Journal of Nuclear Materials

Title Publication Date Language Citations
The surface and grain boundary properties of uranium boride: A DFT calculation2023/11/01English
Dissolution of (U,Th)O2 heterogeneous mixed oxides2023/12/01English
Mixed analytical model to estimate the anisotropic effective thermal conductivity and in-reactor performance of metallic microcell UO22023/11/01English
Molybdenum and Tungsten isotope compositions of UO2 fuel pellets: Implications for isotopically enriched taggants2023/10/01English
Lithium-deuterium co-deposition2023/10/01English
Effect of radiation damage on liquid Pb corrosion at the Fe/Pb solid-liquid interface2023/09/01English
On the sink strength of coherent nanoparticles in irradiated crystals2023/10/01English
Microstructure and corrosion resistance of Z3CN20.09M stainless steels after different thermo-mechanical processing2023/04/01English
Wrinkled strong quasi-atomic planes caused by partial covalent bonds in α-uranium2023/04/01English
High-temperature steam oxidation of (Ti, Mo)C-forming FeCrAlY alloy2023/02/01English
Cavity evolution and void swelling in dual ion irradiated tempered martensitic steels2023/04/01English
Preparation of ZrC coating in TRISO fuel particles by precise transportation of solid precursor and its microstructure evolution2023/02/01English
Testing fast reactor fuels in a thermal reactor: Comparison of transmutation metallic fuel alloys behavior by scanning electron microscopy2023/03/01English
The Thermo-Elastic Properties and Damping of U-6wt%Nb2023/04/01English
Microstructural evolution in tungsten binary alloys under proton and self-ion irradiations at 800 °C2023/03/01English
Synergistic influence of dislocations and helium cluster on hydrogen atom in tungstenEnglish
Editorial BoardEnglish
Fabrication of americium containing transmutation targetsEnglish
Positron annihilation spectroscopy evaluation of using proton irradiation as a surrogate for early neutron radiation damageEnglish
Effect of hydride orientation on tensile properties and crack formation in zirconium alloy cladding tubesEnglish
Corrosion tests on austenitic samples with alumina and alumina-forming coatings in oxygen-containing stagnant Pb and turbulently flowing PbBiEnglish
Detecting irradiation defects in materials: A machine learning approach to analyze helium bubble imagesEnglish
Deformation localisation in ion-irradiated Fe and Fe10CrEnglish
Synthesis, Characterization and Dissolution behavior of Nitrides of U-Zr and U-Pu-Zr metallic alloy fuels for Aqueous reprocessingEnglish
Fracture mechanics approach to TRISO fuel particle failure analysisEnglish
Phase field-volumetric lattice Boltzmann model of ion uptake in porous nuclear waste form materials under continuous flowEnglish
Temperature-time dependence and mechanisms of redox reaction in Cr-coated Zr alloy cladding during steam oxidation at 900–1250 ℃English
High-temperature degradation behavior of PEO-coated ZIRLO alloy in N2 and N2+steam environments at 900 and 1000 °CEnglish
Analysis of radially resolved thermal conductivity in high burnup mixed oxide fuel and comparison to thermal conductivity correlations implemented in fuel performance codesEnglish
Chemical and electrochemical reduction of solid oxide fuelEnglish