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Phase field-volumetric lattice Boltzmann model of ion uptake in porous nuclear waste form materials under continuous flow J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-16 Zirui Mao, Xiaoyu Zhang, Yulan Li, Vanessa Proust, Alban Gossard, Thomas David, Robert Montgomery, Agnes Grandjean, Huidan Yu, Hans-Conrad zur Loye, Shenyang Hu
The flow field within the mesopores of sorbent particles plays a crucial role in radionuclide diffusion and ion uptake kinetics, thus, impacting the overall performance of porous nuclear waste form materials. To fundamentally understand the influence of microstructures and material properties on the radionuclide absorption and retention processes requires a coupled multi-physics model that considers
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Temperature-time dependence and mechanisms of redox reaction in Cr-coated Zr alloy cladding during steam oxidation at 900–1250 ℃ J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-16 Jianxi Deng, JiaDong Zuo, Donghui Geng, Qiaoyan Sun, Zhongxiao Song, Jun Sun
The structural integrity of the Cr coating can be significantly affected with increasing temperature during steam oxidation. Herein, the elemental diffusion, layered structure, and kinetics evolution of Cr coating in steam environment of 900–1250 ℃ for 0.5 to 3 h were studied. The oxidation behavior of Cr-coated Zr claddings associated with redox reaction between CrO and Zr were investigated in detail
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Evolution of Zr(Fe,Cr)2 second phase particles in Zircaloy-2 under heavy ion irradiation J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-16 Kieran Lynch, Ömer Koç, Graeme Greaves, Alexander Carruthers, Mia Maric, Michael Preuss, Aidan Cole-Baker, Philipp Frankel, Joseph Robson
The Zr(Fe,Cr) second phase particles (SPPs) found in Zircaloy-2 and -4 are known to amorphize and dissolve under irradiation. In the present work, their evolution has been studied in situ in a transmission electron microscope (TEM) under 600 keV Ar irradiation at 320°C, taking samples to two different doses of 13 and 24 dpa. Using scanning transmission electron microscopy coupled with energy-dispersive
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Thermal creep in a pre-hydrided Zr1 % Nb nuclear fuel cladding tube J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-15 Vaclav Sklenicka, Kveta Kucharova, Petr Kral, Jiri Dvorak, Marie Kvapilova, Vera Vrtilkova, Jakub Krejci
The effect of hydrogen on the thermal creep behaviour of Zr1 %Nb alloy fuel cladding tubes for use in light water-cooled nuclear reactors was investigated. The tubes were hydrogen charged using an alternative method with a hydrogen content in the range 371–656 wppm H. Comparative constant-stress creep tests were carried out at 350 °C, under stress ranging from 150 to 225 MPa, on both non-hydrided and
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First evidence of metallisation circle in a Chromium doped UO2 pellet submitted to a thermal gradient J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-11 A. Devillaire, L. Le Berre, L. Desgranges, F. Martin, P. Charmasson, C. Colin, N. Tarisien, X. Iltis, K. Hanifi, A. Canizares
In order to better understand the formation of metallic fission product precipitates at the centre of power ramped UO fuel, we performed a separate effects experiment. An annular pellet made of chromia doped UO with an initially uniform distribution of chromium oxide precipitates, was subjected to a temperature gradient. The chromia precipitates became metallic in the hottest part of the pellet but
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Mandrel diameter effect on ring-pull testing of nuclear fuel cladding J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-10 Peter M. Beck, Mathew L. Hayne, Cheng Liu, James Valdez, Thomas Nizolek, Samuel A. Briggs, Stuart A. Maloy, Tarik A. Saleh, Benjamin P. Eftink
Inconsistencies in ring-pull testing methods for thin-walled tubes make it difficult to compare a material's mechanical properties in the hoop direction presented in past publications. The effect of test setup, specifically mandrel diameter, was investigated for nuclear fuel cladding tubes of the iron-chromium-aluminium (FeCrAl) alloy, C26M (Fe-12Cr-6Al-2Mo). Mandrel diameter for ring-pull testing
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Metallothermic reduction of Cerium Chloride in molten Salt using Li, Na, and Ca Metal J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-10 Mario Gonzalez, Sierra Freitas, Chao Zhang, Michael F. Simpson
Options for reducing purified metal chlorides to metallic form were evaluated in support of the development of a metal purification process based on the conversion of impure metal via hydriding followed by chlorination. Metallothermic reduction of CeCl was tested using Li, Ca, and Na as reductant metals with and without a molten salt solvent in MgO and YO-coated MgO crucibles. The effects of added
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Surface blistering and D desorption in high-energy-rate forging W, W-Y2O3, W-TiC exposed to deuterium plasma J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-10 X.F. Xie, Z.M. Xie, H. Wang, S.S. Wu, J.R. Luo, X.X. Zhang, P. Wang, R. Liu, Q.F. Fang, C.S. Liu, X.B. Wu
Pure tungsten (W) and two representative W alloys of W-1 wt.%YO and W-0.5 wt.%TiC were subjected to deuterium (D) plasma exposure at a fluence up to 1 × 10 D/m at 400 K to evaluate their surface blistering and D desorption behaviors. Thermal desorption spectroscopy (TDS) results revealed that high-temperature desorption peaks (> 650 K) dominated in HERF-W, whereas the two W alloys featured higher intensities
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Study on eutectic-oxidation coupling reaction of Cr-Zr system in high temperature steam environment J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-09 Dong Wang, Ruhao Zhong, Xiaocheng Wu, Yapei Zhang, Xiurui Li, Jian Yu, Yicong Lan, G.H. Su, Suizheng Qiu, Wenxi Tian, Yehong Liao, Zhenxun Peng, Chao Guo, Zhongxiao Song, Jun Sun
The isothermal steam oxidation behavior of Cr-coated Zr-1Nb alloy at 1350 ℃ and 1400 ℃ was investigated. The temperatures are above the Cr-Zr eutectic temperature. The reliability of the experimental setup and methodology was confirmed by conducting isothermal steam oxidation experiments on bare Zr-1Nb specimens at 1100∼1400 ℃, and the corresponding oxidation kinetics correlation was obtained. After
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Chloride-induced stress corrosion cracking in Austenitic steels for SNF storage canisters - Recent understanding and advances in mitigation and repair J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-09 Haozheng J. Qu, Jonathan Tatman, Janelle P. Wharry
Chloride-induced stress corrosion cracking (CISCC) is a critical threat to stainless steel (SS) spent nuclear fuel (SNF) and radioactive materials storage canisters currently in service worldwide. In the past decade, extensive research has been conducted to understand fundamental mechanisms of CISCC, as well as to develop effective repair and mitigation methods to prevent catastrophic failure of the
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Improving hydrogen isotope permeation resistance of (TiVAlCrZr)O multi-component metal oxide coating by O+ Ion implantation J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-09 Fen Zhong, Guangxu Cai, Enkai Guo, Yifu He, Bowen Fu, Wentao Ge, Mengqing Hong, Changzhong Jiang, Feng Ren
Oxygen vacancy can greatly affect the performance of metal oxide to hydrogen isotope permeation resistance. In this work, we report a new strategy to improve the permeation resistance by oxygen ion implantation. A (TiVAlCrZr)O multi-component metal oxide film was prepared as a tritium permeation barrier on a 316 L substrate by RF magnetron sputtering deposition, and then implanted by oxygen ions to
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Manufacturing Oxide Dispersion Strengthened (ODS) steel plate via cold spray and friction stir processing J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-08 Xiang Wang, Dalong Zhang, Jens T. Darsell, Kenneth A. Ross, Xiaolong Ma, Jia Liu, Tingkun Liu, Ramprashad Prabhakaran, Lan Li, Iver E. Anderson, Wahyu Setyawan
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First-principles investigation of the interaction between oxygen and alloy atoms in [formula omitted]-zirconium J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-07 Guodong Lu, Zhixiao Liu, Wangyu Hu, Tianguo Wei, Yi Zhao, Dong Wang, Huiqiu Deng
Zirconium alloys, being crucial materials in the nuclear industry, frequently encounter issues related to oxidative corrosion. Understanding the interaction between O atom and alloying atoms at the atomistic scale is helpful for optimizing Zr alloys with improved corrosion resistance. In the present study, a first-principles approach is used to understand the stability and diffusion properties of an
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High energy radiation tolerance of iron phosphate glasses: Molecular dynamics study J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-05 C. Cockrell, K. Joseph, M.K. Patel, R.W. Grimes, K. Trachenko
We report the results of massive parallel molecular dynamics simulations of high-energy radiation damage in phosphate glasses. This damage is created by overlapping multiple 70 keV collision cascades. We quantify different aspects of radiation-induced structural changes including at different stages of damage development, including coordination numbers, cluster sizes and density. The overall trend
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Radiation-induced effects in self-passivating W-Cr-Y alloy J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-03 O.V. Ogorodnikova, A.A. Nikitin, S.V. Rogozkin, E. Sal, C. García-Rosales, Yu.M. Gasparyan, V. Gann
Irradiation of the promising self-passivating W-10wt%Cr-0.5wt%Y alloy suggested as the first wall material for demonstration power plant, DEMO, was performed with 5.9 MeV Co ions at 300 °C and 5.6 MeV Fe ions at 500 °C to a dose peak of 12 dpa (displacements per atom). Atom probe tomography and transmission electron microscopy were used to study the chemical compositions of this alloy at the atomic
