J. Galambos

Results of an International Study on a High-Volume Plasma-Based Neutron Source for Fusion Blanket Development

AbstractAn international study conducted by technical experts from Europe, Japan, Russia, and the United States has evaluated the technical issues and the required testing facilities for the develo...

Commercial tokamak reactor potential with advanced tokamak operation

The attractiveness of future commercial tokamak reactors is sensitive to the attainable plasma performance, notably plasma energy confinement and allowable beta. The impact of varying levels of confinement and beta on the size and cost of the resulting tokamak reactor is systematically quantified. Several different classes of tokamak reactors are considered, and designs are optimized in terms of cost of electricity (COE) via a coupled physics/engineering/costing systems code. Surprisingly narrow ranges of plasma confinement and beta are found to be simultaneously useful in minimizing the reactor COE, i.e. improvement in only one of these quantities is not useful beyond some point without accompanying improvements in the other. For steady state, current driven reactors characterized by H mode confinement (where τE=HτE,L; τE,L being the confinement time predicted by the ITER.89 L mode scaling, and H ~ 2), the maximum useful Troyon β coefficient (βN) is only ~ 4.3%.mT/MA. These confinement levels are similar to those observed in present day experiments. If slightly better confinement is achievable (i.e. an enhancement factor over L mode of H ~ 2.5), the maximum useful Troyon coefficient increases to βN ~ 6 and the reactor COE decreases by 20%. Inductively driven, pulsed reactors have somewhat increased useful ranges of confinement relative to the steady state cases. In general, increasing the allowable beta over presently accepted limits offers the single biggest improvement in reactor attractiveness of the tokamak concept

A ``SUPERCODE`` for systems analysis of tokamak experiments and reactors

A new code, named SUPERCODE, has been developed to fill the gap between currently available zero dimensional systems codes and highly sophisticated, multidimensional plasma performance codes. The former are comprehensive in content, fast to execute, but rather simple in terms of the accuracy of their physics and engineering models. The latter contain state-of-the-art plasma physics modeling but are limited in engineering content and are time consuming to run. The SUPERCODE upgrades the reliability and accuracy of systems codes by calculating the self consistent 1 1/2-D plasma evolution in a realistic engineering environment. By a combination of variational techniques and careful formulation there is only a modest increase in CPU time over 0-D runs, thereby making the SUPERCODE suitable for use as a systems studies tool. In addition, we have expended considerable effort to make the code user- and programmer friendly, as well as operationally flexible, with the hope of encouraging wide usage throughout the fusion community.

XAL Application Programming Structure

XAL is an application programming framework used at the Spallation Neutron Source (SNS) project in Oak Ridge. It is written in Java, and provides users with a hierarchal view of the accelerator. Features include database configuration of the accelerator structure, an online envelope model that is configurable from design or live machine values, an application framework for quickstart GUI development, a scripting interface for algorithm development, and a common toolkit for shared resources. To date, about 25 applications have been written, many of which are used extensively in the SNS beam commissioning activities. The XAL framework and example applications will be discussed.

ORBIT-a ring injection code with space charge

ORBIT (Objective Ring Beam Injection and Tracking) is a new particle tracking code for rings. Modelling capabilities include H/sup -/ foil injection mechanisms, longitudinal and transverse space charge effects, and second order matrix transport. Additional code features include a programmable interactive driver shell, and interactive plotting.

Study of a spherical tokamak based volumetric neutron source

Abstract With the worldwide development of fusion power focusing on the design of the International Thermonuclear Experimental Reactor (ITER), developmental strategies for the demonstration fusion power plant (DEMO) are being discussed. A relatively prudent strategy is to construct and operate a small deuterium–tritium fuelled volumetric neutron source (VNS) in parallel with ITER. The VNS is to provide, over a period less than 20 years, a relatively high fusion neutron fluence of 6 MW year m−2 and wall loading of 1 MW m−2 or more, over an accessible blanket test area of more than 10 m2. Such a VNS would complement ITER in testing, developing, and qualifying nuclear technology components, materials, and their combinations for DEMO and future commercial power plants. The effort of this study has established the potential of the spherical tokamak as a credible VNS concept that satisfies the above requirements.

