A broad review is presented of the status and outlook of achieving fusion energy production by the confinement of high temperature plasmas with magnetic fields. The paper describes the basic features of the fusion process, plasma confinement principles and concepts with emphasis on the Tokamak, issues of materials development, descriptions of the major components and subsystems of fusion reactors, prospects for power plants, the policy in the United States of magnetic fusion energy and the international aspects of this endeavor.

This report discusses the US ITER research and development program on the following topics: Magnets; Neutral Beam; PF Magnets; Blanket/Shield; Containment Structures; Remote Handling and Maintenance; Fueling; Vacuum Technology; Plasma Facing Components; Impurity Control, Divertor, First Wall, and Disruptions; RF Heating and Current Drive (ICH); RF Heating and Current Drive (ECH); RF Heating and Current Drive (LH); Fuel Cycle; Safety and Environment; Diagnostics; Plasma Physics; and The US Physics R&D Plans Relevant to ITER.

A solid-breeder, water-cooled blanket option based on a multilayer configuration was developed for the ITER (International Thermonuclear Experimental Reactor). The blanket uses beryllium for neutron multiplication and lithium oxide for tritium breeding. The material forms are sintered products for both materials with 0.8 density factor. The lithium-6 enrichment is 90%. The blanket can accommodate a factor of two change in the neutron wall loading without violating the different design guidelines. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. At the same time, the reliability and the safety aspects of the blanket are enhanced by using a low-pressure coolant and separating the tritium purge lines from the coolant system. The blanket modules are made by hot vacuum forming and diffusion bonding a double-wall structure with integral cooling channels. The different aspects of the blanket design including tritium breeding, nuclear heat deposition, activation analyses, thermal-hydraulics, tritium inventory, structural analyses, and water-coolant conditions are summarized.< >

A water-cooled solid-breeder blanket concept was developed for ITER. The main function of this blanket is to produce the necessary tritium for the ITER operation. Several design features are incorporated in this blanket concept to increase its attractiveness. The main features are the following: a) a multilayer concept which reduces fabrication cost; b) a simple blanket configuration which results in reliability advantages; c) a very small breeder volume is employed to reduce the tritium inventory and the blanket cost; d) a high tritium breeding ratio eliminates the need for an outside tritium supply; e) a low-pressure system decreases the required steel fraction for structural purposes; f) a low-temperature operation reduces the swelling concerns for beryllium; and g) the small fractions of structure and breeder materials used in the blanket reduce the decay heat source. It is assumed that the blanket operation at commercial power reactor conditions can be sacrificed to achieve a high tritium breeding ratio with minimum additional research and development, and minimal impact on reactor design and operation.Operating temperature limits are enforced for each material to insure a satisfactory blanket performance. In fact, the design was iterated to maximize the tritium breeding ratio and satisfy these temperature limits. The other design constraint is to permit a large increase in the neutron wall loading without exceeding the temperature limits for the different blanket materials. The blanket concept contains 1.8 cm of Li2O and 22.5 cm of beryllium both with a 0.8 density factor. The water coolant is isolated from the breeder material by several zones which reduces the tritium buildup in the water by permeation, reduces the chance for water-breeder interaction, and permits the breeder to operate at high temperature with a low temperature coolant. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. The key features and design analyses of this blanket are summarized in this paper. Notesa Work supported by the U.S. Department of Energy, Office of Fusion Energy.

