Charles Kessel

Overview of the ARIES-RS reversed-shear tokamak power plant study

The ARIES-RS tokamak is a conceptual, D‐T-burning 1000 MWe power plant. As with earlier ARIES design studies, the final design of ARIES-RS was obtained in a self-consistent manner using the best available physics and engineering models. Detailed analyses of individual systems together with system interfaces and interactions were incorporated into the ARIES systems code in order to assure self-consistency and to optimize towards the lowest cost system. The ARIES-RS design operates with a reversed-shear plasma and employs a moderate aspect ratio (A4.0). The plasma current is relatively low (Ip11.32 MA) and bootstrap current fraction is high ( fBC 0.88). Consequently, the auxiliary power required for RF current drive is relatively low ( 80 MW). At the same time, the average

IDEAL MHD STABILITY LIMITS OF LOW ASPECT RATIO TOKAMAK PLASMAS

The ideal magnetohydrodynamic (MHD) stability limits of low aspect ratio tokamak plasmas are computed numerically for plasmas with a range of cylindrical safety factors q*, normalized plasma pressures beta , elongations kappa and central safety factors q(0). Four distinct regimes are optimized, namely: (a) low-q* plasmas with q(0)=1.1 with and without a stabilizing wall, (b) low-q* plasmas with no wall and 1.1<q(0)<2, (c) high- beta , high bootstrap fraction plasmas at moderate kappa requiring a wall and edge current drive and (d) high- beta , very high bootstrap fraction plasmas with moderate to high kappa requiring a stabilizing wall but little external current drive. A stable equilibrium is found at an aspect ratio of A=1.4 and an elongation of kappa =3.0, with 99.3% of the current provided by the plasma pressure and beta =45%. Special attention is paid to the issues of numerical convergence and the proper definition of bootstrap current fraction

Physics design of the national spherical torus experiment

The mission of the National Spherical Torus Experiment (NSTX) is to prove the principles of spherical torus physics by producing high-beta toroidal plasmas that are non-inductively sustained, and whose current profiles are in steady-state. NSTX will be one of the first ultra low a[P(input) up to 11 MW] in order to produce high-beta toroidal (25 to 40%), low collisionality, high bootstrap fraction (less than or equal to 70%) discharges. Both radio-frequency (RF) and neutral-beam (NB) heating and current drive will be employed. Built into NSTX is sufficient configurational flexibility to study a range of operating space and the resulting dependences of the confinement, micro- and MHD stability, and particle and power handling properties. NSTX research will be carried out by a nationally based science team.

The ARIES-I Tokamak Reactor Study †

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-3He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.

THE PROSPECTS FOR MAGNETOHYDRODYNAMIC STABILITY IN ADVANCED TOKAMAK REGIMES

Stability analysis of advanced regime tokamaks is presented. Here advanced regimes are defined to include configurations where the ratio of the bootstrap current, IBS, to the total plasma current, Ip, approaches unity, and the normalized stored energy, βN* = 80π〈p2〉1/2a/IpB0, has a value greater than 4.5. Here, p is the plasma pressure, a the minor radius in meters, Ip is in mega‐amps, B0 is the magnetic field in Tesla, and 〈⋅〉 represents a volume average. Specific scenarios are discussed in the context of Toroidal Physics Experiment (TPX) [Proceedings of the 20th European Physical Society Conference on Controlled Fusion and Plasma Physics, Lisbon, 1993, edited by J. A. Costa Cabral, M. E. Manso, F. M. Serra, and F. C. Schuller (European Physical Society, Petit‐Lancy, 1993), p. I‐80]. The best scenario is one with reversed shear, in the q profile, in the central region of the tokamak. The bootstrap current obtained from the plasma profiles provides 90% of the required current, and is well aligned with the...