Small tokamaks for fusion technology testing

Small steady-state tokamaks for testing divertors and fusion nuclear technologies are considered. Based on present physics and technology data and explanation to reduce R{sub 0}/a, H-D-fueled tokamaks with R{sub 0} {approximately} 0.6--0.75 m, R{sub 0}/a {approximately} 1.8--2.5, and B{sub t0} {approximately} 1.4--2.2 T can be driven with P{sub tot} {approximately} 4.5 MW to maintain I{sub p} {approximately} 0.5 MA and produce the ITER-level plasma edge and divertor conditions. Given an adequate steady-state divertor solution and Q{approximately}1 operation based on fusion through the suprathermal component, D-T-fueled tokamaks with R{sub 0} {approximately} 0.8 m, R{sub 0}/a {approximately} 2, and B{sub t0} {approximately} 4 T can be driven with P{sub tot} {approximately} 15 MW to maintain I{sub p} {approximately} 4.6 MA and produce an peak neutron wall load W{sub L} {approximately} 1 MW/m{sup 2}. Such devices appear possible if the plasma properties at the power R{sub 0}/a remain tokamak-like and, for the D-T case, can unshielded center core is feasible. The use of a single conductor as the inboard leg of the toroidal field coils for this purpose is discussed. The physics issues and the design features are identified for such tokamaks with a testing duty for factor goal of 10--20%.

The EPICS based virtual accelerator-concept and implementation

A virtual accelerator (VA) concept and an implementation founded on TRACE3D and PARMILA codes are presented. This virtual accelerator is suitable for accelerators with a control system based on EPICS and consists of the EPICS portable channel access server (PCAS), the EPICS client providing communication between a simulation model and PCAS, and the simulation model itself. The virtual accelerators for the SNS linac and experience in using these VAs are discussed.

Two-point transport model for cold divertor plasmas

Abstract A two-point model for plasma transport along divertor field lines is presented here. Solutions for the steady-state plasma temperature, density, and flow speed at the divertor throat and plate are found for a given throat heat flux, divertor field line length, neutral recycling coefficient, and throat density. Because of the simplicity of the model, it is useful for parameter surveys and indications of general trends. Results from the model show that more than one steady-state solution can exist for a particular throat density and that the range of throat densities for which solutions exist is limited. Also, for a constant plasma flux at the divertor throat, as neutral recycling increases, the throat density must also increase to compensate for a large decrease in plasma flow speed at the throat.

Spherical tokamak (ST) transmutation of nuclear wastes

The concept for an ST fusion core that drives a He-cooled, actinide-bearing, molten-salt blanket of moderate power density to generate electricity is examined for the first time. The results show that the fusion core is suited for this purpose and require a level of plasma, power density, engineering, and material performances moderate in comparison with what has been considered desirable for fusion-only power plants. The low aspect ratio of ST introduces a relatively thick, diverted scrape-off layer which leads to reduced heat fluxes at the limiter and divertor tiles. The use of a demountable, water-cooled, single-turn copper center leg for the toroidal field coils enables simplifications of the fusion core configuration and improves overall practicality for future power applications. These result in much reduced size and cost of the fusion core for the transmutation power plant relative to an optimized fusion-only fusion core. Surrounded by a separate tritium-breeding zone, the molten-salt blanket concept is in principle less complex and costly than the thermal breeding blankets for fusion. These combine to effect major reductions in the cost and weight of the power core equipment for the transmutation power plant. The minimum cost of electricity for such a power plant is thus reduced from the best fusion-only counterpart by more than 30%, based on consistent but approximate modeling. The key issues, development steps, and the potential value inherent in the ST fusion core in addressing the world needs for nuclear waste reduction and energy production are discussed.