Plasma facing materials (PFMs) in future magnetic fusion devices will face various challenges, such as 14.1 MeV neutron and transmutation gas irradiation at high temperatures. W has been considered as one of the most effective candidates for a PFM in recent years. However, pure W exhibits some drawbacks that limit its applications. Conversely, W-ZrC (W-0.5%ZrC, mass fraction) alloy demonstrates excellent performance, such as a relatively low ductile-brittle transition temperature (DBTT), high ductility, and high strength, which will be particularly useful in future fusion reactors. In this work, the Doppler-broadening slow positron beam analysis (DB-SPBA) and SEM were used to characterize the W-ZrC alloy, which had been irradiated by pure H neutral beam or H+6%He (atomic fraction) neutral beam. In the DB-SPBA, parameters S and W were used to characterize the open volume defects in the samples. Under the pure H neutral beam irradiation, the defects were mainly H-V complexes with a large ratio of vacancy to H in the sample at the surface temperature of 850^oC. When the surface temperature of the irradiated sample was 1000^oC, there was only one kind of vacancy-type defect without any defect damage layer due to the recovery of defect damage in the sample. The surface morphology was smooth and flat at the irradiated sample surface temperature of 1000^oC, and the most of pinhole damage structures disappeared compared to the surface temperature of 850^oC. The S value in the sample subjected to the H+6%He neutral beam irradiation at the surface temperature of 800^oC was larger than that at 700^oC because of the increasing vacancy-type defect volume, and defect types were more complex in the 800^oC sample. The defect damage layer in the 800^oC sample was wider than that in the 700^oC sample. Both the 700^oC and 800^oC samples presented more than one type of defects, but the sample surface damage was significantly more serious at 800^oC.