Advanced Nuclear Reactor Materials and Corrosion Science
This community investigates the structural integrity, thermal performance, and chemical stability of materials used in advanced nuclear reactor systems, specifically focusing on fuel cladding, coolant interactions, and long-term durability under extreme conditions.
The research centers on the behavior of zirconium alloys, stainless steels, and high-entropy alloys when exposed to aggressive environments such as molten salts, liquid lead-bismuth eutectics, and high-temperature water. Key topics include corrosion mechanisms, heat transfer optimization, and the development of accident-tolerant fuel cladding. The work spans multiple reactor types, including molten salt reactors, sodium-cooled fast reactors, and pressurized water reactors, with a strong emphasis on understanding how temperature, atmosphere, and irradiation affect material failure modes and phase transformations.
The largest share of this community's output is found in zirconium research, accounting for 7.5% of all zirconium research, and 3,165 papers here. Plutonium research also features prominently, representing 8.3% of all plutonium research with 1,296 papers.
This group comprises 19,424 papers, published primarily in the Journal of Nuclear Materials, Annals of Nuclear Energy, and Nuclear Engineering and Design.
Recent work continues to focus on corrosion resistance in lead-bismuth eutectics and molten salts, with specific studies on FeCrAl alloys, zirconium coatings, and high-entropy alloys. Newer research also applies machine learning techniques to uncertainty quantification in reactor thermal modeling and investigates defect engineering in ceramics for radiation resistance.