Fusion reactivities and neutron source characteristics of beam-driven toroidal reactors with both D and T injection

The reactor performance is considered for intensely beam-driven tokamak plasmas with 50:50 D-T composition maintained by neutral-beam injection of both D and T, together with plasma recycling. The D and T are injected with equal intensity and velocity. This mode of operation is most appropriate for high-duty- factor, high-power-density operation, in the absence of pellet injection. The isotropic velocity distributions of energetic D and T ions (for multi-angle injection) are calculated from a simple slowing-down model, but include a tail above the injection velocity. The neutron source characteristics are determined from fusion reactivities calculated for beam-target, hot-ion, and thermonuclear reactions. For conditions where Q approximates 1, beam-target reactions are dominant, although reactions among the hot ions contribute substantially to P/sub fusion/ when n/sub hot//n /sub e/ greater than or equal to 0.2. (auth

Tokamak engineering test reactor

The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m$sup 2$ is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m$sup 2$. The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10$sup 12$ cm$sup -3$s. The plasma temperature is maintained by injection of 177 MW of 200- keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10$sup -6$. If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m$sup 2$ can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth

Counterstreaming-ion tokamak neutron source for large area surface radiation studies

A tokamak neutron source that produces a neutron flux approximately 10$sup 13$ n/cm$sup 2$/s over a wall test-area of 30 m$sup 2$ is designed using near state-of-the-art tokamak and neutral-beam injection technologies. To maximize fusion reactivity, D and T plasma ions are grouped in two distinct quasi- thermal velocity distributions, oppositely displaced in velocity along the magnetic axis. Such counterstreaming distributions can be set up by tangential injection of all plasma ions by oppositely directed D and T neutral beams, by facilitating removal of completely decelerated ions, and by minimizing plasma recycling. Fusion energy is produced principally by head-on collisions between D and T ions in the counterstreaming distributions. For injection energies of 40- 60 keV, and typical tokamak parameters, the fusion power density can be approximately 1 W/cm$sup 3$, with Q approximately 1 attainable for T/sub e/ = 3.4 keV and electron energy confinement parameter n/sub e/tau/sub E/ approximately equal to 2 x 10$sup 12$ cm$sup -3$s. All plasma fueling is carried out by the injected beams, and when a significant fraction of the electron population is trapped, the plasma current can be maintained by the beams. (auth

Electrical energy requirements for ATW and fusion neutrons

This note compares the electrical energy requirements of accelerator (ATW) and fusion plants designed to transmute nuclides of fission wastes. Both systems use the same blanket concept but for each source neutron the fusion system must utilize one blanket neutron for tritium breeding. The ATW and fusion plants are found to have the same electrical energy requirement per available blanket neutron when the blanket coverage is comparable and fusion Q {approx} 1, but the fusion plant has only a fraction of the energy requirement when Q {much{underscore}gt} 1. If the blanket thermal energy is converted to electricity, the fusion plant and ATW have comparable net electrical energy outputs per available neutron when Q {>=} 2

Maximum neutron wall loadings in beam-driven tokamak reactors

If a beam-driven D--T tokamak reactor is operated at the maximum density allowed both by pressure limitation and by adequate neutral-beam penetration, the 14-MeV neutron wall loading increases approximately linearly with magnetic field or vertical elongation of the plasma. With elongation = 3, B/sub tmax/ equals 15T, W/sub beam/ = 200 keV, Q approximately 1.0, maximum wall loading is about 5 MW/m$sup 2$. (auth

Energy spectra of fusion neutrons from plasmas driven by reacting ion beams

Neutron spectra are calculated for two-component tokamak reactors using appropriate steady-state velocity distributions for fast deuterons. The angle- averaged neutron spectra are shown for D-T and D-T interactions. (MOW

Prospects for thermonuclear ignition in a ''collisional'' tokamak

The parameters are described for a tokamak reactor plasma that attains ignition in the same regime of collisionality as present-day ohmic-heated tokamak plasmas, where the confinement scaling ntau$alpha$n$sup 2$ is observed. The use of Nb$sub 3$Sn toroidal field coils and a plasma elongation greater than or equal to 1.5 are necessary to attain the high plasma density (n approximately 10$sup 15$ cm$sup -3$) required for ignition in this collisional regime. Under these conditions, the fusion power density is of order 10 W/cm$sup 3$. This high value is probably necessary for an economic tokamak reactor. (auth

Design considerations for neutron activation and neutron source strength monitors for ITER

The International Thermonuclear Experimental Reactor will require highly accurate measurements of fusion power production in time, space, and energy. Spectrometers in the neutron camera could do it all, but experience has taught us that multiple methods with redundancy and complementary uncertainties are needed. Previously, conceptual designs have been presented for time-integrated neutron activation and time-dependent neutron source strength monitors, both of which will be important parts of the integrated suite of neutron diagnostics for this purpose. The primary goals of the neutron activation system are: to maintain a robust relative measure of fusion energy production with stability and wide dynamic range; to enable an accurate absolute calibration of fusion power using neutronic techniques as successfully demonstrated on JET and TFTR; and to provide a flexible system for materials testing. The greatest difficulty is that the irradiation locations need to be close to plasma with a wide field of view. The routing of the pneumatic system is difficult because of minimum radius of curvature requirements and because of the careful need for containment of the tritium and activated air. The neutron source strength system needs to provide real-time source strength vs. time with {approximately}1 ms resolution and wide dynamic range in a robust and reliable manner with the capability to be absolutely calibrated by in-situ neutron sources as done on TFTR, JT-60U, and JET. In this paper a more detailed look at the expected neutron flux field around ITER is folded into a more complete design of the fission chamber system