Article
Materials Science, Multidisciplinary
T. Jailin, N. Tardif, J. Desquines, P. Chaudet, M. Coret, M-C Baietto, V Georgenthum
Summary: This study aims to investigate the creep behavior of Zircaloy-4 fuel claddings under simulated reactivity initiated accident (RIA) thermo-mechanical conditions. A FEMU-based identification method is proposed to determine the creep law of the cladding, which is weakly coupled to the phase transformation of the material at high temperatures. The identified Norton creep law accurately reproduces the experimental results with a mean error of about 10% for the radial displacement rates in the first 10 seconds.
JOURNAL OF NUCLEAR MATERIALS
(2022)
Article
Materials Science, Multidisciplinary
Lijun Chen, Baifeng Luan, Hongyan Yang, Ruiqian Zhang, Chao Sun, Xiaoling Yang, Shuyu Ma, Peng Wan, Hongling Zhou
Summary: Short-term annealing tests were conducted on Cr-coated Zircaloy-4 cladding by arc ion plating in an Ar environment at various temperatures. The microstructure and adhesion property of the coating were studied using SEM-EDS, EBSD, and STEM-EDS. The results showed that the diffusion of elements increased with higher annealing temperatures, resulting in the formation of an intermetallic layer at the interface. The adhesion properties of the coating remained good below 1300 degrees C.
MATERIALS CHARACTERIZATION
(2023)
Article
Materials Science, Multidisciplinary
S. Guilbert-Banti, A. Viretto, J. Desquines, C. Duriez
Summary: The study showed that pre-oxidation has a protective effect against high temperature oxidation, with pre-oxidized samples exhibiting lower weight gains after steam oxidation compared to bare cladding. Laboratory-grown unirradiated pre-oxidation layers have a stronger protective influence than a corrosion layer formed under irradiation, likely due to differences in oxide microstructure and intergranular porosity.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Chemistry, Physical
Robert Balerio, Hyosim Kim, Andres Morell-Pacheco, Laura Hawkins, Ching-Heng Shiau, Lin Shao
Summary: Plasma nitridation was used to modify the surface of Zircaloy-4, forming a thin layer of ZrN phase with a temperature-dependent hardness. The hardness of the polycrystalline ZrN phase increases with nitridation temperature, reaching a maximum at 700 degrees C.
Article
Nuclear Science & Technology
Ligang Song, Bo Huang, Jianghua Li, Xianfeng Ma, Min Liu, Jishen Jiang, Yanying Hu
Summary: In heavy ion irradiation studies, CrN coating exhibited the best irradiation resistance due to its compact crystal structure and higher density of grain boundary, which contributed to reduce dislocations and voids. Compared to TiAlCrN coating, Cr coating showed better resistance in terms of irradiation dislocation loops.
ANNALS OF NUCLEAR ENERGY
(2021)
Article
Materials Science, Multidisciplinary
M. C. Collins, G. J. Francolini, P. Obreja, N. Scuro, R. Varga, B. Breeden, D. Rosas, B. W. N. Fitzpatrick, E. Geiger, G. Harvel, M. H. A. Piro
Summary: This study investigates the high temperature oxidation of Zircaloy-4 in light water steam and heavy water steam. The reaction kinetics are compared in both environments, and the results show that the oxidation behavior is similar, confirming the previous assumption.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Chemistry, Physical
Wenzhe Wang, Guojun Zhang, Caixia Wang, Tao Wang, Yagang Zhang, Tong Xin
Summary: By changing the bias voltage using high power impulse magnetron sputtering, Cr coatings with different columnar structures were constructed to optimize the oxidation behavior of chromium-coated Zircaloy-4 alloys under 1200 degrees C high-temperature steam for accident tolerant fuel (ATF) claddings. The microstructure, mechanical properties and high-temperature steam oxidation properties of the Cr coatings were studied systematically. The results showed that increasing the bias voltage changed the microstructure of the Cr coatings from a porous columnar structure to a compact columnar structure. The Cr coatings with optimal compact columnar structure exhibited improved hardness and elevated oxidation resistance compared to the coatings with porous columnar structure.
