J.H. Schultz

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

The design of the KSTAR tokamak

Abstract The Korea Superconducting Tokamak Advanced Research (KSTAR) Project is the major effort of the Korean National Fusion Program (KNFP) to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. Major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 Tesla, and plasma current 2 mA with a strongly shaped plasma cross-section and double-null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beam, ion cyclotron waves, lower hybrid waves, and electron–cyclotron waves for flexible profile control. A comprehensive set of diagnostics is planned for plasma control and performance evaluation and physics understanding. The project has completed its conceptual design phase and moved to the engineering design phase. The target date of the first plasma is set for year 2002.

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 ARIES-CS COMPACT STELLARATOR FUSION POWER PLANT

Abstract An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore attractive compact stellarator configurations and to define key research and development (R&D) areas. The large size and mass predicted by earlier stellarator power plant studies had led to cost projections much higher than those of the advanced tokamak power plant. As such, the first major goal of the ARIES-CS research was to investigate if stellarator power plants can be made to be comparable in size to advanced tokamak variants while maintaining desirable stellarator properties. As stellarator fusion core components would have complex shapes and geometry, the second major goal of the ARIES-CS study was to understand and quantify, as much as possible, the impact of the complex shape and geometry of fusion core components. This paper focuses on the directions we pursued to optimize the compact stellarator as a fusion power plant, summarizes the major findings from the study, highlights the key design aspects and constraints associated with a compact stellarator, and identifies the major issues to help guide future R&D.

The Levitated Dipole Experiment (LDX) magnet system

In the Levitated Dipole Experiment (LDX), a hot plasma is formed about a levitating superconducting dipole magnet in the center of a 5 m diameter vacuum vessel. The levitated magnet is suspended magnetically during an eight hour experimental run, then lowered and recooled overnight. The floating F-coil magnet consists of a layer-wound magnet with 4 sections, designed to wrap flux lines closely about the outside of the levitated cryostat. The conductor is a niobium-tin Rutherford cable, with enough stabilizer to permit passive quench protection. Lead strips are used as thermal capacitors to slow coil heating. An optimized system of bumpers and cold-mass supports reduces heat leak into the helium vessel. Airbags catch the floating coil on quenches and faults, preventing collision with the vacuum vessel.

Spike voltages seen during "quick charge" ramp limitation tests on Nb/sub 3/Sn cable-in-conduit conductors

Spike voltages observed during ramp rate limitation tests on sub-sized Nb/sub 3/Sn cable-in-conduit superconductors are analyzed using current loop model. The effects of loop currents on the ramp limitations of multi strand superconducting cables are discussed. Current loops existing in multi strand cables generate excess local currents that quench strands and produce voltage spikes. Experimental results previously reported as abnormal ramp rate limitations are explained by loop current phenomena.

Ramp-rate limitation experiments using a hybrid superconducting cable

Abstract Ramp-rate limitation experiments were done in a new facility at the MIT (Massachusetts Institute of Technology) Plasma Fusion Center. The features of this new facility include (1) a superconducting pulse coil that can superimpose high ramp-down rates, up to 25 T s−1, (2 T in 80 ms) at a background field up to 5 T, (2) new power supplies that can supply high rates of dl dt and dB dt to the sample under test and (3) a forced-flow supercritical helium system for cooling CICCs (Cable-In-Conduit Conductors). This paper discusses the results of the ramp-rate limitation experiments on a 27-strand hybrid Nb3Sn cable. The cable was tested under field ramps of up to 2.5 T s−1 with various operating currents. It did not quench with dB dt , field and average strand currents that were simultaneously above the operating range of TPX-PF (Tokamak Physics Experiment Poloidal Field) coils. Further ramp-rate limitation experiments revealed that the tested 27-strand hybrid cable has very high transient stability at ramped fields, extending out to average strand currents that are nearly triple the TPX-PF operating current.

Ramp-rate limitation test of cable-in-conduit conductors with supercritical helium

It has been found on the United States Demonstration Poloidal Coil (US-DPC) and in 27 strand subsized cables of pool boiling cable-in-conduit conductor (CICC), that there is critical current degradation due to fast ramping of the magnetic field. The characteristics of this ramp-rate limitation phenomenon are investigated by using a 27 strand Nb/sub 3/Sn cable in supercritical helium at 6 atm. A 3 m long cable-in-conduit conductor is prepared noninductively and tested in a background field up to 9.5 tesla with maximum ramp rate of 1.6 tesla/second. The ramp-rate limitation results are compared with results of the ramp rate test of the US-DPC and previous experiments. The experimental data are analyzed to identify and understand possible sources of ramp-rate limitation.<<ETX>>

Protection of superconducting magnets

The protection of superconducting magnets can be narrowly considered as the problem of rapidly detecting an irreversible loss of superconductivity (quench) and converting the magnetic stored energy to thermal energy (dump) without permanent damage (protection). It can also be broadly considered as a method for avoiding all structural and electrical failure mechanisms through design, fabrication, and operation of a magnet. Energy dump and quench detection methods are reviewed. General magnet failure is surveyed historically.

The KSTAR tokamak

The KSTAR (Korea Superconducting Tokamak Advanced Research) project is the major effort of the Korean National Fusion Program to design, construct, and operate a steady-state-capable superconducting tokamak. The project is led by Korea Basic Science Institute and shared by national laboratories, universities, and industry along with international collaboration. It is in the conceptual design phase and aims for the first plasma by mid 2002. The key design features of KSTAR are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T, plasma current 2 MA, and flexible plasma shaping (elongation 2.0; triangularity 0.8; double-null poloidal divertor). Both the toroidal and the poloidal field magnets are superconducting coils. The device is configured to be initially capable of 20 s pulse operation and then to be upgraded for operation up to 300 s with non-inductive current drive. The auxiliary heating and current drive system consists of neutral beam, ICRF, lower hybrid, and ECRF. Deuterium operation is planned with a full radiation shielding.