0.08 mm exhibited better mechanical properties and higher pseudo ductility. In addition, recrystallization of the tungsten foils occurred after sintering, and the presence of tungsten was confirmed by XRD.
NUCLEAR MATERIALS AND ENERGY
(2022)
Article
Materials Science, Multidisciplinary
Tatsuya Hinoki, Fumihisa Kano, Sosuke Kondo, Yoshiyuki Kawaharada, Yumiko Tsuchiya, Moonhee Lee, Hiroyuki Sakai
Summary: This study aims to understand the high-temperature water corrosion and steam oxidation behavior of liquid phase sintering silicon carbide and develop stable liquid phase sintering silicon carbide composites. The results show that the formation of silicate and Yttrium Aluminum Garnet improves the corrosion resistance and thermal shock resistance of the materials. The modified particle-dispersion liquid phase sintering silicon carbide composites are promising materials for light water reactors.
Article
Materials Science, Multidisciplinary
Yina Du, Baopu Wang, Yansong Zhong, Tatsuya Hinoki
Summary: In this work, various ceramic coatings were evaluated for their ability to suppress the reaction between tungsten (W) and silicon carbide (SiC). The multi-dipped Er2O3 coating and the sputtered nitrides showed good performance compared to other coatings. The study provides suggestions for choosing an appropriate interface material between SiC and W.
Article
Materials Science, Multidisciplinary
Minsuk Seo, Ke Wang, John R. Echols, A. Leigh Winfrey
Summary: Tungsten undergoes residual helium changes and microstructure deformation in high heat flux and helium plasma environment. The absence of helium in the resolidified tungsten matrix and the presence of defects, reduced grain size, and microstructure deformation affect hardening.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Shuhei Nogami, Itsuki Ozawa, Daisuke Asami, Naoya Matsuta, Seiji Nakabayashi, Siegfried Baumgaertner, Philipp Lied, Kiyohiro Yabuuchi, Takeshi Miyazawa, Yuta Kikuchi, Marius Wirtz, Michael Rieth, Akira Hasegawa
Summary: The addition of tantalum to tungsten-tantalum alloys improves their mechanical properties, resistance against recrystallization, and resistance to high heat flux exposure. This makes them a promising material for fusion reactor applications.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Ryo Ishibashi, Yasunori Hayashi, Huang Bo, Takao Kondo, Tatsuya Hinoki
Summary: The study demonstrates that using coating technology can effectively reduce hydrothermal corrosion of SiC fuel cladding during normal operation, and after irradiation, the adhesion strength between the coating and the SiC substrate is good without delamination and cracking.
Article
Materials Science, Multidisciplinary
Chinthaka M. Silva, Keith J. Leonard, Lauren M. Garrison, Chris D. Bryan, Kiel S. Holliday
Summary: The study shows that low-temperature neutron irradiation has a significant impact on the tensile behavior of 316L stainless steel. Irradiated specimens, free of defects, exhibit reduced tensile ductility and increased radiation-induced hardening. Both base metal and e-beam welded specimens exhibit similar fracture surfaces indicative of ductile rupture after low-temperature neutron irradiation.
METALLURGICAL AND MATERIALS TRANSACTIONS A-PHYSICAL METALLURGY AND MATERIALS SCIENCE
(2022)
Review
Nuclear Science & Technology
L. M. Garrison, Y. Katoh, T. Hinoki, N. Hashimoto, J. R. Echols, J. W. Geringer, N. C. Reid, J. P. Allain, B. Cheng, D. Dorow-Gerspach, V. Ganesh, H. Gietl, S. A. Humphry-Baker, E. Lang, I. McCue, J. Riesch, L. L. Snead, G. D. W. Smith, J. R. Trelewicz, Y. Yang, S. J. Zinkle
Summary: The plasma-facing components (PFCs) of future fusion reactors require intricate structures and multiple materials. The behavior of internal solid interfaces in PFCs under neutron irradiation is being explored through the FRONTIER U.S.-Japan collaboration. Promising materials in various areas are presented, including copper alloys, tungsten-copper composites, tungsten-steel composites, additively manufactured tungsten, particle-reinforced tungsten, and tungsten and SiC fiber composites.
FUSION SCIENCE AND TECHNOLOGY
(2023)
Article
Materials Science, Ceramics
Kazuya Shimoda, Tatsuya Hinoki
Summary: BN-nanoparticle-containing SiC-matrix-based composites without a fiber/matrix interface were fabricated by SPS. The mechanical properties of the composites were investigated and the composites with a BN nanoparticle content of 50 vol.% showed quasiductile fracture behavior. The composites also exhibited high strength and bending, proportional limit stress, and ultimate tensile strength values under ambient conditions.
INTERNATIONAL JOURNAL OF APPLIED CERAMIC TECHNOLOGY
(2023)
Article
Materials Science, Multidisciplinary
Takeshi Miyazawa, Yuta Kikuchi, Masami Ando, Ju-Hyeon Yu, Kiyohiro Yabuuchi, Takashi Nozawa, Hiroyasu Tanigawa, Shuhei Nogami, Akira Hasegawa
Summary: This study explores the effects of alloying elements (Re and Ta) on the microstructural evolution of recrystallized tungsten (W) under proton and self-ion irradiations. It is found that the addition of Re and Ta suppresses the formation of voids in W. The presence of Re inhibits the mobility of small dislocation loops and SIA clusters, while Ta inhibits the mobility of SIA clusters. In self-ion irradiation, solute Re suppresses the raft formation and void formation. The main reason for the irradiation hardening of W-3%Re is the presence of voids and dislocation loops.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Materials Science, Multidisciplinary
Bo Huang, Meng She, Lin Feng, Yansong Zhong, Kanjiro Kawasaki, Fujio Shinoda, Tatsuya Hinoki
Summary: The effect of Y2O3-Al2O3 sintering additive on the irradiation response of LPS-SiC materials was investigated. CVD-SiC and LPS-SiC specimens were subjected to ion irradiation and compared. The volumetric swelling of LPS-SiC was attributed to the sintering additive YAG.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Materials Science, Multidisciplinary
Jing Wang, Yuji Hatano, Takeshi Toyama, Tatsuya Hinoki, Kiyohiro Yabuuchi, Yi-fan Zhang, Bing Ma, Alexander V. Spitsyn, Nikolay P. Bobyr, Koji Inoue, Yasuyoshi Nagai
Summary: This study systematically investigates the irradiation responses of binary W alloys, focusing on the binding energy of an alloying element with a W self-interstitial atom (W-SIA). Plates of W, W-0.3 at.% Cr, W-5 at.% Re, W-2.5 at.% Mo, and W-5 at.% Ta alloys were irradiated, and the formation of vacancy-type defects, the precipitation of alloying elements, and the changes in hardness were studied. It was found that the addition of Cr and Re effectively suppresses the formation of vacancy-type defects, while Ta and Mo have no significant suppression effect. Irradiation hardening was observed in all materials, but its degree was smaller in the W-5 at.% Re alloy.
MATERIALS & DESIGN
(2023)
Article
Nuclear Science & Technology
Yuya Imagawa, Ryuta Hashidate, Takeshi Miyazawa, Takashi Onizawa, Satoshi Ohtsuka, Yasuhide Yano, Takashi Tanno, Takeji Kaito, Masato Ohnuma, Masatoshi Mitsuhara, Takeshi Toyama
Summary: This study conducted creep tests on 9Cr-ODS steel and found that a single equation can express creep rupture strength from 700°C to 1000°C. The validation of the ring creep test method was also conducted.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
(2023)
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)
