Role of doped ZrC on deuterium trapping in W-ZrC alloy

Understanding and predicting hydrogen isotopes (His) retention in plasma-facing materials (PFMs) is crucial for the safe and efficient operation of fusion reactors. Here, a new candidate PFM, W-ZrC alloy, was exposed to D plasma at temperatures in the range from 400 K to 850 K. Surface morphology analysis revealed that the maximum blistering effect on W-ZrC alloy occurred at 600 K, which is 100 K higher than for pure tungsten (W). A quantitative statistical analysis revealed that higher temperature led to an increase in blister diameter and a concomitant decrease in areal density for both W and W-ZrC alloy. The blisters on the W and W-ZrC alloy originated from subsurface cavities, with nucleation sites localized in both intragranular and intergranular regions. The intergranular cavity beneath the blisters of W-ZrC alloy tended to extend along the phase boundaries between ZrC particles and W grains, and the ZrC particles at grain boundaries effectively suppressed the intragranular cavity propagation. If the exposure temperature exceeded 500 K, implanted D atoms dissolved and accumulated within ZrC grains in the W-ZrC alloy, leading to a new high-temperature thermal desorption spectroscopy (TDS) peak ∼ 980 K. Additionally, the total D retention in W-ZrC alloy is higher than that in pure W, especially at the exposure temperatures of 600 K and 700 K. This work provides key insights into surface blistering and D retention behavior in W-ZrC alloys, establishing a foundational basis for optimizing their performance as PFMs in fusion reactor applications.