Journal of Nuclear Materials

Papers
(The H4-Index of Journal of Nuclear Materials is 32. 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 2021-04-01 to 2025-04-01.)
ArticleCitations
Response of oxide nano-particles to helium ion irradiation in oxide dispersion strengthened steels with CeO2 and Y2O3 additions164
Elevated tritium diffusion barrier from varied segregation strengths in Li4TiO4 (1–10) grain boundary122
Ab initio study of neutral point defect properties in 6H-SiC based on the SCAN functional68
Corrigendum to “Non-electric mass transfer between stainless steel 316H and glassy carbon in NaF-KF-UF4 salt” [Journal of Nuclear Materials 604 (2025) 155534]60
Radiation damage study of POCO ZXF-5Q graphite for neutrino production targets using 4.5 MeV helium ions56
Investigating the interactions between hydrotalcite and U(IV) nanoparticulates54
In-situ TEM characterization of dislocation loops in tungsten under simultaneous helium and hydrogen irradiation54
The role of microstructural evolution in irradiation hardening of Alloy 718 under low dose proton irradiation54
Crack self-healing by high-temperature annealing of a 90W7Ni3Fe tungsten heavy alloy54
Neutron flux impact on rate of expansion of quartz53
Dispersed barrier hardening modeling on depth-distributed helium bubbles in iron-based alloys52
Insight into threshold creep behaviour in Zr-2.5Nb alloy50
Fracture behavior and grain boundary cohesion of alumina scales formed on ion-irradiated FeCrAl-ODS alloy49
Radiation damage of Ceria and Ceria-gadolinia mixed oxides: effect of the Gd content and Ion stopping power45
New insights into the carburization and tensile behavior of superalloys Inconel 617 and Incoloy 800H in helium containing trace methane at 950 °C44
Ab initio study of helium in titanium beryllides42
Hydrogen permeability of non-stoichiometric tungsten oxides40
Systematic experimental and model-based evaluation of the synergistic effects of alloy composition and damage rate on the formation of Cr-rich precipitates in Fe–Cr–Al alloys under ion irradiation40
An improved analysis of small punch deformation for evaluating tensile properties39
Measurement of local mechanical properties for Cr-coated accident tolerant fuel cladding39
Effect of the cycle number on fretting wear behavior of alloy 690TT tube in high-temperature pressurized water39
Evaluation of the anisotropic grain boundaries and surfaces of α-U via molecular dynamics38
Positron annihilation study of the reactor pressure vessel model steels irradiated in the high flux reactor37
Correlation between impact angle and corrosion-erosion damage behavior of ferritic/martensitic steel exposed to flowing oxygen-saturated lead-bismuth eutectic36
Modelling the brittle-to-ductile transition of high-purity tungsten under neutron irradiation35
Diffusion behavior of oxygen in the electro-deoxidation of uranium oxide in LiCl-rich melt35
Characterization of modeling and experimental data inconsistencies from burst testing for high-burnup commercial fuel rod applications34
High temperature oxidation of cold spray Cr-coated accident tolerant zirconium-alloy cladding with Nb diffusion barrier layer34
Exploring the inhibitory effect of WTaVCr high-entropy alloys on hydrogen retention: From dissolution, diffusion to desorption34
Parameterization of vacancy production rate in phase-field models of fission gas bubble evolution in nuclear fuel34
Hydrogen diffusion in zirconium cladding alloys with an inner liner as quantified by neutron radiography and nanoindentation33
Efficient atomistic simulations of radiation damage in W and W–Mo using machine-learning potentials33
Parametrization of embedded-atom method potential for liquid lithium and lead-lithium eutectic alloy32
Self–ion irradiation of high purity iron: Unveiling plasticity mechanisms through nanoindentation experiments and large-scale atomistic simulations32
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