Journal of Nuclear Materials

Papers
(The TQCC of Journal of Nuclear Materials is 8. The table below lists those papers that are above that threshold based on CrossRef citation counts [max. 250 papers]. The publications cover those that have been published in the past four years, i.e., from 2020-03-01 to 2024-03-01.)
ArticleCitations
High temperature Cr-Zr interaction of two types of Cr-coated Zr alloys in inert gas environment74
The role of TeO2 insertion on the radiation shielding, structural and physical properties of borosilicate glasses70
Additive manufacturing of silicon carbide for nuclear applications53
Mechanical properties of tungsten: Recent research on modified tungsten materials in Japan53
Cr-coated Zr-4 alloy prepared by electroplating and its in situ He+ irradiation behavior51
Lead-bismuth eutectic (LBE) corrosion behavior of AlTiN coatings at 550 and 600゜C49
A review of the LIBS analysis for the plasma-facing components diagnostics49
Mechanical and chemical properties of PVD and cold spray Cr-coatings on Zircaloy-446
Radiation damage tolerance of a novel metastable refractory high entropy alloy V2.5Cr1.2WMoCo0.0445
Uranium nitride tristructural-isotropic fuel particle44
Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding44
Efficient capture of radioactive iodine by a new bismuth-decorated electrospinning carbon nanofiber40
Effects of contents of Al, Zr and Ti on oxide particles in Fe–15Cr–2W–0.35Y2O3 ODS steels40
SCIANTIX: A new open source multi-scale code for fission gas behaviour modelling designed for nuclear fuel performance codes40
Evaluation of Equivalent Cladding Reacted parameters of Cr-coated claddings oxidized in steam at 1200 °C in relation with oxygen diffusion/partitioning and post-quench ductility39
Highly effective prussian blue-coated MXene aerogel spheres for selective removal of cesium ions39
A review of in-pile fuel safety tests of TRISO fuel forms and future testing opportunities in non-HTGR applications38
TRISO particle fuel performance and failure analysis with BISON37
Cold spray deposition of 304L stainless steel to mitigate chloride-induced stress corrosion cracking in canisters for used nuclear fuel storage36
Review of manufacturing technologies for coated accident tolerant fuel cladding36
Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan35
High-heat-flux technologies for the European demo divertor targets: State-of-the-art and a review of the latest testing campaign35
Design considerations for high entropy alloys in advanced nuclear applications35
Perspectives on multiscale modelling and experiments to accelerate materials development for fusion35
Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: Uranium mononitride33
Thermal aging behaviors of duplex stainless steels used in nuclear power plant: A review33
Impact of neutron irradiation on the strength and ductility of pure and ZrC reinforced tungsten grades31
Fracture-mechanical properties of neutron irradiated ITER specification tungsten31
A study of the oxidation behaviour of FeCrAl-ODS in air and steam environments up to 1400 °C31
Fission gas diffusion and release for Cr2O30
High-temperature oxidation and quenching of chromium-coated zirconium alloy ATF cladding tubes with and w/o pre-damage30
The structure of molten FLiNaK30
Cluster dynamics simulation of xenon diffusion during irradiation in UO230
Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation30
Hydrogen in zirconium: Atomistic simulations of diffusion and interaction with defects using a new embedded atom method potential30
Fabrication of yttrium hydride for high-temperature moderator application29
U3Si2 and UO2 composites densified by spark plasma sintering for accident-tolerant fuels29
A threshold density of helium bubbles induces a ductile-to-brittle transition at a grain boundary in nickel28
Revealing the synergistic effects of sequential and simultaneous dual beam irradiations in tungsten via in-situ TEM28
α-U and ω-UZr2 in neutron irradiated U-10Zr annular metallic fuel28
Molecular dynamics simulations of high-dose damage production and defect evolution in tungsten28
STEM Characterization of Dislocation Loops in Irradiated FCC Alloys27
Correlation of microstructural and mechanical properties of neutron irradiated EUROFER97 steel27
