Steady-state Operation Scenarios Research Articles

Progress on advanced tokamak and steady-state scenario development on DIII-D and NSTX

Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction fBS ⩾ 60%, q95 ∼ 4–5 and . Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E × B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at βN ⩾ 4 with qmin ⩾ 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. βN = 4, H89 = 2.5, with fBS > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q ⩾ 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

Steady-State Operation Scenarios with a Central Current Hole for JT-60SC

A 1.5-dimensional time-dependent transport analysis has been carried out to investigate steady-state operation scenarios with a central current hole by off-axis current drive schemes consistent with a high bootstrap current fraction for the JT-60SC large superconducting tokamak. A steady-state operation scenario with HHy2 = 1.4 and βN = 3.7 has been obtained at Ip = 1.5 MA, Bt = 2 T, and q95 = 5, where noninductive currents are developed during the discharge to form a current hole with beam-driven currents by tangential off-axis beams in combination with bootstrap currents by additional on-axis perpendicular beams. The bootstrap fraction increases up to ~75% of the plasma current, and the current hole region is enlarged up to ~30% of the minor radius at 35 s from the discharge initiation. The current hole is confirmed to be sustained afterward for a long duration of 60 s. The present transport simulation shows that heating equipment designed for JT-60SC is capable of forming and sustaining the current hole only by using off-axis beam-driven currents and bootstrap currents. The stability analysis shows that the beta limit with the conducting wall can be ~βN = 4.5, which is substantially above the no-wall ideal magnetohydrodynamic limit.

Performance of ITER as a burning plasma experiment

Recent performance analysis has improved confidence in achieving Q (= fusion power/auxiliary heating power)⩾ 10 in inductive operation in ITER. Performance analysis based on empirical scalings shows the feasibility of achieving Q ⩾ 10 in inductive operation, particularly with improved modelling of helium exhaust. Analysis has also indicated the possibility that ITER can potentially demonstrate Q ∼ 50, enabling studies of self-heated plasmas. Theory-based core modelling indicates the need for a high pedestal temperature (3.2–5.3 keV) to achieve Q ⩾ 10, which is in the range of projections with presently available pedestal scalings. Pellet injection from the high-field side would be useful in enhancing Q and reducing edge localized mode (ELM) heat load in high plasma current operation. If the ELM heat load is not acceptable, it could be made tolerable by further tilting the target plate. Steady state operation scenarios at Q = 5 have been developed with modest requirements on confinement improvement and beta (HH98(y,2) ⩾ 1.3 and βN ⩾ 2.6). Stabilization of the resistive wall modes (RWMs), required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structures taken into account. Recent analysis shows a potential of high power steady state operation with a fusion power of 0.7 GW at Q ∼ 8. Achievement of the required βN ∼ 3.6 by RWM stabilization is a possibility. Further analysis is also needed on reduction of the divertor target heat load.

Overview of physics basis for ITER

ITER will be the first magnetic confinement device with burning DT plasma and fusion power of about 0.5 GW. Parameters of ITER plasma have been predicted using methodologies summarized in the ITER Physics Basis (1999 Nucl. Fusion 39 2175). During the past few years, new results have been obtained that substantiate confidence in achieving Q ⩾ 10 in ITER with inductive H-mode operation. These include achievement of a good H-mode confinement near the Greenwald density at high triangularity of the plasma cross section; improvements in theory-based confinement projections for the core plasma, even though further studies are needed for understanding the transport near the plasma edge; improvement in helium ash removal due to the elastic collisions of He atoms with D/T ions in the divertor predicted by modelling; demonstration of feedback control of neoclassical tearing modes and resultant improvement in the achievable β-values; better understanding of edge localized mode (ELM) physics and development of ELM mitigation techniques; and demonstration of mitigation of plasma disruptions. ITER will have a flexibility to operate also in steady-state and intermediate (hybrid) regimes. The ‘advanced tokamak’ regimes with weak or negative central magnetic shear and internal transport barriers are considered as potential scenarios for steady-state operation. The paper concentrates on inductively driven plasma performance and discusses requirements for steady-state operation in ITER.

Noninductive Current Drive and Steady-State Operation in JT-60U

Studies on noninductive current drive (CD) and development of an integrated steady-state high performance operation in JT-60 are reviewed. Experiments on lower hybrid current drive (LHCD) in JT-60 have shown a large noninductive current up to 3.6 MA, high current drive efficiency of 3.5 × 1019 m-2A/W, and a flexible current profile control. Basic studies on LH waves, such as an effect of accessibility condition, fast electron behaviors, and so on, in JT-60 have contributed to understanding LHCD physics. Significant progress in neutral beam current drive (NBCD) has been made in JT-60 by testing the performance of negative ion-based (N) neutral beam injection (NBI) (N-NBI). The CD efficiency of ~1.5 × 1019 m-2A/W and negative ion-based neutral beam (N-NB) driven current of ~1 MA have been demonstrated in N-NBCD. Strongly localized noninductive driven current by electron cyclotron current drive (ECCD) was identified with a fundamental O-mode scheme from a low field side injection. ECCD in JT-60 has shown CD efficiency of 0.5 × 1019 m-2A/W and EC-driven current of 0.2 MA. Modification of local current profile was demonstrated and was used for suppression of neoclassical tearing mode. Based on these developments, two integrated steady-state operation scenarios were developed in JT-60, which are reversed magnetic shear (R/S) plasmas and high βp ELMy H-mode. In these operation regimes, discharges have been sustained near the steady-state current profile under full noninductive current drive (High βp; HHy2 ~ 1.4 and βN ~ 2.5 with N-NB, R/S; HHy2 ~ 2.2 and βN ~ 2 with fBS ~ 80%). High performance plasmas with a high nDoτETio and at high normalized density were also produced under fully noninductive condition in high βp ELMy H-mode and R/S mode.

Profile control for steady-state operation

Large departures from standard L-mode confinement scaling laws are now being observed in most tokamaks which appear promising for relaxing the plasma current constraint and therefore open the route to high-performance steady-state operation with high bootstrap current fraction. Prospects of the weak or reversed magnetic shear scenarios for steady-state tokamak operation are presented in the light of recent experimental progress to control pressure and current profiles with the combined use of various heating and current drive schemes. Implications of the present experiments on the modelling of future non-inductive operation in existing tokamaks and in ITER are addressed.

Comparison of ITER Single-Null and Double-Null Operation

The nominal International Thermonuclear Experimental Reactor (ITER) configuration is a double-null (DN) divertor, which requires precise plasma vertical position control. Vertical displacements of only about 1 cm (out of a plasma height of 4.7 m) are estimated to destroy the up/down symmetric distribution of power flow to the divertor plates. As an alternate configuration to avoid this difficulty, we look at the single-null (SN) option, where all the charged power flow is deposited on the lower divertor plate. The primary consideration in this study is that of technology phase performance (maximum neutron wall load) for the ITER divertor heat load and plasma constraints. With regard to the divertor heat loads, the SN case has the advantages of (a) longer scrape-off field line connection lengths and (b) more vertical space, which allows a greater spreading of the heat load on the divertor plates. These advantages offset the SN case disadvantage of having fewer divertor plates, and therefore the potential for higher heat fluxes for a given core plasma condition. The attainable wall loads for the SN and DN divertors are found to be similar for steady-state and hybrid operation scenarios.