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Surface damage of refractory high entropy alloys subject to He irradiation J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-03 Yaoxu Xiong, Kun Wang, Shijun Zhao
Tungsten is a leading material considered for nuclear fusion applications. However, W usually suffers from severe surface modification under exposure to He ions generated from fusion reactions. The development of high entropy alloys provides a feasible way to overcome such vulnerabilities. Herein, we explore the response of refractory multi-principal element alloys (MPEAs), WTaTi, WTaCrV, and WTaCrVTi
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Sintering behavior analysis of compacted dry recycled U0.7Pu0.3O2 powder using master sintering curve theory J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-03 Shinya Nakamichi, Takeo Sunaoshi, Shun Hirooka, Romain Vauchy, Tatsutoshi Murakami
Using dry recycled powders for uranium and plutonium mixed oxide (MOX) fuel production can reduce unnecessary storage and accountability of nuclear material in facilities. The shrinkage behavior of green compacts of dry recycled powders differs from that of conventional raw powders because the dry recycled MOX powder is obtained from the fabrication scrap of sintered pellets. The shrinkage behavior
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Corrosion behaviors of Zr-Sn-Nb alloys: Influence of different concentration dissolved oxygen environment and trace V element on the anti-corrosion properties J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-03 Xutong Wang, Jiuxiao Li, Zhiwei Zhao, Yixiao Yu, Huigang Shi
Zr-0.4Sn-0.7Nb-0.3Fe-0.1Cr-0.15Mo-0.12O-V, =0, 0.05 wt% alloys were designed to explore the effect of water with pure water and 1000 ppb dissolved oxygen on the corrosion behavior of the alloys. Results showed that dissolved oxygen can accelerate the corrosion of the alloys. The V-containing alloy possesses improved corrosion resistance in the DO environment, which can be attributed to the enhancement
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Physical simulation of the underclad heat affected zone in a reactor pressure vessel to study intergranular cracking J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-02 Alessandro Cattivelli, Mary Grace Burke, Jean Dhers, John Anthony Francis
Underclad cracking in nuclear pressure vessels was of significant concern in the 1970s and 1980s before mitigating adjustments were made both to steel compositions and manufacturing practice. Unfortunately, the cracking mechanisms are still not well understood, and this can undermine confidence when changes to cladding operations are under consideration. In this work, Charpy-sized test coupons of 18MND5
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Immobilization of hexavalent uranium U(VI) by hydroxyapatite under oxic conditions J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-02 Seoha Kim, Yongmoon Lee, Minji Park, Hoon Young Jeong
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Corrosion behavior and mechanical properties of V-4Cr-4Ti alloy exposed at 500 and 700 °C to static Pb with ∼10-9mass% dissolved oxygen for 1000 h J. Nucl. Mater. (IF 3.1) Pub Date : 2024-04-01 Valentyn Tsisar, Takuya Nagasaka, Olga Yeliseyeva, Jun Lim, Jürgen Konys, Takeo Muroga, Carsten Schroer
The corrosion tests of V-4Ti-4Cr alloy (NIFS HEAT 2) tensile samples were carried out in static liquid Pb with ∼10 mass% dissolved oxygen at 500 and 700 °C for 1000 h. A clear intergranular corrosion attack was observed on the surface of the samples. Corrosion losses increase ten times, from around 3.0 µm to 30 µm, as the temperature increased from 500 °C to 700 °C, respectively. The alloy remained
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On the stable/metastable nature of the γ-hydride phase in Zircaloy-2: Microstructural characterization by electron diffraction, electron energy-loss spectroscopy, and diffraction line profile analysis J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-29 N.N. Badr, F. Long, T. Lucas, Y. Luo, M. Topping, L. Balogh, L.K. Béland, Z. Yao, G. King, M.R. Daymond
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Raman spectroscopy of uranium nitride kernels J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-29 Eddie Lopez-Honorato, Liangbo Liang, J. Matthew Kurley, Katherine Montoya, William F. Cureton, Rachel Siebert, Rodney D. Hunt, Nathan Capps, Andrew T. Nelson
Uranium nitride is an advanced fuel candidate for a wide variety of advanced nuclear reactors. This work summarizes the first characterization of UN kernels by Raman spectroscopy. First-principles density functional theory calculations were performed to predict the Raman spectra of uranium sesquinitride (UN), uranium dinitride (UN), uranium mononitride (UN), uranium monocarbide (UC), as well as U-N-C
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First-principles analysis of intergranular fracture in UN by Σ5(210) grain boundary segregated Xe and vacancy J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-29 Yuanhai Jiang, Jiajun Zhao, Liu Xi, Jijun Zhao, Yuanyuan Wang
