-
FCC-Zr phase transformation-induced defect substructure in Zr2Si precipitates equilibrated in Si-modified Zircaloy-4 J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-18 Muhammad Ali, Fusen Yuan, Fuzhou Han, Wenbin Guo, Jie Ren, Jianan Hu, Qichen Wang, Yingdong Zhang, Geping Li
[Display omitted]
-
Performance and properties evolution of near-term Accident Tolerant Fuel: Cr-doped UO2 J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-17 Adrien J. Terricabras, Sean M. Drewry, Keri Campbell, Elizabeth J. Judge, Darrin D. Byler, Emily S. Teti, Arjen van Veelen, Scarlett Widgeon Paisner, Joshua T. White
Chromium-doped UO fuel has received significant interest due to the ability for chromium to produce pellets with large average grain size (> 30 μm), which has shown to increase fission gas retention during operation. Sintering of chromium-doped UO pellets was pursued with oxygen potential and sintering atmosphere controlled to tailor the final microstructure of the material. Chromium additions in this
-
An improved analysis of small punch deformation for evaluating tensile properties J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-17 Aishwary Vardhan Pandey, V. Karthik, Abdul Rahman Shaik, Ashish Kolhatkar, T.K. Haneef, Divakar R
Small punch (SP) method is a promising technique for evaluating tensile properties using small volumes of test material, yet its applicability is bound by large errors in estimation of tensile yield strength (YS) using the yield force (F) determined by the existing two tangent, offset or bilinear function fitting methods. In this study, a yield force (F) based on the occurrence of yielding across the
-
Effect of grain boundaries and rigid inclusions on plasticity in nickel bicrystals containing helium bubbles and radiation-induced self-interstitial atom clusters J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-16 Tung Yan Liu, Michael J. Demkowicz
We use molecular dynamics to assess the effect of grain boundaries and rigid intergranular inclusions on plastic deformation in nickel (Ni) containing helium (He) bubbles and self-interstitial atom clusters. Our simulations show that plasticity in Ni bicrystals is relatively uniform, with no localized slip bands or nano-twins. We attribute this behavior to grain boundaries, which block dislocations
-
Comparison of PM-HIP to Forged SA508 Pressure Vessel Steel Under High-Dose Neutron Irradiation J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-16 Wen Jiang, Yangyang Zhao, Yu Lu, Yaqiao Wu, David Frazer, Donna P. Guillen, David W. Gandy, Janelle P. Wharry
[Display omitted]
-
Effect of lithium on the atmospheric corrosion characteristics of TZM as the first-wall material in EAST J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-16 Wei Zheng, Rong Yan, Guizhong Zuo, Lei Mu, Niuxian Liu, Xiancai Meng, Rui Ding, Junling Chen
The effect of Lithium (Li) coating on the corrosion behavior of molybdenum-titanium-zirconium (TZM) alloy was investigated in low- and high-humidity air at room temperature (RT). Additionally, the corrosion mechanism was also studied through immersion tests in deionized water and LiOH solution. The test results confirm that the corrosion of the TZM alloy depends on the presence of HO and O, particularly
-
Effect of alternating current on the corrosion behavior of Inconel 693 alloy in borosilicate glass J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-16 Tianyu Liu, Jing Ma, Yingju Li, Ce Zheng, Xiaohui Feng, Yuansheng Yang
The corrosion behavior of Inconel 693 alloy in borosilicate glass at 1150°C with different alternating current (AC) densities was investigated. It was found that the samples with AC applied had greater dimensional loss than the samples without AC applied. The corrosion products were all the double-layer film in two cases, with the outer layer of CrO and the inner layer of AlO. In addition, corrosion
-
Environmentally-Assisted Cracking of Electropolished 316L Stainless Steel in Molten FLiNaK Salt J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-16 Xavier Quintana, Jake Quincey, Samuel A. Briggs
Environmentally-assisted cracking is a well-known and extensively studied phenomenon in light water reactor environments but remains relatively unexplored in proposed molten salt reactor systems where the mechanisms at play are expected to be quite different. Here, slow strain rate testing has been performed on 316L stainless steel tensile specimens during simultaneous exposure to a molten LiF-KF-NaF
