Abstract: Compared to deuterium–tritium (D–T) fusion, proton–boron (p–¹¹B) fusion requires higher plasma temperatures and densities, posing greater thermal load challenges for components such as the divertor and first wall. This study focuses on the research of the candidate material carbon fiber composite (CFC) for EHL-2 divertor based on its advantages such as high thermal conductivity, low atomic number, and excellent plasma compatibility, while avoiding issues like neutron irradiation and tritium retention found in D–T fusion. Using a high-heat-flux linear plasma device and a high-energy electron gun, the study simulated the steady-state and transient thermal load conditions of EHL-2 divertor, respectively. The erosion and damage patterns of three types of CFC materials under different heat flux were investigated. The results indicated that under steady-state thermal loads exceeding 10 MW·m⁻², all three CFC materials demonstrate significant erosion, with the highest erosion rate reaching 0.14 mm·min⁻¹. However, the erosion rate can be significantly reduced through water-cooling structures. Microstructural analysis revealed no notable changes in the irradiated zone. Transient thermal load tests showed that CFC materials exhibit minimal damage under plasma disruption conditions, but severe damage occurs under vertical displacement event (VDE) conditions. We developed a multi-mechanism coupled plasma-material simulation model to analyze the erosion behavior of CFC under helium plasma irradiation. With this model we calculated the erosion rate, assessed material lifespan, and laid the foundation for determining the content and distribution of impurities in the plasma-wall interaction (PWI) process. Comparisons with the experimental data showed good consistency, confirming the reliability of the model.