Effect of brazing temperature, filler thickness and post brazing heat treatment on the microstructure and mechanical properties of W-Eurofer joints brazed with Cu interlayers27
Evaluation of the effects of neutron irradiation on first-generation corrosion mitigation coatings on SiC for accident-tolerant fuel cladding27
Vacancy cluster growth and thermal recovery in hydrogen-irradiated tungsten27
Effect of Ce4+-substitution at A and B sites of Nd2Zr2O7: A study for plutonium incorporation in pyrochlores26
Ferrite formation and its effect on deformation mechanism of wire arc additive manufactured 308 L stainless steel26
Processing of tungsten through electron beam melting26
A systematic investigation of the phase assemblage and microstructure of the zirconolite CaZr1-xCexTi2O7 system26
Synthesis and characterization of zirconolite-sodium borosilicate glass-ceramics for nuclear waste immobilization25
Crystalline phosphates for HLW immobilization - composition, structure, properties and production of ceramics. Spark Plasma Sintering as a promising sintering technology25
Visualization of hydrogen isotope distribution in yttrium and cobalt doped barium zirconates25
A refined oxidation mechanism proposed for ferritic-martensitic steels exposed to oxygen-saturated liquid lead-bismuth eutectic at 400°C for 500 h25
Atomistic simulation study of clustering and evolution of irradiation-induced defects in zirconium24
Helium-induced damage behavior in high temperature nickel-based alloys with different chemical composition24
Recrystallization at high temperature of two tungsten materials complying with the ITER specifications24
A Critical Review of High Burnup Fuel Fragmentation, Relocation, and Dispersal under Loss-Of-Coolant Accident Conditions24
Effect of radiation damage and water radiolysis on corrosion of FeCrAl alloys in hydrogenated water24
Design and strategy for next-generation silicon carbide composites for nuclear energy24
Impact of microstructural properties on hardness of tungsten heavy alloy evaluated by stand-off LIBS after PSI plasma irradiation24
Microstructural examination of neutron, proton and self-ion irradiation damage in a model Fe9Cr alloy24
Using external ion irradiations for simulating self-irradiation damage in nuclear waste glasses: State of the art, recommendations and, prospects24
Electronic stopping in molecular dynamics simulations of cascades in 3C–SiC23
Sputtering of nanostructured tungsten and comparison to modelling with TRI3DYN23
Oxidation and anion lattice defect signatures of hypostoichiometric lanthanide-doped UO223
Effect of helium bubbles on irradiation hardening of additive manufacturing 316L stainless steel under high temperature He ions irradiation23
DLI-MOCVD CrxCy coating to prevent Zr-based cladding from inner oxidation and secondary hydriding upon LOCA conditions23
On the irradiation tolerance of nano-grained Ni–Mo–Cr alloy: 1 MeV He+ irradiation experiment23
Development of CuCrZr via Electron Beam Powder Bed Fusion (EB-PBF)23
Tensile properties and microstructure of additively manufactured Grade 91 steel for nuclear applications23
Effects of carbon doping on irradiation resistance of Fe38Mn40Ni11Al4Cr7 high entropy alloys22
Architecture and properties of TCR fuel form22
First steps toward predicting corrosion behavior of structural materials in molten salts22
Heavy-ion irradiation effects of Gd2Zr2O7 nanocrystalline ceramics as nuclear waste immobilization matrix22
Results of the QUENCH-LOCA experimental program at KIT22
Protection of graphite from salt and gas permeation in molten salt reactors22
Creep behavior of 316 L stainless steel manufactured by laser powder bed fusion22
Correlation between microstructure and mechanical properties in the age-hardenable Cu-Cr-Zr alloy22
Radiation effects on structure and mechanical properties of borosilicate glasses22
Uniform corrosion of FeCrAl cladding tubing for accident tolerant fuels in light water reactors22
Helium bubble formation in refractory single-phase concentrated solid solution alloys under MeV He ion irradiation22
Analysis of fuel rod behavior during loss-of-coolant accidents using the BISON code: Cladding modeling developments and simulation of separate-effects experiments21
Comparison of tritium release behavior in Li2TiO3 and promising core-shell Li2TiO3–Li4SiO4 biphasic ceramic pebbles21