Segregation of irradiation-generated xenon (Xe) atoms and vacancies to grain boundary (GB) leads to instability and brittleness of uranium mononitride (UN). However, theoretical knowledge about the energetics of its defective interface is lacking, whereas there has been considerable attention on unirradiated UN in experiments. Using first-principles methods, herein we compute the energies of segregation
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Temperature-Dependent Mechanical Anisotropy in Textured Zircaloy Cladding J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-29 Malachi Nelson, Shmuel Samuha, David Kamerman, Peter Hosemann
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Molecular dynamics simulation study of void collapse mechanisms in cubic metals under shock compression J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-28 Yuanfang Lu, Hongxian Xie, Guang-Hong Lu
Molecular dynamics simulations have revealed the collapse mechanism of a void in cubic metals under shock compression along [100] direction. The results show that the void collapse is caused by emitting dislocations from its surface. The type of cubic metal plays a decisive role in the subsequent microstructural evolution around the void. Void in face-centered-cubic metals collapses by emitting Shockley
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Mesoscale modeling of the effects of accelerated burnup on UO2 microstructural evolution J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-28 Amani Cheniour, Floyd W. Hilty, Christian M. Petrie, Nathan A. Capps
Accelerating the nuclear fuel qualification process will rely on some combination of advanced modeling and simulation techniques with accelerated irradiation testing and separate effects experiments to enable the development of new fuel concepts in a shorter time frame. One of the key challenges to successfully leveraging accelerated irradiation tests will be understanding the artifacts that may be
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A facility for studying corrosion via in-situ Raman spectroscopy J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-27 V.S. Ramsundar, K. Daub, S.Y. Persaud, M.R. Daymond
Several in-core components in nuclear power systems are exposed to high-temperature water in the presence of radiation fields. The dynamic effect of radiation and water chemistry on material performance in these environments is not well understood partly due to significant experimental challenges. A facility consisting of a high temperature/pressure corrosion loop coupled with Raman spectroscopy has
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Predicting and visualizing crack propagation in nuclear graphite J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-27 Gongyuan Liu, Khalid Hattar, William Windes, Aman Haque, Jing Du
Nuclear graphite exhibits a complex fracture behavior due to the presence of defects ranging from nanoscale basal cracks to sub-millimeter scale voids as well as the heterogeneity of constituents. This study aims to develop insights on whether finite element (FE) modeling can effectively reproduce fracture behavior observed in the experiments coupled with micro X-ray computed tomography (CT). Two-phase
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Thermophysical characterization of UFe3B2 and USiNi: An experimental study J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-27 Yifan Sun, Yuji Miyawaki, Masaya Kumagai, Shun Fujieda, Hiroaki Muta, Ken Kurosaki, Yuji Ohishi
In the development for advanced nuclear fuels for LWRs, the focus has traditionally centered on a limited array of materials like doped-UO, UN, and USi, leaving the vast majority of potential uranium compounds unexplored. To expand the search and expedite the discovery of advanced LWR fuels, we developed a machine learning (ML) classification model dedicated to identifying uranium compounds with excellent
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Blistering and retention behavior of laser powder bed fused tungsten alloys under hydrogen plasma irradiation J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-26 Kailun Li, Baorui Du, Li Yu, Dianzheng Wang, Haodong Liu, Hiashan Zhou, Guangnan Luo, Jun Yao, T.W. Morgan, Wei Liu, Wanqi Chen
In this study, pure tungsten (W) fabricated by powder metallurgy and laser powder bed fusion (LPBF) formed W-Ta and W-ZrC alloys were exposed to hydrogen plasma. Severe blistering was found in LPBFed W-Ta specimens due to a strong [111] crystallographic texture and the large number of particle-matrix interfaces in the LPBFed W-ZrC inhibits the blister growth. The retention of LPBFed W alloys is lower
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Effects of Gamma-ray and neutron irradiation on infrared optical properties of ZnSe and ZnS for ITER divertor thermography J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-26 Tomohiko Ushiki, Ryota Imazawa, Sin-Ichi Kitazawa, Masao Ishikawa, Hidetoshi Murakami, Kosuke Shimizu, Tatsuo Sugie, Hiroyuki Okazaki, Masaya Seki, Butch Buenavidez, Naoto Kasano, Yuzi Katayanagi, Yoshihiko Nunoya
Gamma-ray and neutron irradiation tests were performed by using antireflective-coated and non-antireflective-coated samples of ZnSe and ZnS to examine the impact on their infrared optical properties. In neutron irradiation tests, ZnSe and ZnS samples showed minimal transmittance reduction even up to 1.44 × 10 n/cm (evaluated by dpa value, the values are 5.77 × 10, 2.07 × 10 for ZnSe and ZnS, respectively)