This blanket concept uses a dilute suspension of fine solid breeder particles (Li2O, LiAlO2, or Li4SiO4) in a carrier gas (He) as the coolant and the tritium breeding stream. A small fraction of this stream is processed outside the reactor for tritium recovery. The blanket consists of a beryllium multiplier and carbon/steel reflector. A steel clad is used for all materials. A carbon reflector is employed to reduce the beryllium thickness used in the blanket for a specific tritium breeding ratio. The breeder particle size has to exceed a few microns (≥ 2 microns) to avoid sticking problems on the cold surfaces of the heat exchanger. The helium gas pressure is in the range of 2–3 MPa to carry the solid breeder particles through the blanket and the heat exchanger loop. The solid breeder concentration in the helium stream is 1 to 5 volume percent. A high lithium-6 enrichment is used to produce a high tritium breeding ratio and to reduce the breeder concentration in the helium gas. At a lithium-6 enrichment of 90%, the local tritium breeding ratio is 2.03 based on a one-dimensional poloidal model. The total thickness of the helium stream is only 4 cm out of the 50 cm total blanket thickness. The blanket uses 35 cm of beryllium for neutron multiplication. A simple multi-layer design is employed where the blanket sector has the helium coolant flowing in the poloidal direction. The blanket concept has several unique advantages which are very beneficial for fusion reactors including ITER. The key advantages are listed below:The blanket operation can be switched between nonbreeding and breeding modes without hardware changes in the reactor.The blanket performance can be adjusted during reactor operation by changing the breeder concentration, lithium-6 enrichment, helium pressure, or helium velocity.The blanket has a very low tritium inventory.The tritium breeding ratio is adjustable during operation in the range of 0 to 2.03.The coolant loop operates at low to medium pressure (2 to 3 MPa).The addition of µ-sized solid particles in the helium gas improves its heat transfer and transport properties.The blanket can be designed to operate in any temperature range suitable for an optimum structure performance.The afterheat source in the blanket is very low because of the low steel fraction and the absence of the solid breeder.The blanket has no impact on the reactor configuration, and it can be designed for vertical or horizontal maintenance schemes.The blanket has very low design uncertainties in its performance.The blanket concept has the potential to extrapolate to power reactor conditions.The main features, key technical issues, and design analyses of this blanket concept are summarized in this paper.

Critical technical issues that affect the feasibility of the International Tokamak Reactor (INTOR) concept for a next-generation tokamak experimental reactor are analyzed. The results of these studies are used to update the recommended design concept for INTOR.

The goal of our study was to provide a limited analytical assessment of the technical and economic benefits associated with the use of polarized fuels in a commercial deuterium-tritium (D-T) tokamak fusion reactor. The effort was focused on determining key benefits accrued in the reactor systems. The balance of plant (BOP) was assumed to not be appreciably affected. The D-T system was chosen since it is likely to be the first commercial reactor system. Two types of reactors were considered: one was the Starfire reactor, which has a moderate beta, a moderate radius, and a moderate magnetic field; the other was an advanced reactor developed in Argonne National Laboratory's Tokamak Power Systems Study, which has a higher beta a smaller radius, and a lower magnetic field. Since the physical size, i.e., major radius of a reactor, significantly affects the cost of the toroidal field magnets, these two systems should provide a means of separating the size effect from effects attributable to the use of polarized fuels. A number of alternative cases were considered for the two reference cases. Our conclusions were that the use of polarized fuels in a commercial reactor can increase its performance. However, the sum of allmore » benefits does not result in a significant improvement in reactor economics. This is especially true in situations where polarization is not 100%. The benefits of polarized fuels are quickly lost if polarization is < 95%. Therefore, the primary value of polarized fuels is as an option that can be used in a given fusion reactor design if a required condition cannot be achieved in other, more standard operating modes.« less

A number of advances in plasma physics and engineering promise to greatly improve the reactor prospects of tokamaks. The following features, in particular, have been examined: (a) large aspect ratio (A ≍ 6), which may ease maintenance; (b) high beta (β ≳ 0.20) without indentation, which brings the maximum toroidal field down to about 7 T; (c) low toroidal current (I ≍ 5MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields produced in the plasma. In addition to matching desirable high-beta equilibria, this method is capable of producing a large variety of new equilibria, many of which look attractive. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or “slot” limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors.

The benefits and limitations arising from the use of polarized deuterium-tritium fuels were assessed for commercial tokamak fusion reactors. The difference in capital costs for a reactor with and without polarized fueling was quantified for two reactors, one with a beta of 0.067 and a major radius of 7 m and the second with a beta of 0.25 and a major radius of 5.25 m. The change in reactor performance was also quantified. The conclusion was that the sum of all benefits associated with the use of polarized fuels does not result in a significant improvement in the tokamak reactor economics.

The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

The US Department of Energy's Office of Fusion Energy has initiated several studies during FY-1985 called Tokamak Power System Studies (TPSS). The TPSS is being carried out by several laboratories, universities and industry with the general objective of developing innovative physics and technology concepts to improve the commercial attractiveness of tokamak power reactors. The effort of Argonne National Laboratory, entitled STARFIRE-II, is an effort to update and improve STARFIRE, which was the last comprehensive conceptual design study in the US of a commercial tokamak power plant. The STARFIRE-II effort has developed a number of goals in order to improve fusion commercial power plants based in part on several recent studies. The primary goals for STARFIRE-II are listed.