John Sheffield

A plan for the development of fusion energy

This is the final report of a panel set up by the U.S. Department of Energy (DOE) Fusion Energy Sciences Advisory Committee (FESAC) in response to a charge letter dated September 10, 2002 from Dr. Ray Orbach, Director of the DOEs Office of Science. In that letter, Dr. Orbach asked FESAC to develop a plan with the end goal of the start of operation of a demonstration power plant in approximately 35 years. This report, submitted March 5, 2003, presents such a plan, leading to commercial application of fusion energy by mid-century. The plan is derived from the necessary features of a demonstration fusion power plant and from the time scale defined by President Bush. It identifies critical milestones, key decision points, needed major facilities and required budgets. The report also responds to a request from DOE to FESAC to describe what new or upgraded fusion facilities will “best serve our purposes” over a time frame of the next twenty years.

A Study of Options for the Deployment of Large Fusion Power Plants

One option for making fusion power plants that could be competitive with other power plants operating during the 21st century is to make them large, e.g., 3 GW(electric) or more, to take advantage of the expected economies of scale. This study examines the effects on electrical utility system hardware, operations, and reliability of incorporating such large generating units. In addition, the study evaluates the use of the coproduction of hydrogen to reduce the grid-supplied electricity and offer the possibility for electrical load-following. The estimated additional cost of electricity (COE) for a large power plant is ~5 mills/kW⋅h. The estimated total COE for 3- to 4-GW(electric) fusion power plants lies in the range of 37 to 60 mills/kW⋅h. Future hydrogen costs from a variety of sources are estimated to lie in the range of 8 to 10

The role of energy efficiency and renewable energies in the future world energy market

/GJ, when allowance is made for some increase in natural gas price and for the possible need for greenhouse gas emission limitations. A number of combinations of fusion plant and electrolyzer were considered, including hot electrolyzers that use heat from the fusion plant. For the optimum cases, hydrogen produced from off-peak power from a 3- to 4-GW(electric) plant is estimated to have a competitive cost. Of particular interest, the cost would also be competitive if some hydrogen were produced during on-peak electricity cost periods. Thus, for a 4-GW(electric) plant, only up to 3 GW(electric) might be supplied to the grid, and load-following would be possible, which would be a benefit to the utility system.

Impact of tritium removal and He-3 recycling on structure damage parameters in a D-D fusion system

The world population is rising rapidly, notably in the developing countries. Historical trends suggest that increased annual energy use per capita is a good surrogate for the standard of living factors which promote a decrease in population growth rate. If these trends continue, the stabilization of the worlds population will require the increased use of all sources of energy as cheap oil and gas are depleted. The improved efficiency of energy use and renewable energy sources will be essential to stabilizing population, while providing a decent standard all over the world.

Physics Guidelines for the Compact Ignition Tokamak

Abstract Removing tritium produced by D–D fusion and recycling part of it after it decays to He-3 significantly reduces the fraction of fusion energy carried by neutrons in a D–D system. For a catalyzed D–D system (no tritium removal), the peak dpa rate in candidate structural materials is 25–35% lower than that in an equivalent D–T system with the same fusion power wall loading. The gas production and transmutation rates are about 60% lower. As tritium is removed gas production and transmutations decrease by more than two orders of magnitude and the dpa rate decreases by a factor of 2.3–2.8. An additional reduction of a factor of 1.6–1.7 in damage parameters is achieved by recycling the removed tritium as He-3. This results in significant lifetime enhancement of structural materials. Information from tests in fission reactor spectra would be directly relevant in determining the lifetime of the structural material in this He-3-recycled system.

Deuterium-Fueled Power Plants with Tritium Suppression

The goal of the Compact Ignition Tokamak (CIT)d program is to provide a cost-effective route to the production of a burning deuterium-tritium plasma, so that alpha-particle effects may be studied. A key issue to be studied in the CIT is whether alpha power behaves like other power sources in affecting tokamak plasma confinement. The program is managed by the Princeton Physics Laboratory and includes broad community involvement. Guidelines for the preliminary design effort have been provided by the Ignition Technical Oversight Committee in discussion with the tokamak community. The reference design is a tokamak with a high filed (10 T), high current (10 MA), poloidal divertor, and liquid-nitrogen-cooled coils. It is a small, high-power-density device of the type proposed by Bruno Coppi (MIT). It has a major radius of 1.23 m, a minor radius of 0.43 m, and plasma elipticity of 1.8. This paper reviews the aims of the program and the basis for the physics guidelines. The role of the CIT in the longer-term tokamak program is briefly discussed. 23 refs., 9 figs., 1 tab.

Generic Magnetic Fusion Reactor Revisited

Abstract Catalyzed D-D is the ultimate fusion cycle, because deuterium is essentially unlimited on earth. In this approach, the 3He and tritium fusion products are recycled to increase the charged particle fusion power. A difficulty with this fusion cycle is that the tritium from fusion, if left in the plasma, produces 14-MeV neutrons, leading to radiation damage comparable to that of the D-T cycle. This paper shows that the damage problems may be alleviated by removing tritium before it can burn. Fortunately, the charged particle fusion power from burning the tritium is small compared to that from the 3He and removing it from the plasma makes little difference to the plasma power balance. Ion cyclotron power might be used to pump out tritium. In this paper, we review the benefits of tritium removal, identify the issues associated with this approach, and determine illustrative parameters required for an advanced tokamak and an advanced stellarator.

Magnetic fusion commercial power plants

Abstract The original generic magnetic fusion reactor paper was published in 1986 for deuterium-tritium reactors. This update describes what has changed in 30 years. Notably, the construction of ITER is providing important benchmark numbers for technologies and costs. In addition, we use a more conservative neutron wall flux and fluence. But, these cost-increasing factors are offset by greater optimism on the thermal-electric conversion efficiency and potential availability. In addition, today’s inflation and interest rates are low, leading to a cost of money well below that used in the original study. The main examples show the cost of electricity (COE) as a function of aspect ratio and neutron flux to the first wall. The dependence of the COE on availability, thermoelectric efficiency, electrical power output, and the present day’s low interest rates is also discussed. Interestingly, at fixed aspect ratio there is a shallow minimum in the COE at neutron flux of ~2.5 MW/m2. The possibility of operating with only a small COE penalty at even lower wall loadings (to 1.0 MW/m2 at larger plant size) and the possible use of niobium-titanium coils are also investigated. It should be emphasized that the variation in the COEs is important rather than their absolute values.

Fusion Energy Advisory Committee (FEAC): Panel 7 report on Inertial Fusion Energy

Toroidal magnetic systems offer the best opportunity to make a commercial fusion power plant. They have, between them, all the features needed; however, no one system yet meets the ideal requirements. The tokamak is the most advanced system, and the proposed International Thermonuclear Experimental Reactor (ITER) and Tokamak Physics Experiment (TPX) will build upon the existing program to prepare for an advanced tokamak demonstration plant. Complementary toroidal systems such as the spherical torus, stellarator, reversed-field pinch, field-reversed configuration, and spheromak offer, between them, potential advantages in each area and should be studied in a balanced fusion development program.

Prospects for toroidal fusion reactors

The charge to FEAC Panel 7 on inertial fusion energy (IFE) is encompassed in the four articles of correspondence. To briefly summarize, the scope of the panel`s review and analysis adhered to the following guidelines. (1) Consistent with previous recommendations by the Fusion Policy Advisory Committee (FPAC) and the National Academy of Science (NAS) panel on inertial fusion, the principal focus of FEAC Panel 7`s review and planning activities for next-generation experimental facilities in IFE was limited to heavy ions. (2) The panel considered the three budget cases: