W.F. Vogelsang

HIBALL—A conceptual design study of a heavy-ion driven inertial confinement fusion power plant

Abstract A preliminary concept for a heavy-ion beam driven inertial confinement fusion power plant is presented. The high repetition rate of the RF accelerator driver is utilized to serve four reactor chambers alternatingly. In the chambers a novel first-wall protection scheme is used. At a target gain of 83 the total net electrical output is 3.8 GW. The recirculating power fraction is below 15%. The main goal of the comprehensive HIBALL study (which is continuing) is to demonstrate the compatibility of the design of the driver, the target and the reactor chambers. Though preliminary, the present design is essentially self-consistent. Tentative cost estimates are given. The costs compare well with those found in similar studies on magnetic confinement fusion reactors.

The INPORT concept — an improved method to protect ICF reactor first walls

Abstract A method to protect the first metallic walls of ICF reactors from X-rays and target debris has been developed. The concept utilizes porous, flexible tubes of woven C or SiC fibers to contain liquid metals inside the vacuum chamber of an ICF system. These porous tubes allow for ablation and recondensation of liquid metal films. The tubes also moderate the neutron spectra and reduce the displacement and transmutation damage in metallic walls.

Aspects of tritium release from neutron-irradiated lithium oxide☆

Out-of-reactor tritium release from well-characterized solidified melt particles of lithium oxide is investigated for dependence on temperature, sweep gas flow rate, effect of hydrogen addition to the processing gas, and particle size. Solid-phase diffusion, characterized by diffusion coefficient in D(cm2/s) = −(13.2 ± 3.7) − (27.6 ± 12 kJ/mol)/RT, is shown to control long-term escape for temperatures 673 K – 923 K. Admixture of 2.4 vol. % H2 to the sweep stream appears to promote initial tritium release.

Tritium and fissile fuel exchange between hybrids, fission power reactors, and tritium production reactors

A mathematical model extending work by Gordon and Harms is developed to describe the fissile fuel and tritium flows in a fusion-fission system consisting of a fusion hybrid reactor, a tritium production reactor, and several fission power reactors. The hybrid reactor plays the role of a fuel factory, providing the fission reactors and the tritium production reactor with their fissile fuel needs. The tritium production reactor ( a fission reactor) is devoted primarily to producing tritium for subsequent use in the hybrid. Different combinations of these systems are found by shifting the tritium breeding function among the various parts. At steady state, the total thermal power in fission reactors per unit of fusion power depends only on the total conversion ratio of the fission reactors and the hybrid. An economic analysis is required to determine which combination of systems will produce electricity at the lowest costs.

Nucleonic Design for a Compact Tokamak Fusion Reactor Blanket and Shield

The nucleonic design features of the NUWMAK are as follows. A tritium breeding ratio of 1.54 is obtained. Li/sub 62/Pb/sub 38/ eutectic is used as the breeding and thermal energy storage material. The total nuclear heating in the blanket and shield is approx. 17.2 MeV per deuterium-tritium neutron. The performance of the superconducting magnet will be satisfactory for more than 2 yr of continuous operation through the use of a 35-cm-thick tungsten shield that extends 2.5 m above the midplane on the inboard part of the torus. The radioactiity is lowered by using a titanium alloy as the structural material and large amounts of lithium lead as the blanket material. One day after shutdown, the dose rate outside the outer shield drops below 2.6 mrem/h, and it is favorable to hands-on shift maintenance.

Activation product transport in fusion reactors

Abstract Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. A computer code, RAPTOR, has been developed to determine the transport of these products in fusion reactor coolant/tritium breeding materials. Without special treatment, it is likely that fusion reactor power plant operators could experience dose rates as high as 8 rem per hour around a number of plant components after only a few years of operation.

Fusion-fission hybrid design with analysis of direct enrichment and non-proliferation features (the SOLASE-H study)

Abstract The role of a fusion-fission hybrid in the context of a nuclear economy with and without reprocessing is examined. An inertial confinement fusion driver is assumed and a consistent set of reactor parameters are developed. The form of the driver is not critical, however, to the general concepts. The use of the hybrid as a fuel factory within a secured fuel production and reprocessing center is considered. Either the hybrid or a low power fission reactor can be used to mildly irradiate fuel prior to shipment to offsite reactors thereby rendering the fuel resistant to diversion. A simplified economic analysis indicates a hybrid providing fuel to 10 fission reactors of equal thermal power is insensitive to the recirculating power fraction provided reprocessing is permitted. If reprocessing is not allowed, the hybrid can be used to directly enrich light water reactor fuel bundles fabricated initially from fertile fuel (either ThO 2 or 238 UO 2 ). A detailed neutronic analysis indicates such direct enrichment is feasible but the support ratio for 233 U or 239 Pu production is only 2, making such an approach highly sensitive to the hybrid cost. The hybrid would have to produce considerable net power for economic feasibility in this case. Inertial confinement fusion performance requirements for hybrid application are also examined and an integrated design, SOLASE-H, is described based upon the direct enrichment concept.