Recent progress on preparation routes and performance evaluation of ODS/CDS-W alloys for plasma facing materials in fusion devices21
Neutron irradiation tolerance of potassium-doped and rhenium-alloyed tungsten21
Corrosion studies of a low alloyed Fe–10Cr–4Al steel exposed in liquid Pb at very high temperatures21
Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions21
Thermal annealing and transformation of dimer F centers in neutron-irradiated Al2O3 single crystals21
Accelerating nuclear fuel development and qualification: Modeling and simulation integrated with separate-effects testing21
Charge compensation mechanisms in Nd-doped UO2 samples for stoichiometric and hypo-stoichiometric conditions: Lack of miscibility gap21
Structural and compositional effects on the electronic excitation induced phase transformations in Gd2Ti2-yZryO7 pyrochlore21
Raman spectroscopy coupled to principal component analysis for studying UO2 nuclear fuels with different grain sizes due to the chromia addition21
Microstructures and mechanical properties of the novel CuCrZrFeTiY alloy for fusion reactor20
Quantifying the effect of hydride microstructure on zirconium alloys embrittlement using image analysis20
Large-scale potassium-doped tungsten alloy with superior recrystallization resistance, ductility and strength induced by potassium bubbles20
Irradiation effects on Al0.3CoCrFeNi and CoCrMnFeNi high-entropy alloys, and 316H stainless steel at 500 °C20
Dislocation loop evolution and radiation hardening in nickel-based concentrated solid solution alloys20
Effects of carbonitrides and carbides on microstructure and properties of castable nanostructured alloys20
B2O3-assisted low-temperature crystallization of pollucite structures and their potential applications in Cs+ immobilization20
Corrosion of commercial alloys in FLiNaK molten salt containing EuF3 and simulant fission product additives20
Brittle-ductile transition temperature of recrystallized tungsten following exposure to fusion relevant cyclic high heat load19
Investigation of the irradiation effects in additively manufactured 316L steel resulting in decreased irradiation assisted stress corrosion cracking susceptibility19
Evolution of vacancy defects in heavy ion irradiated tungsten exposed to helium plasma19
Non-Destructive post-irradiation examination results of the first modern fueled experiments in TREAT19
Oxidation mechanism and kinetics of nuclear-grade FeCrAl alloys in the temperature range of 500–1500 °C in steam19
10B(n, α)7Li reaction-induced gas bubble formation in Al–B4C neutron absorber irradiated in spent nuclear fuel pool19
Nano-hardening features in high-dose neutron irradiated Eurofer97 revealed by atom-probe tomography19
Modeling high burnup structure in oxide fuels for application to fuel performance codes. part I: High burnup structure formation19
The effect of Zr on precipitation in oxide dispersion strengthened FeCrAl alloys19
Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding19
Effects of temperature on helium bubble behaviour in Fe–9Cr alloy19
Estimation of reliable displacements-per-atom based on athermal-recombination-corrected model in radiation environments at nuclear fission, fusion, and accelerator facilities19
Atomistic simulations to study the effect of helium nanobubble on the shear deformation of nickel crystal18
Hot deformation behavior and processing map of Zr-4 alloy18
Effects of dissolved oxygen on partial slip fretting corrosion of Alloy 690TT in high temperature pure water18
Ni coating on 316L stainless steel using cage plasma treatment: Feasibility and swelling studies18
High temperature zirconium alloys for fusion energy18
Application of machine learning in understanding the irradiation damage mechanism of high-entropy materials18
The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel18
Microstructure, oxidation and corrosion properties of FeCrAl coatings with low Al content prepared by magnetron sputtering for accident tolerant fuel cladding18
Iodosodalite synthesis with hot isostatic pressing of precursors produced from aqueous and hydrothermal processes18
High temperature steam oxidation dynamics of U3Si2 with alloying additions: Al, Cr, and Y18
Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants18
