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 2020-04-01 to 2024-04-01.)

Article	Citations
High temperature Cr-Zr interaction of two types of Cr-coated Zr alloys in inert gas environment	76
The role of TeO2 insertion on the radiation shielding, structural and physical properties of borosilicate glasses	74
Mechanical properties of tungsten: Recent research on modified tungsten materials in Japan	55
Additive manufacturing of silicon carbide for nuclear applications	53
A review of the LIBS analysis for the plasma-facing components diagnostics	52
Cr-coated Zr-4 alloy prepared by electroplating and its in situ He+ irradiation behavior	52
Lead-bismuth eutectic (LBE) corrosion behavior of AlTiN coatings at 550 and 600゜C	52
Mechanical and chemical properties of PVD and cold spray Cr-coatings on Zircaloy-4	47
Radiation damage tolerance of a novel metastable refractory high entropy alloy V2.5Cr1.2WMoCo0.04	47
Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding	47
Uranium nitride tristructural-isotropic fuel particle	45
SCIANTIX: A new open source multi-scale code for fission gas behaviour modelling designed for nuclear fuel performance codes	42
A review of in-pile fuel safety tests of TRISO fuel forms and future testing opportunities in non-HTGR applications	41
Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan	41
Efficient capture of radioactive iodine by a new bismuth-decorated electrospinning carbon nanofiber	41
Effects of contents of Al, Zr and Ti on oxide particles in Fe–15Cr–2W–0.35Y2O3 ODS steels	40
Highly effective prussian blue-coated MXene aerogel spheres for selective removal of cesium ions	40
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 ductility	39
TRISO particle fuel performance and failure analysis with BISON	39
Perspectives on multiscale modelling and experiments to accelerate materials development for fusion	37
Review of manufacturing technologies for coated accident tolerant fuel cladding	37
Design considerations for high entropy alloys in advanced nuclear applications	37
Cold spray deposition of 304L stainless steel to mitigate chloride-induced stress corrosion cracking in canisters for used nuclear fuel storage	36
Thermal aging behaviors of duplex stainless steels used in nuclear power plant: A review	35
High-heat-flux technologies for the European demo divertor targets: State-of-the-art and a review of the latest testing campaign	35

Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: Uranium mononitride	34
A study of the oxidation behaviour of FeCrAl-ODS in air and steam environments up to 1400 °C	34
Fracture-mechanical properties of neutron irradiated ITER specification tungsten	32
Cluster dynamics simulation of xenon diffusion during irradiation in UO2	32
Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation	31
High-temperature oxidation and quenching of chromium-coated zirconium alloy ATF cladding tubes with and w/o pre-damage	31
Fission gas diffusion and release for Cr2O31
The structure of molten FLiNaK	31
Impact of neutron irradiation on the strength and ductility of pure and ZrC reinforced tungsten grades	31