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Stress-corrosion cracking sensitization by hydrogen upon oxidation of nickel-base alloys by water – An experiment-guided first-principles study J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-24 Ageo Meier de Andrade, Christine Geers, Jiaxin Chen, Itai Panas
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Magnetism and finite-temperature effects in UZr2: A density functional theory analysis J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-24 Shehab Shousha, Benjamin Beeler
The structure and the thermophysical properties of -UZr are investigated using 0 K density functional theory and molecular dynamics (AIMD). Modeling the true paramagnetic state of this intermetallic compound has been challenging using first-principles calculations. For the first time, we find that the generalized gradient approximation method without applying an on-site Coulomb interaction term (Hubbard
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Ab-initio study of point defects in Th and U alloy J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-22 Jacob Startt, Chaitanya Deo
Point defect formation energies are calculated along with several structural and elastic properties for the () and () Th crystal structures. Th-self defects and single uranium impurity defects are modeled in the phase, whereas only Th self-defects are modeled in the high temperature phase. Additionally, an analysis of several well known exchange-correlation functionals is given in regard to defect
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The lattice contraction of UO2 from Cr doping as determined via high resolution synchrotron X-ray powder diffraction J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-21 Gabriel L. Murphy, Volodymyr Svitlyk, Maximilian Henkes, Daniil Shirokiy, Christoph Hennig, Philip Kegler, Dirk Bosbach, Andrey Bukaemskiy
High resolution synchrotron powder X-ray diffraction analysis of Cr-doped UO samples with additions of CrO as 0, 500, 1000, 1500, 2500 to 3500 ppm prepared under sintering conditions of -420 kJ/mol and 1700 °C is reported. The lattice dependence from Cr doping is established through the Rietveld refinement method where the rate of linear lattice parameter contraction from Cr doping, , was found to
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Impact of grain boundary and surface diffusion on predicted fission gas bubble behavior and release in UO2 fuel J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-21 Md Ali Muntaha, Sourav Chatterjee, Sophie Blondel, Larry Aagesen, David Andersson, Brian D. Wirth, Michael R. Tonks
In this work, we quantify the impact of grain boundary (GB) and surface diffusion on fission gas bubble evolution and fission gas release in UO nuclear fuel using simulations with a hybrid phase field/cluster dynamics model. We begin with a comprehensive literature review of uranium vacancy and xenon atom diffusivity in UO through the bulk, along GBs, and along surfaces. In our model we represent fast
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Effect of the triangular prism dent on stress corrosion cracking behavior of alloy 690TT heat transfer tube in a lead-containing alkaline solution J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-20 Yun Ding, Sui Yuan, Renquan Wu, Shichen Wei, Shuo Wang, Jian Xu, Hongying Yu, Dongbai Sun
In this paper, the microstructural changes and mechanical characteristics of alloy 690TT heat transfer tube with a triangular prism dent are investigated by combining experimental and simulation methods, and subsequently examines their influence on stress corrosion cracking (SCC) behavior. The results indicate that the deformation mechanism caused by impact dent is slip and twinning. The structural
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An FFT based approach to account for elastic interactions in OkMC: Application to dislocation loops in iron J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-20 Rodrigo Santos-Güemes, Christophe J. Ortiz, Javier Segurado
Object kinetic Monte Carlo (OkMC) is a fundamental tool for modeling defect evolution in volumes and times far beyond atomistic models. The elastic interaction between defects is classically considered using a dipolar approximation but this approach is limited to simple cases and can be inaccurate for large and close interacting defects. In this work a novel framework is proposed to include elastic
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Investigation on microstructural evolution and strengthening mechanism in Zr-2.5Nb alloy with multi-scale lamellar structure J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-19 Kangkai Song, Conghui Zhang, Xunliang Zhang, Guodong Song, Wenguang Zhu, Tongguang Zhai, Xuan Zhou
Zr-2.5Nb alloy was subjected to a solid solution treatment at 900 °C for 30 min, followed by aging between 500 and 700 °C for 60 min. It was found that there were two distinct martensitic microstructures after the solid solution treatment: nano-twins and dislocation lamellar structures. As the aging temperature increased, the α’ martensite lath gradually decomposed, which was mainly manifested in the
