B. Misra

The trio experiment

The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an analytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion.

The TRI0-01 experiment: In-situ tritium recovery results

The TRIO-01 experiment is a test of in-situ tritium recovery from ..gamma..-LiAlO/sub 2/ with test conditions chosen to simulate those anticipated in fusion power reactors. A status report is presented which describes qualitatively the results observed during the irradiation phase of the experiment. Both the rate of tritium release and the chemical forms of tritium were measured using a helium sweep gas which flowed past the breeder material to a gas analysis system.

STARFIRE, a commercial tokamak power plant design

Abstract STARFIRE is a design for a conceptual commercial tokamak electrical power plant based on the deuterium/tritium/lithium fuel cycle. In addition to the goal of being technologically credible, the design incorporates safety and environmental considerations. STARFIRE is considered to be the tenth in a series of commercial fusion power plants. STARFIRE has a 7-m major radius reactor producing 1200 MW of net electrical power from 4000 MW of thermal power, with an average neutron wall load of 3.6 MW/m 2 . The aspect ratio is 3.6 and a D-shaped plasma with a height-to-width ratio of 1.6 and average toroidal beta of 0.067 is used. The maximum magnetic field is 11T. Availability goals have been set at 85% for the reactor and 75% for the complete plant including the reactor. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum for impurity control, most superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield.

Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles.

Wildcat: A commercial deuterium-deuterium tokamak reactor

AbstractWILDCAT is a conceptual design of a catalyzed deuterium-deuterium tokamak commercial fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing deuterium-tritium (D-T) designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete conceptual design.

A COMPARISON OF TOKAMAK BURN CYCLE OPTIONS

Experimental confirmation of noninductive current drive has spawned a number of suggestions as to how this technique can be used to extend the fusion burn period and improve the reactor prospects of tokamaks. Several distinct burn cycles, which employ various combinations of Ohmic and noninductive current generation, are possible, and we will study their relative costs and benefits for both a commercial reactor as well as an INTOR-class device. We begin with a review of the burn cycle options.

Materials selection for the U.S. INTOR divertor collector plate

Abstract The divertor collector plate in the INTOR reactor will be subjected to high heat, particle, and neutron fluxes, making it the most severely damaged torus component. The collector plate is composed of a protection plate, which is directly exposed to the particle flux, and a heat sink which provides support for the protection plate and carries the water coolant. The high-Z refractory metals have been considered for use as the protection plate material and austenitic stainless steels and copper alloys have been considered as the heat sink material. Tungsten and Type 316 stainless steels have been selected for the protection plate and heat sink, respectively. The protection plate has a sputtering lifetime of 1.75 y at a 50% duty factor, while the heat sink is expected to last the lifetime of the reactor.

Comparative study of pulsed and steady-state tokamak reactor burn cycles

Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles.

Evaluation of materials for EPR power generation

Abstract The blanket materials employed for heat generation in the Argonne Experimental Power Reactor (EPR) are evaluat, ed. The EPR blanket consists of annealed Type 316 stainless steel sections cooled by pressurized water and Inconel 718 sections cooled by steam. The predicted lifetimes of the two different blanket sections is approximately 2 years of normal operation for the Inconel 718 sections and approximately 3.5 years for Type 316 stainless steel sections. The lifetime of annealed Type 316 stainless steel is limited by swelling considerations, while the lifetime of Inconel 718 is limited by ductility considerations.