As part of the ITER Design Review, the physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

The impacts of plasma disruptions on ITER have been investigated in detail to confirm the robustness of the design of the machine to the potential consequential loads. The loads include both electro-magnetic (EM) and heat loads on the in-vessel components and the vacuum vessel. Several representative disruption scenarios are specified based on newly derived physics guidelines for the shortest current quench time as well as the maximum product of halo current fraction and toroidal peaking factor arising from disruptions in ITER. Disruption simulations with the DINA code and EM load analyses with a 3D finite element method code are performed for these scenarios. Some margins are confirmed in the EM load on in-vessel components due to induced eddy and halo currents for these representative scenarios. However, the margins are not very large. The heat load on various parts of the first wall due to the vertical movement and the thermal quench (TQ) is calculated with a 2D heat conduction code based on the database of heat deposition during disruptions and simulation results with the DINA code. For vertical displacement event, it is found that the beryllium (Be) wall does not melt during the vertical movement, prior to the TQ. Significant melting is anticipated for the upper Be wall and the tungsten divertor baffle due to TQ after the vertical movement. However, its impact could be substantially mitigated by implementing a reliable detection system of the vertical movement and a mitigation system, e.g. massive noble gas injection. Some melting of the upper Be wall is anticipated at major disruptions. At least several tens of unmitigated disruptions must be considered even if an advanced prediction/mitigation system is implemented. With these unmitigated disruptions, the loss of the Be layer is expected to be within ≈30–100 µm/event out of a 10 mm thick Be first wall.

The helium ash exhaust function of a divertor has been experimentally demonstrated. Helium atoms accumulate in the divertor region as the electron density of the main plasma increases. With a helium concentration of ~ 1.6% of electron density in the main plasma, neutral helium pressure at the divertor region is as high as 1.0 x 10" 4 Torr. This experiment indicates the possibility of helium ash exhaust in an a-particleheated diverted tokamak with use of pumping ducts of a practical size.

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The successful operation of a single-null poloidal divertor in Doublet-Ill has demonstrated several new advantages of a diverted tokamak in addition to the suppression of impurity influx as demonstrated in DIVA: 1) The impurity contamination and radiation loss of the main plasma has been reduced by an open divertor geometry, i.e. without a divertor chamber; 2) The radiative cooling and formation of a dense and cold (n e >5Xl0 13 crn~3, T e <7 eV, P H , < 4 X 10' 3 torr) divertor plasma have been observed. -Up to 50% of the Ohmic input power is radiated in the divertor region, thus cooling the plasma in front of the divertor plate down to several eV. This remote radiative cooling greatly reduces the heat load on the divertor plate without cooling the main plasma. -The feasibility of remote radiative cooling in INTOR was studied by use of a volume integration technique of the radiation power along the field line.

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Stable plasmas with surface elongations of up to 1.8 (aspect ratio = 3.4) have been produced in the upper lobe of Doublet III with the use of both passive and active controls. The growth rate of the vertical instability has been measured at various values of elongation by disabling the feedback circuit of the vertical position control power supply. A rigid-shift analysis of growth rates indicates that the passive stabilization effect of the field-shaping coils plays a key role in achieving a high elongation of 1.8. Experiments have demonstrated that the maximum stable elongation is determined by the strength of the passive stabilization effect even with active feedback control. The dee-shape is found to be preferable to an elliptical shape because the triangularity reduces the absolute value of the decay index required to produce a given elongation.