OOIR: Observatory of International Research

Papers

(The H4-Index of Journal of Nuclear Materials is 30. 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 2021-02-01 to 2025-02-01.)

Article	Citations
Proton irradiation effects in Molybdenum-Carbide-Graphite composites	122
Fracture behavior and grain boundary cohesion of alumina scales formed on ion-irradiated FeCrAl-ODS alloy	112
Ab initio study of helium in titanium beryllides	52
Insight into threshold creep behaviour in Zr-2.5Nb alloy	50
Radiation damage of Ceria and Ceria-gadolinia mixed oxides: effect of the Gd content and Ion stopping power	50
The role of microstructural evolution in irradiation hardening of Alloy 718 under low dose proton irradiation	50
Elevated tritium diffusion barrier from varied segregation strengths in Li4TiO4 (1–10) grain boundary	49
Response of oxide nano-particles to helium ion irradiation in oxide dispersion strengthened steels with CeO2 and Y2O3 additions	49
A new model of fission gas bubble growth and mechanism analysis for U-xZr fuels	48
Ab initio study of neutral point defect properties in 6H-SiC based on the SCAN functional	46
New insights into the carburization and tensile behavior of superalloys Inconel 617 and Incoloy 800H in helium containing trace methane at 950 °C	44
Neutron flux impact on rate of expansion of quartz	43
In-situ TEM characterization of dislocation loops in tungsten under simultaneous helium and hydrogen irradiation	42
Evolution of IFMIF-DONES' heart: System overview of the Test Cell	40
Quantitative in-situ investigation of dislocation loop motion in Fe9Cr1.5W ferritic/martensitic steel under the coupling effects of constant stress and irradiation	40
Calculation of grain boundary diffusion coefficients in γU-Mo using atomistic simulations	40
Mechanisms of plastic deformation and fracture of austenitic chromium-nickel steel irradiated during 45 years in WWER-440	39
Deuterium transport and retention properties of representative fusion blanket structural materials	39
Investigating the interactions between hydrotalcite and U(IV) nanoparticulates	39
Dose rate effects on damage accumulation and void growth in self-ion irradiated tungsten	38
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 irradiation	38
Microstructural evolution in tungsten binary alloys under proton and self-ion irradiations at 800 °C	35
No ball milling needed: Alternative ODS steel manufacturing with gas atomization reaction synthesis (GARS) and friction-based processing	34
Diffusion behavior of oxygen in the electro-deoxidation of uranium oxide in LiCl-rich melt	33
Solubility of monoclinic and yttrium stabilized cubic ZrO2: Solution and surface thermodynamics guiding ultra-trace analytics in aqueous phase	33

An improved model for the prediction of He bubbles driven W fuzz growth at different temperatures	31
Parameterization of vacancy production rate in phase-field models of fission gas bubble evolution in nuclear fuel	31
Effect of irradiation temperature on the fracture-mechanical behaviour of tungsten irradiated to 1 dpa	30
Coupling effects in borosilicate glass leaching: A study on La/V doping	30
Self-ion irradiation response of (CoCrFeNi)94Ti2Al4 alloy containing coherent nanoprecipitates	30
Characterization of modeling and experimental data inconsistencies from burst testing for high-burnup commercial fuel rod applications	30