Neutron damage in reduced activation martensitic alloys in inertial and magnetic fusion reactors: Comparison with fission reactors

Abstract

Neutron irradiation damage in terms of displacement and transmutant production, is assessed in inertial (ICF) and magnetic (MCF) confinement fusion reactors, for the reduced activation martensitic steel, B-TAHF. These results, along with current experimental and theoretical information available on ferritic/martensitic steels, have been used to infer irradiation responses, such as swelling and ductile to brittle transition temperature (DBTT). The ICF approach can lead to concepts with a first structural wall (FSW) able to last the power plant's entire lifetime. In MCF, the material at end of life does not fulfill the technological requirements. Damage calculations have also been carried out in the HFIR and the FFTF reactors, for purposes of planning and assessing irradiation experiments in fission reactors in the light of fusion reactor needs. Fission irradiations can be very useful for assessing ICF reactor materials, but appear to have little engineering relevance for MCF commercial applications.