-
Towards atomistic modelling of solid Pb-O formation and dissolution in liquid lead coolant: interatomic potential development J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-15 G.V. Khnkoian, V.S. Nikolaev, V.V. Stegailov
Microscopic description of solid Pb-O formation and dissolution in liquid lead coolant is important for the modelling of fast neutron reactors. For this purpose, in this work we develop interatomic potential models for lead melt with dissolved oxygen. Potentials fitting is based on density functional theory (DFT) calculations and quantum molecular dynamics. Two different potential models are trained
-
Characterization of Fluidized Bed Chemical Vapor Deposition ZrC Coatings on PyC/YSZ Kernels Deposited under Differing Conditions J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-13 Peter Doyle, Eddie Lopez-Honorato, Gokul Vasudevamurthy, Jim Miller, Harry Meyer III, Tyler Gerczak
Coated fuel particle architectures with ZrC coatings are candidate fuels for advanced power reactors and space nuclear propulsion (SNP) concepts. Owing to its relevance to SNP, the composition, microstructure, and mechanical properties of eight ZrC coatings prepared by fluidized bed chemical vapor deposition were evaluated. Evaluation by SEM and EBSD showed that all grains were columnar. Across the
-
Prior Implantation of Hydrogen as a Mechanism to Delay Helium Bubbles, Blistering, and Exfoliation in Titanium. J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-13 Svetlana Fink Ilyasafov, Nitzan Maman, Ulrich Kentsch, Victor Y. Zenou, Moshe Vaknin, Yevgeny Rakita, Gabriel Zamir, Itzhak Dahan, Roni Z. Shneck
This study explores the delaying of the formation of helium bubbles and blisters in pure titanium by hydrogen pre-implantation. Titanium, implanted with helium (40 KeV, 5 × 10 ions/cm²), exhibited large bubbles that cause exfoliation after heat treatment, whereas hydrogen pre-implantation inhibited bubble growth at room temperature and reduced the exfoliation after heat treatment.
-
Development of multi-scale computational frameworks to solve fusion materials science challenges J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-11 A. Lasa, S. Blondel, M.A. Cusentino, D. Dasgupta, P. Hatton, J. Marian, D. Perez, W. Setyawan, B.P. Uberuaga, Q. Yu, B.D. Wirth
Over the past two decades, the US-DOE has funded multiple projects that rely on high-performance computing and exascale computing platforms to accelerate scientific discoveries and address grand scientific challenges, such as harnessing fusion energy. In this article, we review in detail one of these efforts aimed at enhancing our capability to model plasma-facing materials subject to plasma and high-energy
-
Size dependence of micro-scale mechanical properties on heavy-ion irradiated tempered-martensitic steel evaluated through nanoindentation and micropillar compression tests J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-11 Diancheng Geng, Hao Yu, Masami Ando, Hiroyasu Tanigawa, Hironori Kurotaki, Takashi Nozawa, Sosuke Kondo, Ryuta Kasada
The size dependence of microscale mechanical properties is crucial for evaluating irradiation hardening using nanoindentation and micropillar compression tests. To predict bulk hardness without the indentation size effect in heavy-ion irradiated materials, this study designed a modified Nix–Gao model incorporating the effect of irradiation hardening. The model was used to determine the bulk hardness
-
Interface-controlled mechanical properties and irradiation hardening in nanostructured Cr/Zr multilayers J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-09 Xiaoxuan Fan, Yaqiang Wang, Kai Wu, Jinyu Zhang, Gang Liu, Jun Sun
Interface as an effective defect-absorbing trap can reduce the accumulation of irradiation damage by promoting the recombination of point defects. As an efficient strategy for designing radiation resistant materials, nanostructured multilayers have attracted extensive interests due to their controllable structure and density of heterogeneous interface. In this work, Cr/Zr nanostructured metallic multilayers
-
Formulation of high-temperature strength equation of 9Cr-ODS tempered martensitic steels using the Larson–Miller parameter and life-fraction rule for rupture life assessment in steady-state, transient, and accident conditions of fast reactor fuel J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-08 Takeshi Miyazawa, Takashi Tanno, Yuya Imagawa, Ryuta Hashidate, Yasuhide Yano, Takeji Kaito, Satoshi Ohtsuka, Masatoshi Mitsuhara, Takeshi Toyama, Masato Ohnuma, Hideharu Nakashima
