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In tokamak devices, the performance of plasma facing components (PFCs) under high heat loads is a key challenge for achieving long-pulse and high-performance plasma operations. The lower divertor of EAST adopts an ITER-like cassette modules (CMs) structure. However, regular distribution of damage along the toroidal direction have been observed on the divertor target plates, which not only affects the lifespan of PFCs but also limits the long-pulse and high-performance plasma operations. Therefore, investigating the toroidal distribution of the heat load on the divertor target in EAST is crucial for understanding the mechanism of damage on the target plate surface. The toroidal characteristics of heat loads on CMs of the lower divertor in EAST was analyzed using a three-dimensional magnetic field line tracing program (PFCFlux). For the inner vertical target (IVT) and outer vertical target (OVT) within a single CM, the heat load varies monotonically along the toroidal direction. In contrast, the heat load distribution on the outer horizontal target (OHT) is uniform, except at the chamfers, which is directly correlated with the target’s geometric structure and the magnetic field configuration. The surface heat load distribution exhibits toroidal non-uniformity, with the heat deposition pattern being generally consistent and displaying periodic variations. The toroidally uneven heat load distribution is attributed to the change in the magnetic field’s tilt angle. Additionally, the peak heat load is observed at the chamfer positions between cassettes, and an analysis is conducted on the relationship between peak heat load and the size of the chamfer. Smaller tilt angles result in lower surface heat load. Furthermore, a comparative analysis of peak heat loads at the chamfer under different plasma current conditions reveals that higher plasma currents generally result in increased peak heat loads. This investigation into the toroidal heat load distribution on the divertor surface contributes to understanding of the damage mechanism of W PFC as the target plate and provides valuable data for divertor design of future fusion devices.

The consequences of tungsten (W) cracking on divertor lifetime and plasma operation are high priority issues for ITER. One actively cooled ITER-like plasma facing unit (PFU) has been pre-damaged in a High Heat Flux (HHF) facility before its installation in WEST in order to assess the damage evolution after tokamak plasma exposure. The resulting pre-damage exhibits micrometer-size crack network and high roughness on the tungsten monoblock (MB) top surface. A total of 10 MBs, equally distributed on the low and high field sides of the lower divertor, have been pre-damaged among the 35 radially aligned MBs characteristic of the WEST PFU. Subsequent plasma exposure was carried out, from the first breakdown achieved in WEST (in 2017) until the removal of the damaged PFU three years later (2020). On top of the whole WEST plasma exposure (covering C1-C4 experimental campaigns), a dedicated experiment has also been performed in the frame of the EU work program to maximize the power and energy loads on one of the damaged MBs featuring a “crack network” pattern. The MB top surface, including both “crack network” damage and “healthy” (undamaged) areas, was monitored with a high spatial resolution IR camera to detect any potential evolution of the damage pulse after pulse. This paper describes the full plasma exposure achieved in the WEST tokamak (including large number of steady-state and transient heat loading cycles), the dedicated “damaged PFU exposure” experiment together with the experimental results (heat loading on the damaged MBs). Post-mortem measurement reveals significant broadening of the cracks and new cracks in the electron beam loaded area only.