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The effect of carbon re-erosion in a mixed-material environment on the results of ITER edge plasma modelling is presented. Divertor performance is investigated with varying erosion rate of re-deposited carbon and varying connection length. The results are unified as scaling relations in the edge parameters, allowing meaningful comparison as model parameters are varied and providing effective boundary conditions for core plasma modelling.

A strong effect of the shape of the divertor floor near the separatrix strike point on power loading of the divertor targets in ITER–FEAT is found in the modelling studies. A similar effect has been demonstrated in JET experiments with Mark I and Mark II divertors (A. Loarte, Nucl. Fusion 38 (1998) 587; R. Monk, et al., in: Proceedings of the 24th EPS Conference on Controlled Fusion and Plasma Physics, Berchtesgaden, 1997, vol. 21A, p. 117) thus providing experimental verification of the modelling results. An operational window in six-dimensional phase space is found and different controls, like pumping speed or impurity seeding, are considered for ITER–FEAT.

The operating parameters of the ITER divertor are varied around the partially attached operating point of Refs. [A.S. Kukushkin et al., Berchtesgaden, 1997; A.S. Kukushkin et al., Contributions to Plasma Physics 38 (1998) 20], to obtain an operating window with low heat load, acceptable He pumping and high divertor radiation. Lower heat diffusivities lead to higher heat loads which can be compensated by somewhat lower input power. The divertor geometry is varied: longer domes lead to higher heat loads, whereas the case without dome exhibits poorer helium pumping. Shorter divertor length cases show higher heat load and little change in helium pumping.

The B2-EIRENE code has been compared with plasma edge measurements from Alcator C-MOD, DIII-D, JET, JT-60U in L-mode and H-mode regimes. Similarities in L-mode discharges between JET and JT-60U are well reproduced by the code as well as the sudden approach to detachment seen in JT-60U when a 1% chemical sputtering yield for carbon is assumed. Comparisons between similar L-mode regimes in JET, JT-60U and Alcator C-MOD reveal the larger importance of anomalous energy convection in Alcator C-MOD with regard to larger tokamaks. H-mode discharges of DIII-D and JET can be satisfactorily modelled with similar values of the transport coefficients (approximately an order magnitude smaller than in L-mode) when the power flow into the scrape-off layer (SOL) is dominated by the ion channel, as previous EDGE2D simulations for JET suggested.

Time-dependent theoretical models have been implemented in a 1.5-D transport code to describe the operational boundaries (L–H transition, ballooning limit) in terms of edge parameters. The width of the H-mode transport barrier (F. Wagner et al., Phys. Rev. Lett. 49 (1982) 1408) is self-consistently calculated and is related to α crit . Access to the second stable branch of the ballooning modes is possible at high triangularity. Time dependent trajectories are calculated for several machines and the L- to H-mode transition conditions are compared with global scalings. Calibrated to one experiment (ASDEX-UP), the model yields good agreement with global scaling laws and experimental pedestal electron temperatures, and is then applied to ITER.

Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

This paper describes models that address both power and particle control with the ITER divertor. For power control, the characteristics of the transition from attached to detached divertor operation are studied with an extension of our previous simple model [Y. Igitkhanov et al., 22nd EPS, Bournemouth (1995) p. IV-317]. This new model has been applied to JET experiments to show that the model can reproduce global features of the transition. The model has been applied to ITER to identify the range for detached operation in terms of the heating power, upstream density, and impurity level, and the relation between divertor control and core plasma conditions. For particle control, we use the PRETOR transport code together with a two point divertor model [K. Borrass and G. Janeschitz, Nucl. Fusion 34 (1994) 1203] for the edge boundary conditions to study the sensitivity of the central helium concentration to the pumping speed and helium enrichment in the divertor.

This paper reviews the status of the design of the divertor and first-wall/shield, the main in-vessel components for ITER. Under nominal ignited conditions, 300 MW of alpha power will be produced and must be removed from the divertor and first-wall. Additional power from auxiliary sources up to the level of 100 MW must also be removed in the case of driven burns. In the ignited case, about 100 MW will be radiated to the first wall as bremsstrahlung. Allowing the remaining power to be conducted to the divertor target plates would result in excessive heat fluxes. The power handling strategy is to radiate an additional 100–150 MW in the SOL and the divertor channel via a combination of radiation from hydrogen, and intrinsic and seeded impurities. Vertical targets have been adopted for the baseline divertor configuration. This geometry promotes partial detachment, as found in present experiments and in the results of modelling runs for ITER conditions, and power densities on the target plates can be ≤ 5 MW / m 2 . Such regimes promote relatively high pressure (> 1 Pa ) in the divertor and even with a low helium enrichment factor of 0.2, the required pumping speed to pump helium is ≤ 50 m 3 / s . An important physics question is the quality of core confinement in these attractive divertor regimes. In addition to power and particle handling issues, the effects of disruptions play a major role in the design and performance of in-vessel components. Both centered disruptions and VDE's produce stresses in the first-wall/shield modules, backplate and the divertor wings and cassettes that are near or even somewhat in excess of allowables for normal operation. Also plasma-wall contact from disruptions, including at the divertor target, together with material properties are major factors determining component lifetime. Considering the potential for impurity contamination and minimizing tritium inventory as well as thermomechanical performance, the present material selection calls for carbon divertor targets near the strike point, tungsten on the rest of the target and on the baffle where the charge-exchange flux could be high, and beryllium elsewhere. All three materials and relevant joining techniques are being developed in the R&D program and the final selection for the first assembly will be made at the end of the EDA.

For ITER divertor operation in the detached regime, erosion of divertor plates made of beryllium, carbon fibre composites and tungsten-rhenium alloys is evaluated. Erosion due to (a) physical and, for CFC, chemical sputtering (including surface temperature variation over the lifetime because of neutron irradiation), (b) disruption thermal quench using published data and (c) evaporation and melt layer loss (metals) for infrequent high-power transients (20 MW/m 2 , up to 10 s) is evaluated both with and without the reduction of net incident power by radiation from the evaporated impurities (‘low density vapour shield’). The composite lifetime is calculated for normal operation (detached, 10% transients, 10% disruptions); the effect of an increase in transient frequency and in incident power (semi-attached operation) is determined. The resulting erosion lifetime excludes beryllium, whereas both CFC and W-3%Re yield acceptable lifetimes.

The thermal performance of divertor plates armoured with beryllium (thickness 2–30 mm), carbon fibre composite (CFC, 2–60 mm) or tungsten (W3Re, 2–60 mm) was modelled for transients of 10 s at 20 MW/m 2 — with and without vapour shielding. Due to the high resulting temperatures special attention was paid to the material properties up to and above melting or sublimation. Pure copper was chosen for the flat heat sink of Be and W3Re, the coolant tube material was DS-Cu in all cases. The melt layer depth, the surface temperature and peak power densities to the coolant were obtained in full 2D geometry using finite element modelling. The code was modified to include melting. Evaporation and vapour shielding at elevated temperatures improve the cooling efficiency and reduce the incident power (converting it to radiation, some of which is radiated elsewhere) and thus reduce the depth of melting. For Be armour 10 mm thick, slow plasma transients at 20 MW/m 2 after 10 s will induce melting to a depth of ∼ 2 mm without vapour shielding, and ∼ 0.15 mm with vapour shielding. For up to 60 mm of CFC the peak temperatures attained stay below sublimation; for up to 20 mm of W3Re the surface temperature stay below melting. The lifetime of the divertor plates is determined from the reduction of thickness due to these transients (10% of shots, 10 s), as well as from disruptions (10% of shots) and sputtering. When the maximum or initial armour thickness is given either by the maximum permitted surface temperature during steady state operation at 5 MW/m 2 (which is 1050 K for Be, permitting 11 mm armour thickness, 1780 K for carbon, permitting 40 mm of CFC after neutron irradiation), or in the case of W3Re by the requirement to stay below melting after 10 s at 20 MW/m 2 (permitting 20 mm), and assuming a minimum remaining thickness of 2 mm for all three armour materials, the predicted lifetimes without vapour shielding during transients are as follows: 120–225 shots for Be, 5840–8170 shots for CFC with moderate chemical sputtering, and 2370–7740 shots for W3Re.

The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions Te in the divertor can be reduced to

From 2-D modelling of the scrape-off plasma of a next step device, the scaling of key divertor parameters with density, power, transport coefficients and geometry (pitch, size) is derived. In high-recycling conditions, both power density and electron temperature at the divertor plate decrease strongly with increasing upstream density and decreasing input power. When the upstream density is adjusted to obtain constant, low electron temperature at peak power load, the peak power per unit area is proportional to the input power, and the scaling with connection length and transport coefficients is relatively benign. When divertor depth increases, the temperature still increases strongly with decreasing upstream density and both electron temperature and peak power per unit area decrease. Only for a very short divertor, a thermally unstable region is found (i.e. the upstream density is a multivalued function of plasma temperature at the divertor plate).