Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor
- JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)
- Belgium
- CEA, IRFM, Cadarache F-13108 Saint-Paul-l Lez-Durance (France)
- Japan Atomic Energy Agency, 801-1 Muko-yama, Naka, Ibaraki 311-0193 (Japan)
- Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States)
- General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)
- ITER Organization, 13115 Saint Paul Lez Durance (France)
- Max-Planck-Institut für Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany)
The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ∼10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ∼ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q{sub 95} ∼ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at n{sub e}/n{sub GW} ∼ 0.85. Most experiments, however, obtain stable discharges at H{sub 98(y,2)} ∼ 1.0 only for β{sub N} = 2.0–2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (l{sub i}(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13–15 MA and densities down to n{sub e}/n{sub GW} ∼ 0.5, long pulse operation (>1000 s) in ITER is possible at Q ∼ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (>50 MW). H-mode operation in helium may be possible at input powers above 35 MW at a toroidal field of 2.65 T, for studying H-modes and ELM mitigation. In hydrogen, H-mode operation is expected to be marginal, even at 2.65 T with 60 MW of input power. Simulation code benchmark studies using hybrid and steady state scenario parameters have proved to be a very challenging and lengthy task of testing suites of codes, consisting of tens of sophisticated modules. Nevertheless, the general basis of the modelling appears sound, with substantial consistency among codes developed by different groups. For a hybrid scenario at 12 MA, the code simulations give a range for Q = 6.5–8.3, using 30 MW neutral beam injection and 20 MW ICRH. For non-inductive operation at 7–9 MA, the simulation results show more variation. At high edge pedestal pressure (T{sub ped} ∼ 7 keV), the codes predict Q = 3.3–3.8 using 33 MW NB, 20 MW EC, and 20 MW ion cyclotron to demonstrate the feasibility of steady-state operation with the day-1 heating systems in ITER. Simulations using a lower edge pedestal temperature (∼3 keV) but improved core confinement obtain Q = 5–6.5, when ECCD is concentrated at mid-radius and ∼20 MW off-axis current drive (ECCD or LHCD) is added. Several issues remain to be studied, including plasmas with dominant electron heating, mitigation of transient heat loads integrated in scenario demonstrations and (burn) control simulations in ITER scenarios.
- OSTI ID:
- 22423774
- Journal Information:
- Physics of Plasmas, Vol. 22, Issue 2; Other Information: (c) 2014 Author(s); Country of input: International Atomic Energy Agency (IAEA); ISSN 1070-664X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
AUXILIARY HEATING
BENCHMARKS
BREAKDOWN
COMPUTERIZED SIMULATION
DEUTERIUM
ECR CURRENT DRIVE
ECR HEATING
EDGE LOCALIZED MODES
HEATING LOAD
HELIUM
H-MODE PLASMA CONFINEMENT
HYDROGEN
ICR HEATING
ITER TOKAMAK
KEV RANGE
METALS
NEUTRAL ATOM BEAM INJECTION
NEUTRON FLUENCE
PLASMA SIMULATION
TRITIUM