Self-toughening mechanism of the ZrO2 scale and the precipitation of Sn during the steam oxidation of ZIRLO™ at 1200° C
The steam oxidation behavior of ZIRLO™ at 1200° C was investigated using a
thermogravimetric analyzer. The microstructure of the oxide was analyzed using SEM …
thermogravimetric analyzer. The microstructure of the oxide was analyzed using SEM …
Effects of steam concentration and flow rate on the high temperature oxidation of PEO-coated zirconium alloy at 1000° C and 1200° C
H Guan, X Wang, C Xu, Y Liao, C Gao, J Du… - Surface and Coatings …, 2022 - Elsevier
A compact oxide coating of 7 μm thick on ZIRLO zirconium alloy was fabricated by plasma
electrolytic oxidation (PEO) in phosphate electrolyte. The effects of steam concentration and …
electrolytic oxidation (PEO) in phosphate electrolyte. The effects of steam concentration and …
A novel strategy for high entropy alloy coating of nuclear fuel cladding: Design, preparation, microstructure, and high-temperature steam oxidation
C Wang, W Yang, W Shao, Z Wang, D Xu… - International Journal of …, 2025 - Elsevier
Designed and constructed the novel bilayer structural Mo/FeCrAlMo coating to enhance the
resistance of nuclear fuel cladding to sudden loss of coolant accidents in this work …
resistance of nuclear fuel cladding to sudden loss of coolant accidents in this work …
Chemical interactions between pre-oxidized Zircaloy-4 and 304 stainless steel-B4C melt at 1300° C
During severe nuclear accidents, control rods rapidly liquefy at temperatures above 1250° C
due to eutectic reaction, forming a 304 stainless steel (304SS)-B 4 C melt. The melt will …
due to eutectic reaction, forming a 304 stainless steel (304SS)-B 4 C melt. The melt will …
Precipitation behavior in Fe-13Cr-6Al-xMo-0.5 Nb (x= 2, 4) alloys for cladding materials during long-term aging at 475° C
H Zhang, Z Wang, G Jia, J Peng, J Li, X Xiao - Materials Science and …, 2020 - Elsevier
Most of materials used for the nuclear fuel cladding and structural components in light-water
reactors are Zr-based alloys [[1],[2],[3],[4]]. However, the severe oxidation of Zr-based alloys …
reactors are Zr-based alloys [[1],[2],[3],[4]]. However, the severe oxidation of Zr-based alloys …
Chemical interactions between ZrO2/α-Zr structured Zircaloy-4 and 316 stainless steel-B4C melt at 1300° C
L Zheng, F Wang, H Li, Z Jiang, S Ueda - Journal of Nuclear Materials, 2022 - Elsevier
To deeply understand chemical interactions among nuclear materials during core meltdown,
pre-oxidized Zircaloy-4 was re-heated under oxygen starvation to form thermodynamically …
pre-oxidized Zircaloy-4 was re-heated under oxygen starvation to form thermodynamically …
Out-of-Pile Performances of Zr-Sn-Nb-Fe Alloys for PWR Fuel Cladding
H Zhuo, ZB Yang, ZQ Cheng, YF Yang… - IOP Conference Series …, 2020 - iopscience.iop.org
The safety and reliability of Pressure Water Reactors (PWRs) is closely related to the
performances of zirconium (Zr) alloy as fuel rod cladding material. Zr-Sn-Nb-Fe series alloys …
performances of zirconium (Zr) alloy as fuel rod cladding material. Zr-Sn-Nb-Fe series alloys …