This paper discusses the applicability of Straalsund et al.’s technique for combining the Larson–Miller parameter (LMP) and life-fraction rule to form a single high-temperature strength equation for 9Cr-oxide-dispersion-strengthened (ODS) tempered martensitic steels (TMS). It uses the extensive dataset on creep rupture, tensile, and temperature-transient-to-burst tests of 9Cr-ODS TMS cladding tubes
-
Preferential intergranular oxidation as a potential degradation mechanism for Alloy X-750 CANDU spacers J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-08 A.E. Yaedu, L. Volpe, J.D. Henderson, S. Ramamurthy, K. Daub, F. Scenini, S.Y. Persaud
Oxidation experiments on Alloy X-750 were performed in reducing CO-CO mixtures to assess the possibility of preferential intergranular oxidation (PIO) in relation to CANDU garter-spring spacers. The testing environments simulate hypothetical off-chemistry scenarios, promoting the oxidation of the alloying elements, but not Ni. Surface and cross-section microscopy corroborated by Auger electron spectroscopy
-
Effect of glass forming additives on low-activity waste feed conversion to glass J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-08 Miroslava Vernerová, Karolína Šůsová, Martina Kohoutková, Jaroslav Kloužek, Petra Cincibusová, Pavel Ferkl, Jose Marcial, Pavel Hrma, Albert A. Kruger, Richard Pokorný
A significant effort was invested in the past to develop and refine mathematical models that relate the composition of nuclear waste glasses with their properties, such as viscosity, electrical conductivity, or chemical durability. However, less attention has been paid to the formulation of the melter feed itself, such as the chemical form and the particle size of the glass forming and modifying additives
-
Thermodynamics and Kinetics of Delayed Hydride Cracking in Zirconium Alloys: a review. J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-07 G.A. McRae, C.E. Coleman
Hydrogen moves in zirconium because of forces associated with gradients in concentration and stress. When solubility limits are reached, stable hydrides form within control volumes that include stabilizing clouds and Cottrell atmospheres of hydrogen in solution, otherwise unstable hydrides form that can continue to grow as observed in delayed hydride cracking and predicted by the Diffusion First Model
-
Coupling effects in borosilicate glass leaching: A study on La/V doping J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-07 Kai Bai, Xiaofen Chen, Jiangjiang Mao, Yuhe Pan, Yuqian Sun, Yuchuan Wang, Haiqiang Zhou, Peng Lv, Tieshan Wang, Haibo Peng
Borosilicate glass could solidify high-level radioactive waste (HLW), and the research on the leaching behavior can optimize the composition, thereby improving its stability and durability. In this work, the leaching behaviors of borosilicate glass in the presence of La/V and their coupling effects were investigated. The results revealed significant coupling effects among La, VO, and CO in an aqueous
-
Insights on the structure and properties of sodium iron phosphate glasses from molecular dynamics simulations J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-07 Jayani Kalahe, T.S. Mahadevan, Xiaonan Lu, John D. Vienna, Brian J. Riley, Jincheng Du
Iron phosphate glasses are promising nuclear waste forms while more detailed understanding of their structures and structure-property relations are still needed to better design waste glass compositions. In this work we report studies of three series of sodium iron phosphate (NFP) glasses: 60PO-(40-)FeO-NaO ( = 0→40), (100–2)PO-FeO-NaO ( = 5→17.5) and one with different iron redox ratio, to understand
-
Dose and compositional dependence of irradiation-induced property change in FeCr J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-07 Kay Song, Dina Sheyfer, Kenichiro Mizohata, Minyi Zhang, Wenjun Liu, Doğa Gürsoy, David Yang, Ivan Tolkachev, Hongbing Yu, David E.J. Armstrong, Felix Hofmann
Ferritic/martensitic steels will be used as structural components in next generation nuclear reactors. Their successful operation relies on an understanding of irradiation-induced defect behaviour in the material. In this study, Fe and FeCr alloys (3–12%Cr) were irradiated with 20 MeV Fe-ions at 313 K to doses ranging between 0.00008 dpa to 6.0 dpa. This dose range covers six orders of magnitude, spanning
-
Effect of heat treatment on the microstructure of medium burn-up U-Mo monolithic fuel foils J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-06 F.G. Di Lemma, T.L. Trowbridge, J.F. Jue, D. Salvato, S. Anderson, C.A. Smith, B.D. Miller, D.D. Keiser, J.J. Giglio, J.I. Cole
Using scanning electron microscopy (SEM), this study evaluates the microstructure evolution of uranium-molybdenum (U-Mo) fuel foils made with and without heat treatment at medium burn-up (of approximately 5 × 10 f/cm). The impact of annealing treatments on critical microstructural properties of the U-Mo fuel foils, including porosity, grain structure, Mo homogeneity, and fuel interaction with the Zr
-
Mechanism of dynamic recrystallization of a FeCrAl alloy during hot compression J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-05 Wei Hou, Xinmin Wang, Peinan Du, Jingyuan Pei, Ruiqian Zhang, Qi Xu, Shaoyu Qiu
FeCrAl alloys have the opportunity to replace current Zr-based cladding materials in light water reactors, and dynamic recrystallization (DRX) plays an important role on microstructural controlling of FeCrAl alloys during hot or warm deformation, such as forging, hot rolling, extrusion. In this research, dynamic recrystallization mechanism of FeCrAl alloy has been systematically investigated by microstructure
-
Thermal creep analysis and correlation development for manufactured HT9 cladding J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-03 Dong-Ha Kim, Cheol Min Lee, Jun-Hwan Kim, Sung-Ho Kim, Sunghwan Yeo, Yong-Kook Lee
HT9 cladding was manufactured through successive cold working and heat treatments, and its creep strains were obtained at various stresses and temperatures in the range of 10–120 MPa and 843–953 K, respectively, for up to 20,000 h. Primary and secondary creep regimes were observed, whereas only a few specimens at extreme conditions experienced a tertiary creep regime and creep rupture. The reference
-
Oxidation Mechanism and Surface Characterization of Pyrolytic Graphite in Simulated Air for Pyrochemical Reprocessing Application J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-02 Rongali Hareesh, E. Vetrivendan, S. Balakrishnan, Ravikumar Sole, S. Ningshen
[Display omitted]
-
The behavior of helium bubble evolution under neutron irradiation in different tungsten surfaces J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-01 Jiarui Jian, Liqun Shi, Bin Zhang
Tungsten, as a kind of material for the plasma-facing wall in fusion reactor and tokamak, is subjected to irradiation by the high-energy neutron and the high fluxes of hydrogen and helium plasmas. Especially, in the presence of helium (He) bubbles, neutron irradiation can significantly exacerbate the deterioration of the material's mechanical properties. However, due to experimental limitations, the
-
Evolution of extended defects in UO2 during high temperature annealing J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-01 Chang-Yu Hung, Joshua Ferrigno, Robert O. Gentile, Marat Khafizov, Lingfeng He
The evolution of thermophysical characteristics of uranium oxide (UO) under irradiation is a critical aspect of nuclear fuel performance of light water reactors. This study examines the effect of high-temperature annealing at 1000 °C - 1600 °C on defect evolution in krypton irradiated UO at room temperature. Transmission electron microscopy analysis reveals that the size of irradiation-induced bubbles
-
Prediction of the energetics of stable self-interstitial atoms at tungsten grain boundaries via machine learning J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-29 Xiaolin Li, Yi Hu, Xiangyan Li, Yange Zhang, Yichun Xu, Xuebang Wu, C.S. Liu
[Display omitted]
-
Grain refinement and associated strengthening in laser additive repaired uranium J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-29 Qingdong Xu, Mingxing Li, Lei Yang, Bin Su, Xue Liu, Dongli Zou, Tao Shi, Xiaobin Yue
Synchronous wire feeding laser additive repairing (LAR) was first developed for uranium to recover the geometry and mechanical properties of localised defects or damaged uranium components. Significant refined grains without evident defects were achieved in the repaired zone of the LARed uranium samples. In the heat affected zone, there is a bimodal distribution of grains which comprises finely recrystallized
-
Extending damage accumulation of commercial reactor irradiated 316 stainless steel with ion irradiation J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-28 Miao Song, Kevin G. Field, Kai Sun, Gary S. Was
Austenitic 316 stainless steel flux thimble tubes (FTTs) removed from a commercial pressurized water reactor (PWR) were re-irradiated using nickel ions to evolve the irradiated microstructure to higher damage levels than those achieved in reactor. The microstructures of the neutron-plus-ion irradiated samples were compared with those of neutron irradiated only samples at the same doses to evaluate
-
Thermal conductivity and deuterium/helium plasma irradiation effect of WTaCrVTi high entropy alloy J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-28 Yongzhi Shi, Zhenyu Jiang, Wenjie Zhang, Tongjun Xia, Xinyu Ren, Meiqi Wang, Lisha Liang, Kaigui Zhu
A refractory WTaCrVTi high-entropy alloy (HEA) film was prepared by magnetron sputtering, and a pure tungsten (W) film was also prepared in the same way used as a comparison sample. The electrical resistivity of the films in the temperature range of 4 K–300 K was measured by the standard four-probe method and the results showed that the resistivity of the HEA film is about 10 times higher than that
-
The microstructure effects on irradiation response of ferritic – martensitic steels J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-27 Weicheng Zhong, Lizhen Tan
Microstructural optimization to achieve greater mechanical strength has been one of the focuses in ferritic–martensitic steels development. However, these optimized microstructures’ effects on the radiation response are not well known. In this work, two ferritic–martensitic steels (9Cr-NbMo and 9Cr-Ta) underwent neutron irradiation in the High Flux Isotope Reactor, and their room-temperature post-irradiation
-
The critical influencing factors responsible for the particle cracking in UMo/Zr dispersion fuel plates during post-irradiation anneal tests J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-27 Feng Yan, Changbing Tang, Xiaobin Jian, Shurong Ding, Yuanming Li, Lin Zhang
The post-irradiation anneal tests for dispersion fuel plates are commonly employed to determine the blistering threshold temperature in order to evaluate their resistance to accidents. In this study, a three-dimensional simulation method for UMo/Zr dispersion fuel plates is developed to predict the normal stresses in the fuel skeleton and some related variables during in-pile irradiation and out-of-pile
-
Raman spectroscopy of zirconium hydride J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-24 Eddie Lopez-Honorato, Liangbo Liang, Yong Yan, Katherine Montoya, Nathan Capps
The characterization of zirconium hydride is important in the nuclear industry because of the hydrogen-induced embrittlement of Zircaloy cladding and its use as a neutron moderator. This paper introduces the use of Raman spectroscopy for the characterization of zirconium hydride. First-principles density functional theory (DFT) calculations were used to predict the Raman spectra of ζ-ZrH, γ-ZrH, δ-ZrH
-
Oxide Layers in Ni-doped FeCrAl Alloy in 320°C Radioactive Hydrogenated Water J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-23 Logan Joyce, Peng Wang, Rajnikant V. Umretiya, Andrew Hoffman, Yi Xie
Iron-chromium-aluminum (FeCrAl) alloys have been the interest of the international nuclear energy industry as an accident tolerant fuel cladding replacement to the Zircaloy cladding alloys due to their enhanced corrosion resistance in both normal operation and accident conditions in light water reactors. It is important to understand the corrosion behaviors of FeCrAl before deploying it to nuclear
-
Study of grain growth kinetics of Zr-2.5%Nb alloy quenched from (α+β) and β-phase region J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-22 Aishwarya Upadhye, Arnomitra Chatterjee, A.K. Bind, Apu Sarkar, R.N. Singh
The Zr-2.5 %Nb is used as the Pressure Tube (PT) material in Pressurized Heavy Water Reactors (PHWRs) in both cold worked and stress relieved as well as heat treated condition. For the development of heat-treated pressure tube, process parameters need to be tuned to achieve optimum microstructure and texture. In microstructural aspect, prior β grain size is a very important factor in influencing the
-
