[en] Tungsten and its alloys have been considered as candidate plasma-facing materials for the EAST tokamak, ITER and future DEMO reactors, due to their favorable properties. A component of W coatings on Cu heat sink prepared by vacuum plasma spraying (VPS), acting as a poloidal movable limiter, has been being tested in the HT-7. During last HT-7 campaign, the W/Cu limiter was subjected to more than 20 similar Ohmic plasmas at different positions, still further exposed to more than 20 long pulse plasmas with durations from 10 to 60 s, driven by low hybrid wave (LHW) of 130 kW. The eroded C and CH particles from the fixed graphite limiters drift and redeposit on the PFM surfaces in the HT-7. It is necessary to understand and predict the PFM surface properties and performance under irradiation of H plasma with C impurities. Atomic scale simulations, that include detailed structure information of the material, are the most attractive tool for understanding such processes. In this work, the calculation is carried out using the classical molecular dynamics (MD) simulation method with a bond-order potential for the ternary W-C-H system. The statistical uncertainties of the MD calculations were kept to be less than one percent. The calculation results are compared qualitatively to the experimental data obtained from visible spectroscopy in the HT-7. By changing the C impurity concentration, target temperature, exposure time and position of the W/Cu limiter inserted into the HT-7 vacuum chamber, the process of chemical erosion of W by carbon coming as a hydrogen plasma impurity to the W surface has been investigated. Simultaneously, the following processes have been studied: ion reflection, CH physisorption and physical sputtering, local atomic composition changes and C concentration changes in the depth profile. Finally, it will be discussed that among above processes which step is crucial for the surface modification and determines the lifetime of W limiter. (authors)