Nobuyuki Asakura

In JA DEMO, a poloidal horseshoe limiter has been developed to protect the breeding blanket from excessive heat loads during plasma operation. The limiter protrudes from the first wall and intercepts charged particle heat load along magnetic field lines. Because this heat load is highly localized and sensitive to magnetic equilibrium, the heat load on limiter was evaluated at representative time points using magnetic field line tracing with the APPLE code. During ramp-up, plasma first touch the inboard limiter, and the peak heat flux reaches 2.8 MW/m2 at 27 seconds after plasma initiation. Owing to the transition from limiter to divertor configuration of the plasma, the heat flux shows spike-like temporal behaviour in this phase. During flat-top, the heat flux evolves more gradually and reaches a peak heat flux of 1.9 MW/m2 at 800 seconds. Based on this heat load scenario, time-dependent coupled thermal-structural analyses were performed for tungsten-monoblock plasma-facing units with copper or titanium interlayers and an F82H reduced activation ferritic/martensitic steel cooling pipe. The results showed that the interlayer material strongly affects temperature, stress, and plastic deformation. Although copper provides high thermal conductivity, its large thermal expansion causes higher von Mises stress and greater plastic deformation, particularly at the inner surface of the F82H pipe. In contrast, titanium reduces the stress by 52%, maintains elastic deformation in the pipe, and improves the structural integrity of the pressure boundary, indicating its promise for JA DEMO limiter applications.

SONIC divertor code enables simultaneous calculations of seeding impurity (Ar) and fusion product (He) transport. D2 and Ar gases were injected at outer midplane and upper location of the outer divertor, respectively. He exhaust in the detached divertor has been investigated for a reference series of JA DEMO with increasing deuterium gas puff rate (50 – 200 Pa m ³ s ^-1 ), where P _sep ~250 MW and the radiation fraction of 0.8. Outer divertor heat load (q _tot ^div ) was peaked at the attached region on the partially detached target, and the peak-q _tot ^div was reduced lower than 10 MWm ^-2 by increasing D2 puff rate, while maintaining relatively low nesep (2x10 ¹⁹ – 2.5x10 ¹⁹ m ^-3 ). Neutral and gas pressures in the divertor (P _D0 ^div , P _D2 ^div ) were evaluated at exhaust slots of the dome and in the sub-divertor. The total pressure (P _Dtot ^div ) in the D puff scan was increased from 1.4 to 3 Pa at the outer exhaust slots, and the outer peak-q _tot ^div was proportional to (P _Dtot ^div ) ^-0.7 . Plasma flow reversal with M _// = 0.2 – 0.3 was generated near the detach-attach boundary above the outer divertor target. Influences of the flow reversal on the plasma and impurity transport were determined. Convective heat flux was reversed, which significantly affected the ion heat flux profile in the upstream SOL. The flow reversal reduced impurity shielding and radiation loss in the outer divertor. Inner-enhanced in-out asymmetry of Ar concentration (c _Ar ) was large near the separatrix in SOL, which suggested that Ar ions are transported from the inner SOL to the confined plasma. In the confined plasma (edge), both c _Ar ^edge and c _He ^edge were reduced and poloidally uniform, where c _Ar ^edge was reduced to the marginal level and c _He ^edge was below the design reference by system code assumption (c _Ar ^main : 0.6%, c _He ^main : 7%). Consequently, these SONIC results were generally consistent with the JA DEMO higher-κ design, and larger D-puff (and smaller Γ _Ar,puff ) operation were preferable for the large power exhaust in the divertor and the impurity control in the edge plasma.