Experimental emulation of 10B(n, α)7Li reaction-induced microstructural evolution of Al-B4C neutron absorber used in the dry storage of spent nuclear fuel J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-20 Woong Ha, Geon Kim, Yunsong Jung, Sangjoon Ahn
-
Manufacturing porous U-10Zr metallic fuels with controllable microstructure by volume control spark plasma sintering J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-20 Dong Zhao, Tiankai Yao, Michael T Benson, Kun Yang, Andre Boussard, Junhua Shen, Fidelma G. Di Lemma, Jie Lian
In this paper, the volume control spark plasma sintering tool has been designed and applied to sinter porous U-10Zr metallic fuels, by which the sintered sample volume can be precisely controlled. Ethanol and NHHCO are used to control the powder compact or as pore formers to control the pore size and pore structure. Without pore formers, the fuel pellet displays an inhomogeneous microstructure consisting
-
Annealing effect on deuterium retention in W-Cr-Y alloy J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-17 Y. Wang, Z. Harutyunyan, Yu. Gasparyan, O.V. Ogorodnikova, D. Sinelnikov, N. Efimov, X. Tan, A. Umerenkova, M. Grishaev
Deuterium (D) trapping in the W-11.4Cr-0.6Y alloy manufactured by field assisted sintering technology (FAST) and pre-annealed at temperatures of 1050, 1250, 1473 K was studied by thermal desorption spectroscopy (TDS). D incorporation in W-11.4Cr-0.6Y alloy was carried out by irradiation with deuterium ions (670 eV/D) to a fluence of 5×10 D/m. The change in Cr concentration in the bulk of the W-Cr-Y
-
Cluster dynamics simulations of tritium and helium diffusion in lithium ceramics J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-16 Ankit Roy, Michel Sassi, Krishna Chaitanya Pitike, Mark S. Lanza, Andrew M. Casella, David J. Senor, Christopher Matthews, David A. Andersson, Ram Devanathan
Tritium (T) and He diffusion in LiAlO and LiAlO phases influences the performance of tritium producing burnable absorber rods (TPBARs) by affecting the gas release, swelling and thermal conductivity of Li-bearing ceramic pellets. Frenkel pair defects and clusters created by irradiation can attract T and He interstitials and form clusters of the type in a Li, Al or O vacancy site (notation denotes x
-
Effect of hydrogen–helium interaction on their segregation and desorption at the W/HfC interfaces by first-principles calculations J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-16 Yuxiang Zhang, Yange Zhang, Xiangyan Li, Yichun Xu, Z.M. Xie, R. Liu, C.S. Liu, Xuebang Wu
Understanding the effect of the interaction between hydrogen (H) and helium (He) in tungsten (W)-based plasma-facing materials is essential for predicting materials performance in the context of fusion environments. In this work, the effect of H-He interaction for their segregation, accumulation and desorption behaviors at typical W/HfC interfaces were systematically investigated by first-principles
-
Fluorination of UO2, La2O3, and Y2O3 using ZrF4 J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-15 Brenton Davis, Jinsuo Zhang
Due to its high affinity towards oxygen, zirconium fluoride can fluorinate oxides found in used nuclear fuel while producing zirconium oxide. This study accomplished this using a lithium fluoride, sodium fluoride, zirconium fluoride (26–37–37 mol%) molten salt known as FLiNaZr with a melting point of 709 K. The high concentration of zirconium fluoride in the FLiNaZr molten salt also prevents the formation
-
Hydrogen Motion in Near Stoichiometric Yttrium Dihydride at Elevated Temperatures J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-15 Ercan Cakmak, M. Nedim Cinbiz, Aditya Sundar, Eugene Mamontov, Jianguo Yu, Xunxiang Hu, Kory D. Linton
The high-temperature motion of hydrogen in near stoichiometric yttrium dihydride (YH, x=1.62 and 1.87 at.%) was investigated using incoherent quasi-elastic neutron scattering and Density Functional Theory (DFT) calculations as a function of hydrogen stoichiometry. Translational motion (diffusivity) of hydrogen in yttrium dihydride was only observed in a temperature range of 1073-1173K under vacuum
-
Preparation of Li4SiO4 pebbles with high strength and inhibited lithium volatilization via lithium alginate J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-15 Runze Geng, Jin Hu, Youwen Zhai, Juemin Yan, Kaijun Wang, Weijun Zhang, Qingming Chen, Xiang Liu, Kaizhao Wang
The tritium breeder's primary research focus is to inhibit lithium's volatilization during sintering and improve the mechanical properties of tritium breeder ceramic pebbles. In this study, ceramic pebbles were prepared by wet method using lithium alginate as a lithium supplement and polyvinyl alcohol. The effects of different contents of lithium alginate (LA) and polyvinyl alcohol (PVA) on ceramic
-
Oxidation behavior of Zr-2.5Nb alloy exposed to steam in the temperature range of 600–1200 °C J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-13 V.S.V. Anantha Krishna, Sai Karthik Nouduru, Kiran K. Mandapaka, Parag M. Ahmedabadi, Geogy J. Abraham, S. Shukla, S. Roychowdhury, V. Kain
Zr-2.5Nb pressure tube material was subjected to isothermal steam exposures in the range 600–1200 °C for different durations to predict its oxidation behavior during hypothetical accident scenarios. The oxidation exponent obtained from power law fitting of the thermogravimetry (TG) curves deviated from parabolic regime. Nevertheless, oxidation kinetics was described by determining the parabolic rate
-
Tungsten fiber-reinforced tungsten composites and their thermal stability J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-13 Daniel Ahlin Heikkinen Wartacz, Johann Riesch, Karen Pantleon, Wolfgang Pantleon
Tungsten will be used as armor material for plasma-facing components in future fusion reactors, but its propensity to embrittlement by microstructural restoration at high temperatures poses a challenge for its use. Tungsten fiber-reinforced tungsten composites (W/W) with drawn tungsten wires embedded in a polycrystalline tungsten matrix remedy the inherent brittleness of tungsten and achieve pseudo-ductile
-
Evaluation of Sb-Nd and Te-Nd phases within the U-Zr fuel matrix and their interactions with HT9 alloy J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-12 Nathan D. Jerred, Rabi Khanal, Michael T. Benson, Samrat Choudhury, Indrajit Charit
Antimony (Sb) and tellurium (Te) were investigated as potential additives for U-10Zr (wt.%) metallic fuel to limit the fuel-cladding chemical interaction (FCCI) with HT-9 alloy. Neodymium (Nd) was utilized to simulate the formation of lanthanide-based solid fission products which are known to play a detrimental role in FCCI. Fuel alloys of U-Zr-Sb-Nd and U-Zr-Te-Nd were evaluated in their annealed
-
Application of the immobilized low-activity waste glass corrosion model to the static dissolution of 24 statistically-designed alkali-borosilicate waste glasses J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-12 Sebastien N. Kerisit, James J. Neeway, Charmayne E. Lonergan, Benjamin Parruzot, Jarrod V. Crum, Richard C. Daniel, Gary L. Smith, R. Matthew Asmussen
-
Diffusion controlled hydrolysis in geopolymers under gamma irradiation J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-12 Vincent Cantarel, Frederic Chupin, Maryline Ortega-Charlot, Isao Yamagishi, Fumiyoshi Ueno
When nuclear waste is immobilized in cement or geopolymer, gases may be generated by corrosion and radiolysis. This production must be accurately predicted, and waste loading and countermeasures selected accordingly to avoid overpressure and limit the risk of explosion in the case of dihydrogen (H). We measured and simulated H generation and release from water-saturated geopolymer confined in a glass
-
-
Off-center positioning of helium in a vacancy in metals J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-09 J.L. Cao, W.Q. Xie, J.B. Lin, X.F. He, V. Wang, Shigenobu Ogata, W.T. Geng
It is a conventional wisdom that the closed-shell structure of a He atom prompts it to occupy the center of a vacancy in metals where the charge density is the lowest. Although it was found accidentally by density functional theory (DFT) calculations that the He sits off-center in a vacancy in vanadium (V), the underlying physics has remained elusive. We have performed systematic DFT study of the positioning
-
Investigation of oxidation behaviors of bi-layer CrAl-Cr coated Zircaloy-4 in steam at 1300 oC J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-08 Chaowen Zhu, Yu Sun, Lin Qi, Muhong Li, Huahai Shen, Chen Chen, Song Zeng, Yan Meng, Xiaosong Zhou, Xiaochun Han
In the high-temperature steam environment, CrO exhibits instability due to redox interaction with Zr. To address this issue, a series of bi-layer CrAl-Cr coatings were designed and compared with the pure Cr coating in steam at 1300 °C. External AlO formed when the Al content exceeded 32 at.%. Conversely, internal AlO formed when the Al content was below 18 at.%, which not only preserved the integrity
