Abstract
This paper is dedicated to an experimental program focused on the evaluation of microstructure and failure mechanisms of WWER 440 type nuclear reactor pressure vessel cladding made from Sv 08Kh19N10G2B stainless steel. Static fracture toughness tests performed on standard precracked single edge bend specimens revealed extreme variations in fracture toughness values, J0.2. Fractured halves of test specimens were subject to detailed fractographic and metallographic analyses in order to identify the causes of this behavior and to determine the relationship between local microstructure, failure mode and fracture toughness. Results indicated that fracture toughness of the cladding was adversely affected by the brittle cracking of sigma particles which caused a considerable decrease in local ductile tearing resistance. Extreme variations in relative amounts of sigma phase, as well as the extreme overall structural heterogeneity of the cladding determined in individual specimens, provided a reasonable explanation for variations in fracture toughness values.
Highlights
Austenitic cladding is an integral component of the WWER 440 nuclear reactor pressure vessel (RPV) designed to ensure anticorrosive protection
Individual phases were identified on the basis of typical amounts of alloying elements determined by energy dispersive X-ray spectrometer (EDX) analysis: the cellular zone (CELZ) was identified in the vicinity of the cladded substrate, i.e., the first cladding made from Sv 07Kh25N13 steel
The results of experimental research focused on the evaluation of fracture toughness tests of WWER 440 reactor pressure vessel austenitic cladding made from Sv
Summary
Austenitic cladding is an integral component of the WWER 440 nuclear reactor pressure vessel (RPV) designed to ensure anticorrosive protection. Only limited attention has been paid to the impact of cladding on RPV integrity. Significant differences in thermal expansion coefficients with respect to RPV base metals [3] could lead to a stress peak during a severe emergency transient associated with abrupt overcooling of the RPV. This event is referred to as pressurized thermal shock (PTS) and presents the potential risk of interfacial crack initiation and propagation [4]. PTS safety analyses have recently become a subject of particular interest to nuclear power plant operators [5,6,7]
Sign-in/Register to view highlights & summary Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callDisclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
