The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems. Benefits to authors We also provide many author benefits, such as free PDFs, a liberal copyright policy, special discounts on Elsevier publications and much more. Please click here for more information on our author services.

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The COMPASS upgrade tokamak (Panek et al 2017 Fusion Eng. Des. 123 11–16) will be a tokamak of major radius R 0 = 0.894 m with the possibility to reach high field (B t ∼ 5 T) and high current (I p ∼ 2 MA). The machine should see its first plasma in 2023 and H-mode plasma will be obtained from 2025. The main auxiliary heating system used to access H-mode will be 4 MW of neutral beam injection (NBI) power. The NBI will have a nominal injection energy of 80 keV, a maximum injection radius R tan = 0.65 m and will create a population of well-confined energetic D ions. In this contribution, our modelling studies the NBI deposition and losses when a significant edge background density of neutrals is assumed. We follow the fast ions in the 3D field generated by the 16 toroidal field (TF) coils using the upgraded EBdyna orbit solver (Jaulmes et al 2014 Nucl. Fusion 54 104013). We have implemented a Coulomb collision operator similar to that of NUBEAM (Goldston et al 1981 J. Comput. Phys. 43 61) and a charge-exchange operator that follows neutrals and allows for multiple re-ionizations. Detailed integrated modelling with the METIS code (Artaud et al 2018 Nucl. Fusion 58 105001) yields the pressure and current profiles for various sets of achievable engineering parameters. The FIESTA code (Cunningham 2013 Fusion Eng. Des. 88 3238–3247) calculates the equilibrium and a Biot–Savart solver is used to calculate the intensity of the perturbation induced by the TF coils. Initial distributions of the NBI born fast ions are obtained from the newly developed NUR code, based on Suzuki et al (1998 Plasma Phys. Control. Fusion 40 2097). We evolve the NBI ions during the complete thermalization process and we calculate the amount of NBI ions loss in the edge region due to neutralizations. Results indicate the NBI losses for various injection geometries, various engineering parameters and various assumptions on the magnitude of the background neutral densities.

Following ELMy H-mode experiments with liquid metal divertor target on the COMPASS tokamak, we predict the behavior of a similar target on COMPASS Upgrade, where it will be exposed to surface heat fluxes even higher than those expected in the future EU DEMO attached divertor. We simulate the heat conduction, sputtering, evaporation, excitation and radiation of lithium and tin in the divertor area. Measured high-resolution data from COMPASS tokamak were rescaled towards the Upgrade based on many established scalings. Our simulation then yields the amount of released metal which ranges from 4 mg s −1 upto 12 g s −1 depending mainly on the geometry and Li/Sn choice, quite independently from active cooling or strike point sweeping. Extreme heat loads are predicted on future fusion reactor divertorsExpected engineering lifetime of plasma-facing components (PFC) loaded by 16 MW m −2 corresponds to an acceptable 3 months of cumulated plasma exposure (figure 18 in [1]), however, already for 25 MW m −2 it's only an hour (totally unacceptable). We predicted [2, 3] plasma heat flux q ⊥ perpendicular to the surface of divertor tiles (far off its edges) in ITER attached L-mode (without any fusion) as already q ⊥ =10 MW m −2 . Q=10 ITER H-mode yields P divertor ITER =(1-f rad )100 MW, ⅔ of which deposits on the outer target A dw =2πR 0 f x λ q,integral =0.4 m 2 area [4,42]. Properly controlled impurity seeding increases natural radiation fraction up to f rad =85% [5] on present tokamaks without undesired cooling of the fusing hot plasma core. Accounting further for the toroidal bevel, one gets q ⊥ =⅔P divertor ITER /A dw 4.2°/2.7°=31 MW m −2 [6], thus twice above the engineering limit.Assuming that rather the very complex and recent turbulence models XGC1 [7] and BOUT++ (figure 3 in [Xu19]) will be closer to reality than these empirical scalings, the energy will be deposited on a much more optimistic 5-10× larger area (λ q =6 mm thanks to stronger SOL turbulence due to larger a/r ion Larmor ), predicting thus 5<q ⊥ [MW m −2 ]<16 [1]. The European DEMO 2 GW fusion power plant study [8, 9] counts with similar P SOL =150 MW due to assumed strong core line radiation, namely to stay within the 16 MW m −2 limit, the DEMO edge+SOL+divertor radiation must never drop below 97% which is a big challenge. Estimate for the RECEIVED