Gamma irradiation-induced defects in borosilicate glasses for high-level radioactive waste immobilisation18
Comparison of K-doped and pure cold-rolled tungsten sheets: Tensile properties and brittle-to-ductile transition temperatures18
Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management18
Post-decontamination treatment of MXene after adsorbing Cs from contaminated water with the enhanced thermal stability to form a stable radioactive waste matrix18
A critical review of the microstructure of U–Mo fuels18
Characterization of as-deposited cold sprayed Cr-coating on Optimized ZIRLO™ claddings18
Interfacial characterization of dissimilar-metals bonding between vanadium alloy and Hastelloy X alloy by explosive welding18
Accelerated/reduced growth of tungsten fuzz by deposition of metals17
Uranium nitride (UN) pellets with controllable microstructure and phase – fabrication by spark plasma sintering and their thermal-mechanical and oxidation properties17
Modeling fission product diffusion in TRISO fuel particles with BISON17
Significant suppression of void swelling and irradiation hardening in a nanograined/nanoprecipitated 14YWT-ODS steel irradiated by helium ions17
Probing the Short-Range Ordering of Ion Irradiated Gd2Ti2-yZryO7 (0.0 ≤ y ≤ 2.0) Pyrochlore under Electronic Stopping Regime17
Liquid metal embrittlement sensitivity of the T91 steel in lead, in bismuth and in lead-bismuth eutectic17
Interatomic potentials of W–V and W–Mo binary systems for point defects studies17
Oxidation properties and microstructure of a chromium coating on zircaloy-4 fuel cladding material applied by atmospheric plasma spraying17
Qualification pathways for additively manufactured components for nuclear applications17
The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC17
Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding17
Microstructural evolution of Mo-UO2 cermets under high temperature hydrogen environments17
Molecular dynamics simulation of primary radiation damage in W-Ta alloys: Effect of tantalum17
Characterizing microstructural evolution and low cycle fatigue behavior of 316H austenitic steel at high-temperatures17
Study of radiation-induced amorphization of M23C6 in RAFM steels under iron irradiations17
Irradiation effects on the fracture properties of UO2 fuels studied by micro-mechanical testing16
Strengthening mechanism of Nb addition in Fe–13Cr–4.5Al–2Mo alloys assessed by internal friction measurement16
Heavy ion irradiation response of an additively manufactured 316LN stainless steel16
Modelling and assessment of thermal conductivity and melting behaviour of MOX fuel for fast reactor applications16
Primary radiation damage characteristics in displacement cascades of FeCrAl alloys16
TEM characterization of dislocation loops in proton irradiated single crystal ThO216
Uranium carbide properties for advanced fuel modeling – A review16
Study of the corrosion characteristics of 304 and 316L stainless steel in the static liquid lithium16
Effects of TiC nanoparticle additions on microstructure and mechanical properties of FeCrAl alloys prepared by directed energy deposition16
Post-LOCA ductility of Cr-coated cladding and its embrittlement limit16
Influence of surface roughness on the sputter yield of Mo under keV D ion irradiation16
Tensile properties of powder-metallurgical-processed tungsten alloys after neutron irradiation near recrystallization temperatures16
Hydrothermal Corrosion of First-Generation Dual-Purpose Coatings on Silicon Carbide for Accident-Tolerant Fuel Cladding16
Interatomic potentials and defect properties of Fe–Cr–Al alloys16
In-situ TEM investigation of nano-scale helium bubble evolution in tantalum-doped tungsten at 800°C16
Dynamic evolution of He bubble and its effects on void nucleation-growth and thermomechanical properties in the spallation of aluminum16
High temperature mechanical properties of fluorite crystal structured materials (CeO2, ThO2, and UO2) and advanced accident tolerant fuels (U3Si2, UN, and UB2)16
Effect of stacking fault energy on damage microstructure in ion-irradiated CoCrFeNiMn concentrated solid solution alloys16
Bubbles and precipitates formation and effects on the hardening of irradiated vanadium alloys16
The kinetics of dynamic recrystallization and construction of constitutive modeling of RAFM steel in the hot deformation process16