Predictions of H-mode performance in ITER

Time-dependent integrated predictive modelling is carried out using the PTRANSP code to predict fusion power and parameters such as alpha particle density and pressure in ITER H-mode plasmas. Auxiliary heating by negative ion neutralbeaminjectionandion-cyclotronheatingofHe 3 minorityionsaremodelled,andtheGLF23transportmodelis used in the prediction of the evolution of plasma temperature profiles. Effects of beam steering, beam torque, plasma rotation, beam current drive, pedestal temperatures, sawtooth oscillations, magnetic diffusion and accumulation of He ash are treated self-consistently. Variations in assumptions associated with physics uncertainties for standard base-line DT H-mode plasmas (with Ip = 15MA, BTF = 5.3T and Greenwald fraction = 0.86) lead to a range of predictions for DT fusion power PDT and quasi-steady state fusion QDT (≡PDT/Paux). Typical predictions assuming Paux = 50‐53MW yield PDT = 250‐720MW and QDT = 5‐14. In some cases where Paux is ramped down or shut off after initial flat-top conditions, quasi-steady QDT can be considerably higher, even infinite. Adverse physics assumptions such as the existence of an inward pinch of the helium ash and an ash recycling coefficient approaching unity lead to very low values for PDT. Alternative scenarios with different heating and reduced performance regimes are also considered including plasmas with only H or D isotopes, DT plasmas with toroidal field reduced 10% or 20% and discharges with reduced beam voltage. In full-performance D-only discharges, tritium burn up is predicted to generate central tritium densities up to 10 16 m −3 and DT neutron rates up to 5 ×10 16 s −1 , compared with the DD neutron rates of 6 × 10 17 s −1 . Predictions with the toroidal field reduced 10% or 20% below the planned 5.3T and keeping the same q98, Greenwald fraction and βn indicate that the fusion yield PDT and QDT will be lower by about a factor of two (scaling as B 3.5 ).

Bootstrap current in a tokamak

The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and βp must be kept below a critical value

Experimental studies of ITER demonstration discharges

Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has been verified in dedicated experiments. New data are obtained from C-Mod, ASDEX Upgrade, DIII-D, JT-60U and JET. Results show that breakdown for Eaxis < 0.23–0.33 V m−1 is possible unassisted (ohmic) for large devices like JET and attainable in devices with a capability of using ECRH assist. For the current ramp up, good control of the plasma inductance is obtained using a full bore plasma shape with early X-point formation. This allows optimization of the flux usage from the PF set. Additional heating keeps li(3) < 0.85 during the ramp up to q95 = 3. A rise phase with an H-mode transition is capable of achieving li(3) < 0.7 at the start of the FT. Operation of the H-mode reference scenario at q95 ~ 3 and the hybrid scenario at q95 = 4–4.5 during the FT phase is documented, providing data for the li (3) evolution after the H-mode transition and the li (3) evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation. The inductance could be kept ≤1.2 during the first half of the current decay, using a slow Ip ramp down, but still consuming flux from the transformer. Alternatively, the discharges can be kept in H-mode during most of the ramp down, requiring significant amounts of additional heating.

Advanced tokamak physics-status and prospects

Experimental and theoretical results from around the world point to the possibility of high confinement, high- beta , and high-bootstrap-fraction steady-state tokamak operating modes. These modes of operation, if fully developed and extended to steady-state, could lead to much less expensive tokamak demonstration power reactors and to a significantly reduced cost-of-electricity from fusion, as compared to projections based on low- beta N, pulsed operating modes. Present results have clear implications in the areas of particle control, plasma shaping, and current-profile control. Thus they have strongly influenced the design of the steady-state advanced tokamak TPX, which has the mission to combine the best results from present experiments and extend them to steady-state. These results also have important implications for follow-up tests in ITER, which have the goal of studying advanced-tokamak operation in an ignited plasma, as well as for the eventual configuration of an advanced-tokamak fusion reactor.

Physics Design of a High-beta Quasi-axisymmetric Stellarator

Note: 8th Toki 11th International Stellarator Conference, Toki-City, Japan, September/October 1997, Proc. published in J. Plasma and Fusion Res., SERIES, Vol. 1, 429 - 432 (1998) Reference CRPP-CONF-1998-055 Record created on 2008-05-13, modified on 2016-08-08