Structural materials for ITER in-vessel component design

The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m(2) in the basic performance phase (BPP)) within a temperature range from 20 to 300 degrees C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 degrees C, for doses of 5-10 dpa, In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. Estimates of radiation damage at the locations for re-welding show that the dose will not exceed 0.05 dpa (with He generation of 1 appm) for the manifold and 0.01 dpa (with He generation 0.1 appm) for the back plate for the BPP of ITER operation. Existing experimental data show that these levels will not result in property changes for SS; however, neutron irradiation and He generation promote crack formation in the heat affected zone during welding. Cu based alloys, DS-Cu (Glidcop A125) and PH-Cu (Cu-Cr-Zr bronze) are proposed as a structural materials for high heat flux components of limiter, baffle, divertor and primary wall. Irradiation significantly changes the mechanical properties, and the electrical and thermal conductivity of these alloys. The ductility of high strength Cu alloys is reduced at relatively low doses (< 0.2 dpa) for irradiation temperature similar to < 150 degrees C. For higher doses of irradiation it remains at the low (saturated) level. This effect is exhibited by both DS-Cu and PH-Cu alloys. For higher temperatures of irradiation, an increase of ductility and decrease of strength are observed resulting from radiation-induced microstructural instabilities. The 'softening' temperature for Cu-Cr-Zr alloys is in the range 230-250 degrees C; the corresponding temperature for DS Cu alloys is ca. 300-400 degrees C.