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Combining the advantages of a wet chemical method and spark plasma sintering, carbide-doped materials W-1wt%TiC and W-1wt%ZrC were prepared. Microstructural evolution in W-1wt%TiC and W-1wt%ZrC under irradiation of 5 keV He⁺ at 600 °C to fluences up to 5.0 × 10²¹ ions/m² with ion flux of about 8.8 × 10¹⁷ ions/m²s was investigated by transmission electron microscopy (TEM). The dislocation loop number density of W-1wt%TiC was higher than that of W-1wt%ZrC, but the average loop size of the W-1wt%TiC was in average smaller. There were no observable helium bubbles in W-1wt%TiC and W-1wt%ZrC, exhibiting higher radiation resistance to He⁺ compared to pure W. He⁺ pre-damaged and undamaged W-1wt%TiC and W-1wt%ZrC samples were irradiated by 5 keV D₂⁺ to estimate the D retention in doped W materials. The irradiation damage impact of He⁺ on deuterium retention was examined by a method of thermal desorption spectroscopy (TDS). Compared with the undamaged samples, it was illustrated that D₂ retention of W-1wt%TiC and W-1wt%ZrC increased after He⁺ pre-irradiation.