A logical approach for zero-rupture Fully Ceramic Microencapsulated (FCM) fuels via pressure-assisted sintering route16
Spark plasma sintered tungsten – mechanical properties, irradiation effects and thermal shock performance16
Hydrogen and its detection in fusion and fission nuclear materials – a review16
Phase-field simulations of intergranular fission gas bubble behavior in U3Si2 nuclear fuel16
Positron annihilation spectroscopy study of vacancy-type defects in He implanted polycrystalline α-SiC16
Investigation of spatial relationship between helium bubbles and dislocation loops in RAFM steel16
Critical behavior of interfacial t-ZrO2 and other oxide features of zirconium alloy reaching critical transition condition16
Influence of the transient conditions on release of corrosion products and oxidation of alloy 690 tubes during pressurized water reactor restart after steam generators replacement16
Deuterium trapping in the subsurface layer of tungsten pre-irradiated with helium ions16
Modeling intra-granular fission gas bubble evolution and coarsening in uranium dioxide during in-pile transients16
Grand potential sintering simulations of doped UO2 accident-tolerant fuel concepts16
Interatomic potentials for irradiation-induced defects in iron16
Mechanical properties of neutron-irradiated single crystal tungsten W(100) studied by indentation and FEM modelling16
Benefit or harm of accident tolerant coatings on the low-cycle fatigue properties of Zr-4 cladding alloy: in-situ studies at 400°C16
Modeling reactivity insertion experiments of TRISO particles in NSRR using BISON15
Comparative study of deuterium retention and vacancy content of self-ion irradiated tungsten15
UN microspheres embedded in UO2 matrix: An innovative accident tolerant fuel15
Characterisation of open volume defects in Fe–Cr and ODS Fe–Cr alloys after He+ and Fe+ ion irradiations15
Galvanic corrosion study between low alloy steel A508 and 309/308 L stainless steel dissimilar metals: A case study of the effects of oxide film and exposure time15
Oxidation of UN/U2N3-UO2 composites: an evaluation of UO2 as an oxidation barrier for the nitride phases15
Validating modern methods for impurity analysis in fluoride salts15
Helium bubble nucleation in Laser Powder Bed Fusion processed 304L stainless steel15
Study of thermodynamic properties of U1-yPuyO2 MOX fuel using classical molecular Monte Carlo simulations15
Microstructural evolution of Cr-coated Zr-4 alloy prepared by multi-arc ion plating during high temperature oxidation15
Raman and infrared spectra of plutonium (IV) oxalate and its thermal degradation products15
High-throughput ion irradiation of additively manufactured compositionally complex alloys15
Beyond U/Pu separation: Separation of americium from the highly active PUREX raffinate15
Deuterium retention in W and binary W alloys irradiated with high energy Fe ions15
Effects of minor alloying addition on He bubble formation in the irradiated FeCoNiCr-based high-entropy alloys15
Tensile behavior of dual-phase titanium alloys under high-intensity proton beam exposure: Radiation-induced omega phase transformation in Ti-6Al-4V15
Simulation of the chemical state of high burnup (U15
Microstructural and crystallographic analysis of hydride reorientation in a zirconium alloy cladding tube15
Microstructures and mechanical properties of a modified 9Cr ferritic-martensitic steel in the as-built condition after additive manufacturing15
Thermal and mechanical properties of U3Si2: A combined ab-initio and molecular dynamics study15
Synthesis and characterization of iron phosphate based glass-ceramics containing sodium zirconium phosphate phase for nuclear waste immobilization15
Radiation damage in uranium dioxide: Coupled effect between electronic and nuclear energy losses15
Performance of U3Si2 in an LWR following a cladding breach during normal operation14
Determining uranium ore concentrates and their calcination products via image classification of multiple magnifications14
Irradiation resistance of chromium coatings for ATFC in the temperature range 300–550°C14
Short positron lifetime at vacancies observed in electron-irradiated tungsten: Experiments and first-principles calculations14
Spent nuclear fuel in dry storage conditions – current trends in fuel performance modeling14