JOURNAL OF ALLOYS AND COMPOUNDS
(2023)
Article
Materials Science, Multidisciplinary
Yoshiyuki Nemoto, Yasuhiro Ishijima, Keietsu Kondo, Yuki Fujimura, Yoshiyuki Kaji
Summary: Previous studies have shown that the oxidation rate of zirconium-based alloy fuel cladding is higher in air-steam mixtures than in dry air under certain conditions. Understanding the mechanism of oxidation in an air-steam environment is crucial for severe accidents in nuclear power plants. Tests were conducted on Zircaloy-4 specimens at 800 degrees C in a mix of air and steam, and the oxidation kinetics, oxide layer details, and hydrogen pick-up were studied. The differences in oxidation mechanism between dry air and air-steam mixtures are discussed based on the experimental results.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Nuclear Science & Technology
Ayumi Itoh, Ryotaro Hagiwara, Shintaro Yasui, Yoshinao Kobayashi, Kan Sakamoto, Masato Mizokami, Mutsumi Hirai, Kenichi Ito
Summary: The interaction between Zircaloy and stainless steel during a severe accident in a light water reactor, leading to eutectic liquefaction, was experimentally investigated at 1573 K. The reaction layer consisted of five phases: α-(Fe,Cr,Ni) phase, metastable (Fe,Cr,Ni)23Zr6 phase, Laves Zr(Fe,Cr,Ni)2 phase, α-Zr(O) phase, and liquid phase. The thickness of the reaction layer on the Fe-rich side followed the time parabolic rate law, while the ones involving liquid phase formation followed a saturation-type convection-controlled function. A formula combining diffusion and convection process was introduced to estimate the reacted volume, showing good agreement with the experimental results. It was realized that a model providing the mass transfer coefficient for the convection-controlled process would be required for improvement of the core degradation model.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY
(2023)
Article
Physics, Multidisciplinary
M. Izerrouken, R. Hazem, U. Yahsi, S. E. Naceri, C. Tav, S. Kuzeci, A. Sari, F. Haid, A. Ishaq, O. Menchi, M. Ghamnia
Summary: The study revealed that the amount of hydride precipitates in Zircaloy-4 decreases after irradiation, while the material exhibits swelling above a certain dose. The hardening of Zircaloy is mainly attributed to dislocation formation, and the transition point from low-dose to high-dose regime is observed around 0.07 dpa.
Article
Materials Science, Multidisciplinary
Jean Desquines, Christian Duriez, Severine Guilbert, Tatiana Taurines
Summary: Experimental studies and modeling have shown that pre-oxidation of nuclear fuel cladding has an impact on post-LOCA quench resistance, with the pre-existing oxide layer exhibiting complex effects that can be modeled using a simple approach. A new correlation was established to evaluate the oxygen enriched alpha-phase thickness, and the growth of the zirconia layer was successfully analyzed and modeled. These models and correlations were implemented in a failure prediction model and compared to experimental results.
JOURNAL OF NUCLEAR MATERIALS
(2021)
Article
Engineering, Mechanical
Bharat Kumar, Deepak Kumar, V. Chaudhry
Summary: Core components of nuclear reactor, made of zirconium alloys, are susceptible to low amplitude vibration leading to material wear at contact interface. Reciprocating wear tests were conducted between zircaloy-4 (Zr-4) under dry and water-submerged conditions, along with wear tests between Zr-4 and stainless steel to confirm material transfer due to adhesion. Analysis using SEM, FESEM, EDS, Raman spectroscopy, and XPS revealed that coefficient of friction (COF) increases with fretting duration but decreases with frequency and displacement amplitude. Greater COF was observed under dry condition compared to submerged condition, with wear mechanism transitioning from adhesion to abrasion as displacement amplitude increased.
TRIBOLOGY INTERNATIONAL
(2023)
Article
Materials Science, Multidisciplinary
Soyoung Kang, Pei-Hsun Huang, Victor Petrov, Annalisa Manera, Taehwan Ahn, Bruce Kammenzind, Arthur T. Motta
Summary: During operation in a nuclear reactor, waterside corrosion of Zr-based nuclear fuel cladding can cause hydrogen ingress. The hydrogen migrates and precipitates as zirconium hydrides if the hydrogen content exceeds the terminal solid solubility. Temperature gradients in the cladding result in preferential hydride precipitation at colder spots. The heat of transport (Q*) of hydrogen in solid solution can be determined by annealing experiments under fixed temperature gradients.
JOURNAL OF NUCLEAR MATERIALS
(2023)
Article
Materials Science, Multidisciplinary
Yang Liu, Said El Chamaa, Mark R. Wenman, Catrin M. Davies, Fionn P. E. Dunne
Summary: By combining experimental and modeling approaches, the study investigated hydride formation in Zircaloy-4 under cyclic thermomechanical loading. It was found that strain rate sensitivity and hydrogen concentration play important roles in controlling the distribution of hydrides.
Article
Chemistry, Analytical
Elizabeth J. Kautz, Ewa C. E. Ronnebro, Arun Devaraj, David J. Senor, Sivanandan S. Harilal
Summary: This study characterizes the spectral features of femtosecond laser-induced plasmas from Zircaloy-4 targets with varying H-1 and H-2 concentrations in a helium gas environment, revealing different ambient pressure dependencies for H-1(alpha) and Zr I emission features, as well as spatial and temporal separation in the laser-induced plasmas. Furthermore, the measured H-2(alpha) emission intensities via femtosecond LIBS for different H-2 concentrations in Zircaloy-4 samples showed a linear trend when plotted against known H-2 concentrations.
JOURNAL OF ANALYTICAL ATOMIC SPECTROMETRY
(2021)
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)