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High-velocity dust impacts in plasma facing materials: Insights from molecular dynamics simulations J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-19 Prashant Dwivedi, Alberto Fraile, Tomas Polcar
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A new interatomic potential of W-Ni-Fe systems for point defects and mechanical property studies J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-19 Xichuan Liao, Yangchun Chen, Rongyang Qiu, Yong Liu, Ning Gao, Fei Gao, Wangyu Hu, Huiqiu Deng
W-Ni-Fe tungsten heavy alloys (WHAs) are a kind of ductile phase toughened composites that are becoming increasingly interesting as promising alternatives to monolithic tungsten for fusion reactor plasma-facing materials, because of their outstanding combination of strength, ductility and toughness. Understanding the radiation-induced microstructural features of WHAs associated with their mechanical
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Development of a genetic algorithm based interatomic potential and application in thermal conductivity study of ThO2 grain boundaries J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-19 Shihui Ning, Hongjian Chen, Dingwang Yuan, Wangyu Hu, Bowen Huang
ThO is a promising fuel for next-generation reactors due to its superior thermo-physical properties. In this study, we present a Coulomb-Buckingham-Morse interatomic potential for ThO, developed using a homemade genetic algorithm (GA) method. The potential accurately captures fundamental properties, such as lattice constants, elastic constants, point defect energies, phonon spectra, and thermal conductivity
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Chlorination mechanism of typical metallic radionuclides by HCl on irradiated graphite surface from first principles J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-19 Kun Fu, Meiqian Chen
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Interaction between Xe bubbles and grain boundaries as well as the influences on structural evolution in UO2: A molecular dynamics simulation J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-19 Hui Ma, Danmin Peng, Hongwei Bao, Zhipeng Sun, Jibin Zhang, Fei Ma
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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
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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
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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
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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
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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
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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
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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
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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
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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
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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
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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.
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Transmission electron microscopy study of the radiation-induced amorphization in the Cr-W-C ternary M23C6 J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-11 Sho Kano, Huilong Yang, Masami Ando, Dai Hamaguchi, Takashi Nozawa, Hiroyasu Tanigawa, Tamaki Shibayama, Hiroaki Abe
To provide mechanistic insight into the phase stability of MC under irradiation, bulk MC (Cr-W-C system) specimens with a W concentration range of 0–12 at.% were first subjected to helium ion irradiation and then investigated by systematic TEM/STEM observation. The atomic resolution electron microscopy analysis revealed that the preferential occupation site of the W atom is the 8c site, and the 48
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Fabrication and characterization of zirconium oxide coating on tubular steel by metal organic decomposition J. Nucl. Mater. (IF 3.1) Pub Date : 2024-03-11 Khiem Do Duy, Ryosuke Norizuki, Hikaru Fujiwara, Kento Shirota, Teruya Tanaka, Juro Yagi, Marcin Rasinski, Takumi Chikada
In fusion reactor blankets, multifunctional ceramic coatings fabricated on structural materials have been investigated for tritium permeation reduction, electrical insulation, and corrosion protection. However, it has been challenging to manufacture a homogeneous coating on the complicated components in the blanket breeder, such as ducts. In terms of sustainable development, it is crucial to establish
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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
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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