Low leaching characteristics and encapsulation mechanism of Cs+ and Sr2+ from SAC matrix with radioactive IER14
Corrosion behavior of ferritic FeCrAl alloys in simulated BWR normal water chemistry14
High spatial resolution thermal conductivity mapping of SiC/SiC composites14
TRANOX: Model for non-isothermal steam oxidation of zircaloy cladding14
Microstructure degradation of austenitic stainless steels after 45 years of operation as VVER-440 reactor internals14
Modeling W fuzz growth over polycrystalline W due to He ion irradiations at an elevated temperature14
Fluctuations in stacking fault energies improve irradiation tolerance of concentrated solid-solution alloys14
Hydrothermal corrosion behavior of CVD SiC in high temperature water14
Insight into interface cohesion and impurity-induced embrittlement in carbide dispersion strengthen tungsten from first principles14
Influence of Ni-Mn contents on the embrittlement of PWR RPV model steels irradiated to high fluences relevant for LTO beyond 60 years14
Collapse of stacking fault tetrahedron and dislocation evolution in copper under shock compression14
In-situ studies on the mechanical properties of He ion irradiated nanotwinned Ag14
A novel methodology for estimating tensile properties in a small punch test employing in-situ DIC based deflection mapping14
Radiation damage buildup and basal vacancy cluster formation in hcp zirconium: A molecular dynamics study14
Effect of cooling rate on the residual ductility of Post-LOCA Zircaloy-4 cladding14
Helium ion irradiation enhanced precipitation and the impact on cavity formation in a HfNbZrTi refractory high entropy alloy14
Embedded sensors in additively manufactured silicon carbide14
Determining the acute oxidation behavior of several nuclear graphite grades14
A comparison study of void swelling in additively manufactured and cold-worked 316L stainless steels under ion irradiation14
Microstructure and tensile behavior of powder metallurgy FeCrAl accident tolerant fuel cladding14
Mechanisms of plastic deformation and fracture of austenitic chromium-nickel steel irradiated during 45 years in WWER-44014
The dendrite growth, morphology control and deposition properties of uranium electrorefining14
Grain-boundary phosphorus segregation in highly neutron-irradiated reactor pressure vessel steels and its effect on irradiation embrittlement14
Fracture properties of an irradiated PWR UO2 fuel evaluated by micro-cantilever bending tests14
Effect of deformation level and orientation on SCC of 316L stainless steel in simulated light water environments14
Implementation and Validation of the Hydride Nucleation-Growth-Dissolution (HNGD) model in BISON14
Design and fabrication of UN composites: From first principles to pellet production14
Irradiation hardening of stainless steel model alloy after Fe-ion irradiation and post-irradiation annealing treatment13
Tungsten–tantalum alloys for fusion reactor applications13
Direct observation of dislocation loops shrinkage upon annealing neutron-irradiated Fe–9Cr alloy13
Development of a grain growth model for U3Si2 using experimental data, phase field simulation and molecular dynamics13
Optimization of the t/10 offset correlation method to obtain the yield strength with the Small Punch Test13
The optical texture of PGA, Gilsocarbon, NBG-18, and IG-110 nuclear graphite13
Solid sorbents for gaseous iodine capture and their conversion into stable waste forms13
Initial investigation of B4C–TiB2 composites as neutron absorption material for nuclear reactors13
Ab initio molecular dynamics (AIMD) simulations of NaCl, UCl3 and13
Using molecular dynamics to predict the solidus and liquidus of mixed oxides (13
Mechanical behavior of a chromium coating on a zirconium alloy substrate at room temperature13
Development and characterization of advanced neutron multiplier materials13
Effect of Salt Composition and Microstructure on Stress Corrosion Cracking of 316L Austenitic Stainless Steel for Dry Storage Canisters13
Microstructural characterization of copper coatings in development for application to used nuclear fuel containers13
Effects of ultrasonic surface rolling processing on the corrosion properties of uranium metal13
Recrystallization behaviour of high-flux hydrogen plasma exposed tungsten13
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