In divertor plasmas, ion–atom collisions critically influence plasma transport and heat flux to divertor targets. Fluid neutral models are widely used in divertor simulations due to their low computational cost and favorable convergence properties. However, source terms representing ion–atom collisions often differ across codes because of variations in the cross section data and in the physical approaches used to model collisions between hydrogen isotope ions and atoms of identical nuclei. This study systematically evaluates these differences and establishes more consistent modeling approaches based on the Boltzmann collision integral (BCI). Using the BCI, we investigate particle transport in velocity space and assess the effect of scattering angles and differential cross sections (DCSs). Several analytical expressions—rarely used in divertor studies—are reviewed. For energy transfer, we derive a new BCI-based expression and validate the analytical expressions by comparing with the direct BCI evaluation. It is also pointed out that the commonly used charge–exchange (CX) cross section data no longer reflect recent experimental and theoretical developments. For deuteron-deuterium atom collisions at T = 10 eV, the CX-based model overestimates the friction force by more than 30% compared with the BCI-based model. For energy transfer between deuterons at T = 20 eV and deuterium atoms at T = 10 eV, the CX-based model overestimates the transferred energy by approximately 25%. The use of DCSs and the analytical expressions introduced in this study is therefore recommended. Implementing them in place of the conventional CX-based model may modify the predicted plasma profiles and the assessment of the detached divertor.

Since the publication of the review Progress in the ITER Physics Basis (PIPB) in 2007, significant progress has been made in understanding the processes at the plasma-material interface. This review, part of the ITPA Nuclear Fusion Special Issue On the Path to Burning Plasma Operation, presents these developments, focusing on key areas such as the physics of plasma exhaust, plasma-material interactions, and the properties of plasma-facing materials and their evolution under plasma exposure. The coordinated efforts of the ITPA Topical Group on Scrape-Off Layer and Divertor Physics (DivSOL) have been instrumental in identifying and addressing critical research and development issues in numerous collaborative experimental and modelling projects.

An integrated divertor simulation code, SONIC, has been developed in order to predict a self-consistent transport solution of the plasma, neutral and impurities in the scrape-off layer (SOL) and divertor. SONIC code has contributed to determining the divertor design and power handling scenarios for the Japanese (JA) fusion demonstration (DEMO) reactor. Radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated to evaluate the power exhaust scenarios with P_sep = 230–290 MW. The simulation identified the decay length of the total parallel heat flux profile as being broader than the electron one, because of the ion convective transport from the outer divertor to the upstream SOL, produced by the plasma flow reversal. The flow reversal also reduced the impurity retention in the outer divertor, which may produce the partial detachment. Divertor operation margin of key power exhaust parameters to satisfy the peak q_target ≤ 10 MWm⁻² was determined in the low n_e^sep of 2 − 3 × 10¹⁹ m⁻³ under severe conditions such as reducing radiation loss fraction, i.e., f*_rad^div = (P_rad^sol + P_rad^div)/P_sep and diffusion coefficients (χ and D). The divertor geometry and reference parameters (f*_rad^div ~ 0.8, χ = 1 m²s⁻¹, D = 0.3 m²s⁻¹) were consistent with the low n_e^sep operation of the JA DEMO concepts. For either severe assumption of f*_rad^div ~ 0.7 or χ and D to their half values, higher n_e^sep operation was required. In addition, recent investigations of physics models (temperature-gradient force on impurity, photon transport, neutral–neutral collision) under the DEMO relevant SOL and divertor condition are presented.

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan–Europe broader approach DEMO design activity (BA DDA). Common critical issues are the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling (P _rad ^tot/P _heat ⩾ 0.8). Different exhaust concepts in the main plasma and divertor have been developed for Japanese (JA) and European (EU) DEMOs. JA proposed a conventional closed divertor geometry to challenge large P _sep/R _p handling of 30–35 MW m⁻¹ in order to maintain the radiation fraction in the main plasma at the ITER-level (f _rad ^main = P _rad ^main/P _heat ∼ 0.4) and higher plasma performance. EU challenged both increasing f _rad ^main to ∼0.65 and handling the ITER-level P _sep/R _p in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding P _sep = 250–300 MW and 150–200 MW, respectively. Both results showed that large divertor radiation fraction (P _rad ^div/P _sep ⩾ 0.8) was required to reduce both peak q _target (⩽10 MW m⁻²) and T _e,i ^div. In addition, the JA divertor performance with EU-reference P _sep of 150 MW showed benefit of the closed geometry to reduce the peak q _target and T _e,i ^div near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130 °C–200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs.