| ||Format||Pages||Price|| |
|PDF (188K)||11||$25||  ADD TO CART|
|Complete Source PDF (27M)||1254||$371||  ADD TO CART|
Cite this document
Radiation embrittlement and its mitigation by annealing of VVER-440 reactor pressure vessel steels are studied. The Russian regulatory approach for prediction of radiation embrittlement of VVER-440 steels is considered. Results of an investigation of materials cut out of operating nuclear power plants are discussed. Different models of re-irradiation embrittlement are compared. A new model for assessment of embrittlement under re-irradiation is developed.
reactor pressure vessel, radiation embrittlement, post-irradiation annealing, re-irradiation, ductile-to-brittle transition temperature
Leading research scientists, Russian Research Center “Kurchatov Institute”, ORM, Moscow,
Senior research scientist, Nuclear Safety Institute, Russian Academy of Sciences, Moscow,