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This paper presents the results of a modelling study conducted to estimate the power crossing the separatrix (P SOL ) in the ITER device during a standard start-up sequence. This is used to calculate the power intercepted by the start-up limiters and the resulting power load distribution. The models and methodologies applied to calculate P SOL and the power loads on the limiters are described in detail elsewhere ([e.g., M. Kobayashi et al., Nucl. Fusion. 47 (2) (2007) 61]) and only a brief mention of some of the main results is included here. These assessments show that for the range of conditions analysed, the maximum P SOL intercepted by the two ITER limiter start-up modules during the current ramp-phase is ∼6 MW. The peak power load to each limiter is calculated to be ∼5 MW/m 2 , but these values depends on assumptions on physical quantities (e.g., transport coefficients, i.e., D ⊥ and Χ ⊥ ), which are uncertain and still await confirmation by experiments. Recommendations are made for modelling and experiments to extend the study presented here.

The values of Q = (fusion power)/(auxiliary heating power) predicted for ITER by three different methods are compared. The first method utilizes an empirical confinement-time scaling and prescribed radial profiles of transport coefficients; the second approach extrapolates from specially designed ITER similarity experiments, and the third approach is based on partly theory-based transport models. The energy confinement time given by the ITERH-98(y, 2) scaling for an inductive scenario with a plasma current of 15 MA and a plasma density 15% below the Greenwald density is 3.7 s with one estimated technical standard deviation of ±14%. This translates, in the first approach, for levels of helium removal, and impurity concentration, that, albeit rather stringent, are expected to be attainable, into an interval for Q of [6–15] at the auxiliary heating power, Paux = 40 MW, and [6–30] at the minimum heating power satisfying a good confinement ELMy H-mode. All theoretical transport-model calculations have been performed for the plasma core only, whereas the pedestal temperatures were taken as estimated from empirical scalings. Predictions of similarity experiments from JET and of theory-based transport models that we have considered—Weiland, MMM, and IFS/PPPL—overlap with the prediction using the empirical confinement-time scaling within its estimated margin of uncertainty.

In the last few years, significant progress has been made in the understanding of H-mode plasmas (e.g. ion temperature profile stiffness, pedestal physics, etc). Based on this improved understanding, a set of rules (models) comprising a physics picture of the H-mode has been implemented in the ASTRA code in order to improve the understanding of experimental observations and ultimately to provide a predictive capability for ITER complementary to the scaling relations. The model has been verified for consistency with experimental observations in ASDEX-UG and JET plasmas. Numerical coefficients for the transport, required because of simplifications or missing quantitative information, are determined for one plasma (e.g. from JET) and then held constant for all others (JET, ASDEX-UP, ITER). After benchmarking the model to experimental results, it was also applied to ITER. It predicts that Q = 10 can be achieved in ITER but only with at least a 50% deep fuelling contribution (inside the H-mode pedestal). However, in existing machines as well as in our model runs for existing machines, gas puffing is sufficient to achieve the observed density pedestal and line average densities. A second important result from the predictive runs for ITER is that electron energy transport in the plasma core, the neoclassical transport in the pedestal and the CX losses at the plasma edge are important constraints for a better performance. Thus future theoretical and experimental work should concentrate on these areas in order to improve our predictions.