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.

The hybrid operating mode observed in several tokamaks is characterized by further enhancement over the high plasma confinement (H-mode) associated with reduced magneto-hydro-dynamic (MHD) instabilities linked to a stationary flat safety factor () profile in the core region. The proposed ITER hybrid operation is currently aiming at operating for a long burn duration (>1000 s) with a moderate fusion power multiplication factor, , of at least 5. This paper presents candidate ITER hybrid operation scenarios developed using a free-boundary transport modelling code, CORSICA, taking all relevant physics and engineering constraints into account. The ITER hybrid operation scenarios have been developed by tailoring the 15 MA baseline ITER inductive H-mode scenario. Accessible operation conditions for ITER hybrid operation and achievable range of plasma parameters have been investigated considering uncertainties on the plasma confinement and transport. ITER operation capability for avoiding the poloidal field coil current, field and force limits has been examined by applying different current ramp rates, flat-top plasma currents and densities, and pre-magnetization of the poloidal field coils. Various combinations of heating and current drive (H&CD) schemes have been applied to study several physics issues, such as the plasma current density profile tailoring, enhancement of the plasma energy confinement and fusion power generation. A parameterized edge pedestal model based on EPED1 added to the CORSICA code has been applied to hybrid operation scenarios. Finally, fully self-consistent free-boundary transport simulations have been performed to provide information on the poloidal field coil voltage demands and to study the controllability with the ITER controllers.

The International Tokamak Physics Activity topical group on integrated operational scenarios has compiled a database of stationary H-mode discharges at q95 ~ 3 from AUG, C-Mod, DIII-D, JET and JT-60U, for both carbon wall and high-Z metal wall experiments with ~3300 entries. The analyses focus on discharges that are stationary for  ⩾5 thermal energy confinement times to evaluate the baseline scenario proposed for ITER at 15 MA for achieving its goals of Q  =  10, fusion power of 500 MW at normalised pressure, βN  =  1.8 and normalised confinement as predicted by the standard H-mode scaling, H98y2  =  1. With the data restricted to stationary H-modes at q95 ~ 3, the database shows significant variation of thermal energy confinement compared to the standard H-mode scaling (IPB98(y,2)) in dimensionless form. The data show similar scaling with normalised gyro-radius, but more favourable scaling towards lower collision frequency and more favourable scaling with plasma beta. Using all the engineering variables employed in IPB98(y,2), results in an overfit due to correlations among the data. Moreover, there are significant residual trends in the confinement for plasma current, device size, loss power, and in particular for the plasma density. Significant differences between results obtained for devices with a carbon wall and high-Z metal wall are observed in the data, with data from carbon wall devices providing a larger operating space, encompassing ITER parameters or even exceeding them. H-modes in high-Z metal wall devices have, so-far, not accessed conditions at low collision frequencies, have lower normalised confinement (H98y2 ~ 0.8–0.9) at low input power or beta, achieving H98y2 ~ 1.0 only at input powers two times the L- to H-mode transition scaling predictions and at βN ~ 2.0. Hence, only the best H-modes with high-Z metal walls reach ITER baseline performance requirements. The data show that operating at high plasma density, with line-averaged density at 85% of Greenwald density is achievable for H98y2  >  0.95 for a range of plasma configurations, and that operation at low plasma inductance with li(3) ~ 0.7–0.75 is feasible. Scenario simulations employed for projecting the plasma performance in ITER should incorporate a lower thermal confinement at low plasma beta for the entry to burn and provide projections using higher levels of plasma core radiation by plasma impurities. Moreover, ITER projections should not subtract the core radiation in the evaluation of the thermal confinement time and H98y2, to allow a fair comparison with experimental data currently available. From the data presented here, it is likely that in ITER the energy confinement time will not increase with plasma density and will have no degradation with plasma beta. The analyses indicate that the data at q95 ~ 3 are consistent with achievement of the ITER mission goals at 15 MA.

the Alcator C-MOD team 16 , the ASDEX Upgrade team 14 , the DIII-D team 2 , the EAST team 5 , JET contributors a , the KSTAR team 3 , the NSTX-U team 4 and the TCV team 12 and ITPA IOS members and experts

The successful performance of a model predictive profile controller is demonstrated in simulations and experiments on the TCV tokamak, employing a profile controller test environment. Stable high-performance tokamak operation in hybrid and advanced plasma scenarios requires control over the safety factor profile (q-profile) and kinetic plasma parameters such as the plasma beta. This demands to establish reliable profile control routines in presently operational tokamaks. We present a model predictive profile controller that controls the q-profile and plasma beta using power requests to two clusters of gyrotrons and the plasma current request. The performance of the controller is analyzed in both simulation and TCV L-mode discharges where successful tracking of the estimated inverse q-profile as well as plasma beta is demonstrated under uncertain plasma conditions and the presence of disturbances. The controller exploits the knowledge of the time-varying actuator limits in the actuator input calculation itself such that fast transitions between targets are achieved without overshoot. A software environment is employed to prepare and test this and three other profile controllers in parallel in simulations and experiments on TCV. This set of tools includes the rapid plasma transport simulator RAPTOR and various algorithms to reconstruct the plasma equilibrium and plasma profiles by merging the available measurements with model-based predictions. In this work the estimated q-profile is merely based on RAPTOR model predictions due to the absence of internal current density measurements in TCV. These results encourage to further exploit model predictive profile control in experiments on TCV and other (future) tokamaks.