OOIR: Observatory of International Research

Papers

(The H4-Index of Journal of Nuclear Materials is 31. 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 2020-11-01 to 2024-11-01.)

Article	Citations
The role of TeO2 insertion on the radiation shielding, structural and physical properties of borosilicate glasses	109
High temperature Cr-Zr interaction of two types of Cr-coated Zr alloys in inert gas environment	106
Additive manufacturing of silicon carbide for nuclear applications	66
Mechanical properties of tungsten: Recent research on modified tungsten materials in Japan	64
A review of the LIBS analysis for the plasma-facing components diagnostics	63
Mechanical and chemical properties of PVD and cold spray Cr-coatings on Zircaloy-4	59
Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan	52
Efficient capture of radioactive iodine by a new bismuth-decorated electrospinning carbon nanofiber	50
Design considerations for high entropy alloys in advanced nuclear applications	50
Review of manufacturing technologies for coated accident tolerant fuel cladding	48
Thermal aging behaviors of duplex stainless steels used in nuclear power plant: A review	48
High-temperature oxidation and quenching of chromium-coated zirconium alloy ATF cladding tubes with and w/o pre-damage	47
Molecular dynamics simulations of high-dose damage production and defect evolution in tungsten	44
A study of the oxidation behaviour of FeCrAl-ODS in air and steam environments up to 1400 °C	44
TRISO particle fuel performance and failure analysis with BISON	44
Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation	44
High-heat-flux technologies for the European demo divertor targets: State-of-the-art and a review of the latest testing campaign	43
Perspectives on multiscale modelling and experiments to accelerate materials development for fusion	43
Crystalline phosphates for HLW immobilization - composition, structure, properties and production of ceramics. Spark Plasma Sintering as a promising sintering technology	42
Processing of tungsten through electron beam melting	40
Cluster dynamics simulation of xenon diffusion during irradiation in UO2	39
Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: Uranium mononitride	39
STEM Characterization of Dislocation Loops in Irradiated FCC Alloys	39
Fission gas diffusion and release for Cr2O37
Fracture-mechanical properties of neutron irradiated ITER specification tungsten	36

Ferrite formation and its effect on deformation mechanism of wire arc additive manufactured 308 L stainless steel	36
High temperature zirconium alloys for fusion energy	36
A refined oxidation mechanism proposed for ferritic-martensitic steels exposed to oxygen-saturated liquid lead-bismuth eutectic at 400°C for 500 h	35
Development of CuCrZr via Electron Beam Powder Bed Fusion (EB-PBF)	34
Uniform corrosion of FeCrAl cladding tubing for accident tolerant fuels in light water reactors	33
Helium bubble formation in refractory single-phase concentrated solid solution alloys under MeV He ion irradiation	32
Design and strategy for next-generation silicon carbide composites for nuclear energy	31