-
Initial oxidation behavior of α-U and γ-U surfaces J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-07 Houao Liu, Fuzhu Liu, Hongxiang Zong, Xiangdong Ding, Jun Sun
[Display omitted]
-
Phase field modeling of irradiation-induced shrinkage fracture in TRISO fuel particle J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-07 Jiatong Tan, Yingwei Wu, Qun Li, Yanan He, Chunyu Yin
With the features in retaining fission products and maintaining structural stability, TRISO fuel particle has been widely implemented and studied in the development of advanced nuclear fuel element. During service, however, the key silicon carbide (SiC) layer of TRISO may fail due to irradiation-induced shrinkage fracture of inner pyrolytic carbon (IPyC) layer. Hence, it has great significance to analyze
-
Spectroscopic and theoretical analyses of the reaction of SrO in molten chloride and fluoride salts J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-07 Dokyu Kang, Choah Kwon, Wonseok Yang, Seokjoo Yoon, Yunu Lee, James T.M. Amphlett, Sang-Eun Bae, Sangtae Kim, Sungyeol Choi
In this paper, the reaction of SrO in molten chloride and fluoride systems is investigated using Raman spectroscopy, combined with density functional theory calculations. The formation of SrOCl was observed in various salt compositions, including LiCl, LiCl-KCl, and LiCl-LiF, at temperatures ranging from 298 K to 923 K. Thermodynamic calculations estimated that SrOCl is more stable than SrO and SrCl
-
Identification of irradiation-induced phases in thermomechanically strengthened P92 steel after Fe ion irradiation at 700℃ J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-07 Jiazheng Wang, Yinzhong Shen
In order to study the changes of precipitate phases in TMT-strengthened high-chromium ferritic/martensitic steels under high-temperature irradiation conditions, TMT-strengthened P92 steel was irradiated with 3.5 MeV Fe ions at 700°C up to 0.74 dpa, and precipitate phases in the steel before and after irradiation were observed and analyzed using transmission electron microscopy and energy dispersive
-
Physical insight into interactions of interstitial loops and dislocation lines in austenitic high entropy alloys: atomic-scale modelling J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-06 Ayobami Daramola, Anna Fraczkiewicz, Giovanni Bonny, Gilles Adjanor, Ghiath Monnet, Christophe Domain
Nuclear reactor materials undergo significant changes in their microstructure and mechanical properties due to radiation exposure. Understanding the role of radiation-induced defects in these materials is a complex challenge, especially for new candidate material. This study utilizes molecular dynamics (MD) simulations with a novel interatomic potential to investigate the interactions between mobile
-
Evaluation of primary radiation damage cross sections with uncertainties for charged particles J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-06 Shuyi Chen, Shengli Chen
Radiation damage is an important challenge for materials employed in strong irradiation environments, such as nuclear reactors and accelerators. The radiation damage is often quantified by the primary radiation damage with the unit of displacement per atom (dpa). To more accurately estimate the displacement damage, the athermal recombination-corrected (arc)-dpa model has been proposed by adding an
-
Investigations on 1200 °C steam oxidation behavior of Cr coatings with distinct crystallographic orientation on Zircaloy-4 alloys J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-05 Wenzhe Wang, Guojun Zhang, Caixia Wang, Tao Wang, Yagang Zhang, Tong Xin
Cr-coatings with distinct crystallographic orientation were deposited on Zircaloy-4 alloy by high-power pulsed magnetron sputtering ion plating technology. The crystallographic orientation transformed from Cr (110) to Cr (200) because of high probability of ion collision with the pressure increase. Cr coatings with (200) crystallographic orientation shows superior 1200 °C steam anti-oxidation behavior
-
Influence of transmutation-induced Re/Os content on defect evolution in neutron-irradiated W J. Nucl. Mater. (IF 3.1) Pub Date : 2024-02-05 M. Klimenkov, U. Jäntsch, M. Rieth, H.C. Schneider, D. Terentyev, W. Van Renterghem