Plasma-facing components in next-step fusion devices: Implications from fuel retention and thermomechanical material property studies

George Tynan

University of California San Diego

Friday, May 18, 2018

3:00pm

NW17-218

PSFC Seminars

In this talk we examine critical aspects of tungsten as a first wall material in nuclear-grade fusion devices. Tritium fuel self-sufficiency is obviously needed for a D-T based device that intends to operate at finite duty cycles. However due to low burn-up fraction and recycling effects, the acceptable tritium retention probability in the first wall is quite low, typically below 10^-6, Here we present D retention results obtained in heavy-ion beam damaged W that is subsequently exposed to D plasmas. X-ray scattering and TEM imaging shows the size and density of the displacement-damage induced defects.  NRA and TDS analysis shows the spatial profile and total inventory of trapped D.  The results show increased trapping in damage sites (as expected); we also show that annealing at material temperatures of ~1000K and higher can quench the defect production, allowing a reduction in retention probability.  The results point to a conclusion that neutron-damaged tungsten first walls should have adequately low T retention to permit tritium breeding ratios exceeding unity.   Using recently developed nano-scale thermal conductivity measurements, we also show a significant reduction in thermal conductivity induced by radiation damage; the thermal conductivity then recovers to nearly undamaged values when the damage occurs at elevated temperatures.  The results provide needed data on the the retention and thermal conductivity properties of tungsten in a D-Tfusion environment.