Michinori Yamauchi

Basic Requirements for a 1000-MW(electric) Class Tokamak Fusion-Fission Hybrid Reactor and Its Blanket Concept

AbstractPlasma size and other basic performance parameters for 1000-MW (electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is based on the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. