13–15 Nov 2023
King Fahd Conference Center, KFUPM, Dhahran, KSA
Asia/Riyadh timezone

Hydrogen and Temper Embrittlement Effects on Fatigue Fracture Behaviour of 2.25Cr-1Mo Nuclear Reactor Pressure Vessel Steel

14 Nov 2023, 15:50
20m
60/Ground-105 - Lecture Hall (Administration Building)

60/Ground-105 - Lecture Hall

Administration Building

80
Show room on map

Speaker

Prof. M. A. Islam (Bangladesh University of Engineering and Technology (BUET), Dhaka-1000, Bangladesh)

Description

Nuclear reactor pressure vessel is a critical component of any nuclear power plant. The vessel is a thick wall container that withstands internal pressure caused by the reactor activity. This also plays important role to provide necessary barrier to keep radioactive materials out of the environment. Considering the functions, high strength low steel is usually proper candidate for making the vessel. However, reactor operation generates subatomic particle neutron under operation causing embrittlement and degradation of the steel vessel. Another source of degradation of the pressure vessel steel is by temper embrittlement due to its exposure at high temperature. Hydrogen embrittlement further deteriorates the situation. In this research work, the effect of hydrogen embrittlement on the fatigue crack growth behaviour of 2.25Cr-1.0Mo pressure vessel steel before and after temper embrittlement has been studied. Experimental results revealed that both hydrogen embrittlement and temper embrittlement contribute in enhancing crack growth and also in changing the fracture morphology.

Primary author

Prof. M. A. Islam (Bangladesh University of Engineering and Technology (BUET), Dhaka-1000, Bangladesh)

Presentation materials

Peer reviewing

Paper