The U.S. contribution to ITER (International Tokamak Experimental Reactor) program includes about 20% of the first wall and shield, consisting of 93 modules each weighing about 3.5T and 375 FW panels. There is a potential for significant cost savings by utilizing casting technology rather than welding/HIPing wrought plate material and employing extensive machining to fabricate the shield module. It is the responsibility of the US-program to demonstrate that the utilization of cast material will not impair the mechanical performance and corrosion behavior of the shield module.
The objective of this project is to investigate the stress corrosion cracking susceptibility of cast stainless steel in both unirradiated and neutron-irradiated condition in order to determine whether cast stainless steel can function in its intended role in ITER. The program includes the development of the facility for testing neutron-irradiated stainless steels in controlled water chemistry at temperatures below 300°C. Experiments performed in a controlled water environment will be conducted to determine the baseline stress corrosion cracking behavior of the unirradiated cast alloy and the behavior of neutron-irradiated cast alloy.