Charles W. Maynard

Summary of Apollo, a D-3He tokamak reactor design

AbstractThe key features of Apollo, a conceptual D-3He tokamak reactor for commercial electricity production, are summarized. The 1000-MW(electric) design utilizes direct conversion of synchrotron radiation power and thermal conversion of transport, neutron, and bremsstrahlung radiation power. The direct conversion method uses rectennas, and the thermal conversion cycle uses an organic coolant. Apollo operates in the first-stability regime, with a major radius of 7.89 m, a peak magnetic field on the toroidal field coils of 19.3 T, a 53-MA plasma current, and a 6.7% beta value. The low neutron production of the D-3He fuel cycle greatly reduces the radiation damage rate and allows a full-lifetime first wall and structure made of standard steels with only slight modifications to reduce activation levels. The reduced radioactive inventory and afterheat give significant safety and environmental advantages over deuterium-tritium reactors.

Lattice Calculations and Three-Dimensional Effects in a Laser Fusion-Fission Reactor

Three-dimensional neutronics models of the SOLASE-H fusion-fission reactor have been analyzed by Monte Carlo. In this design, light water reactor (LWR) fertile ThO/sub 2/ fuel bundles are enriched in the fissile isotope /sup 233/U and then shipped for burning in the LWRs. A concept where the fertile fuel bundles constitute a lattice configuration with the moderator-multiplier material is investigated. Parametric lattice calculations as a function of the neutron moderator-multiplier to fuel volume ratio (v/sub m//v/sub f/) in the lattice show that it is possible in such a concept to enhance the fissile nuclei production density in the fertile fuel, compared to cases where a lattice configuration is not used.

The use and performance of graphite and metal ISSECs in tokamak fusion reactors

Abstract Carbon and the metallic elements molybdenum, niobium, vanadium and tungsten have been considered for use as an ISSEC (Internal Spectral Shifter and Energy Converter) in tokamak fusion reactors. All five materials have been shown to reduce the radiation damage in the 316 SS structural first wall and thus increase the first wall lifetime. On a per unit thickness basis, a tungsten ISSEC is most effective in this regard followed by Mo, Nb, V and carbon ISSECs in decreasing order. If the ISSEC is restricted to transfer its heat to the first structural wall by thermal radiation only, the maximum allowable thickness a carbon ISSEC can have is limited to 9.5 cm, Mo ISSEC to 7.5 cm, Nb to 8.5 cm, V to 4.5 cm and a W ISSEC to 6 cm for a 1 MW/m2 neutronic and 4 W/cm2 surface heat loading. If the ISSEC is cooled by radiation plus conduction, the maximum allowable thickness goes up to 13 cm for C, 10 cm for Mo, 8 cm for V and stays the same for Nb and W. The only ISSEC material to result in an overall reduction in total blanket radioactivity at shutdown is carbon while all of the metallic ISSECs increase the total activity. On the other hand, the long term activity (at 1000 years after shutdown) is increased for Mo, Nb and V ISSECs while it is reduced by C and W. The carbon and V ISSECs reduce the energy production per fusion while Nb and W increase it slightly and Mo results in a 15–17% energy production increase. On a relative cost basis, metallic ISSECs cost 30–55 times more than a carbon ISSEC when used at the maximum thicknesses given above. Among the four metals studied, Mo is considered to be the best material for use as an ISSEC. A definitive choice between a graphite and Mo ISSEC is difficult at this time as both materials have strong positive features; carbon being superior from radioactivity, afterheat, cost and fabricability standpoint, but molybdenum being more effective in reducing the radiation damage in the first structural wall and increasing the energy multiplication in the blanket.

COMBINED EFFECT OF ULTRAHIGH PRESSURE AND NUCLEAR IRRADIATION ON SOLID- SOLID PHASE TRANSFORMATIONS

A special Bridgman opposed diamond anvil apparatus was constructed to generate useful static pressures up to 50 kbar at ambient temperatures. It was designed to function in the environment of a nuclear reactor to thermal fluxes of 1018 nvt. The bulk of the study deals with ZrO2; other materials investigated to varying degrees were PbO2, BN, B2O3, B, silicic acid, SiO2, BaTiO3, graphite, P, Ni, Zr, and U. The samples were exposed while under pressure to fast neutron fluxes (>2.8 MeV) of 4.1×1015 nvt. The boron‐containing samples suffered up to 2×1018 fissions/cc while under pressure. PbO2, BN, B2O3, silicic acid, graphite, SiO2, and Zr were bombarded by uranium fission fragments to 6.5×1011/cm2. ZrO2 was bombarded with 2×1016 nvt epithermal neutrons and 1.6×1014 fission fragments/cm2. There was no evidence of any irradiation‐pressure‐promoted phase change. Pressure alone converted tetragonal ZrO2 to the monoclinic form and cubic ZrO2 transformed to tetragonal plus monoclinic phases. A PbO2 rutile → orthorh...

Time-dependent single-collision kernels for integral transport theory

The authors discuss time-dependent integral transport equation single-collision kernels for one-dimensional geometries corresponding to the steady-state single-collision kernels found in the available literature calculated by making use of the Laplace transform technique, simple geometric transformation relationships, and point kernel integrations. Using the convolution theorem, the time-dependent scalar flux is obtained by convoluting the single-collision kernel with the time-dependent source. Using the multiple collision formulation of the integral transport solution isotropic sources thar are delta distributions in time are considered in several examples. Analytical solutions for the uncollided and first-collided scalar fluxes are obtained for a boundary source having an isotropic angular distribution directed into a semi-infinite medium and into a slab of thickness b and for a point source at the origin of an infinite medium and finite sphere of radius a.

The Availability Analysis of Fusion Power Plants as Applied to MARS

The computer program AVSYS has been developed to analyze the availability of fusion power plants. A parametric study has been conducted on MARS. In order to bring up the availability to acceptable ...

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.

The discrete cones method for two-dimensional neutron transport calculations

Abstract A novel method, the discrete cones method (DCN), is proposed as an alternative to the discrete ordinates method (SN) for solutions of the two-dimensional neutron transport equation. The new method utilizes a new concept, discrete cones, which are made by participating a unit spherical surface that the direction vector of particles covers. In this method particles in a cone are simultaneously traced instead of those in discrete directions so that an anomaly of the SN method, the ray effects, can be eliminated. The DCN method has been formulated for X-Y geometry and a program has been created by modifying the standard SN program TWOTRAN-II. Our sample calculations demonstrate a strong mitigation of the ray effects without a computing cost penalty.

Safety and Environmental Characteristics of Recent D- 3 He and DT Tokamak Power Reactors

A comparison of the key features of the D-He Apollo and the DT ARIES fusion power reactor designs is made. The reduction in neutron production from the D-He reaction has a major effect on the performance of tokamak reactors. One of the biggest impacts is the low radiation damage rate in D-He systems which allows a permanent first wall to be utilized. The reduction in radioactivity in D-3He reactors has a particularly advantageous effect on the storage of wastes as well as on the safety to the public in the event of the worst conceivable accident. The more difficult D-He physics requirements are offset by the technological advantages of using this fuel in place of the DT cycle.

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.