The operational space (I p − n) for long-pulse scenarios ( t burn ∼ 1000 s, Q 5) of ITER has been assessed by 1.5D core transport modelling with pedestal parameters predicted by the EPED1 code by a set of transport codes under a joint activity carried out by the Integrated Operational Scenario ITPA group. The analyses include the majority of transport models (CDBM, GLF23, Bohm/gyroBohm (BgB), MMM7.1, MMM95, Weiland, scaling-based) presently used for interpretation of experiments and ITER predictions. The EPED1 code was modified to take into account boundary conditions predicted by SOLPS4 for ITER. In contrast to standard EPED1 assumptions, EPED1 with the SOLPS boundary conditions predicts no degradation of the pedestal pressure as density is reduced. Lowering the plasma density to n e ∼ (5-6) × 10 19 m −3 leads to an increased plasma temperature (similar pedestal pressure), which reduces the loop voltage and increases the duration of the burn phase to t burn ∼ 1000 s with Q 5 for I p 13 MA at moderate normalized pressure (β N ∼ 2). These ITER plasmas require the same level of additional heating power as the reference Q = 10 inductive scenario at 15 MA (33 MW NBI and 17-20 MW EC heating and current drive power). However, unlike the 'hybrid' scenarios considered previously, these H-mode plasmas do not require specially shaped q profiles nor improved confinement in the core for the transport models considered in this study. Thus, these medium density H-mode plasma scenarios with I p 13 MA present an attractive alternative to hybrid scenarios to achieve ITER's long-pulse Q 5 scenario and deserve further analysis and experimental demonstration in present tokamaks.

This paper summarises the modelling studies of steady-state divertor operation being performed for the ITER-FEAT design. Optimisation of the divertor geometry reveals the importance of the proper target shape for a reduction of the peak power loads. A high gas conductance between the divertor legs is also essential for maintaining acceptable conditions in the outer divertor which receives higher power loading than the inner. Impurity seeding, which would be necessary if tritium co-deposition concerns preclude the use of carbon as plasma-facing material, can ensure the required high radiation level at acceptable Z eff , and the divertor performance is not very sensitive to the choice of the radiating impurity.

Abstract. Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha (EDA)H

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H 96-L = 0.6 or H IPB98 = 0.4) has been validated on existing experiments for predicting the li dynamics within +/-0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than +/-0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of I p = 15MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free boundary equilibrium solvers and a Poloidal Field (PF) systems controller are also described, focusing on ITER current ramp-down.

ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1+SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5-7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W redeposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on W transport assumptions during the ELM, a maximum ELM frequency is also identified above which core tungsten accumulation takes place.