The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m−2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s−1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m−2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m−2 heat load, albeit with a fatigue life of approximately 0.55 years based on the expected operation scenario of a prototype or test reactor. This study extends the state of knowledge of the technological limit of a divertor based on a RAFM steel pipe structure.

The long-pulse high-confinement (H-mode) plasma regime is considered to be a preferable scenario in future fusion devices, and in the period of normal operation during H-mode, edge-localised modes (ELMs) are one of the most serious threats to the performance and capability of divertor targets. The EAST recently achieved a variety of H-mode regimes with ELMs. For the purpose of studying the performance of the EAST upgraded divertor during type I ELMs, a series of simulations were performed by using three-dimensional (3D) finite element code. To make a visible outcome of the direct ELM impact on the divertor targets, a preliminary evaluation system with three indices to exhibit the influence has been developed. The indices that comprise temperature evolution, thermal penetration depth and crack initiation life, which could reveal the process of micro-crack formation, are calculated in both low and high-power scenarios for type I ELMs. The initial results indicate that the transient heat load has a significant influence in a very short thickness layer along the direction perpendicular to the plasma-facing surface throughout its duration. The conclusion could offer a pertinent guide to the next-step high-power long-pulse operation in EAST and would also be helpful for scientifically studying the damage and fatigue mechanism of the divertor in ITER and future fusion power reactors.

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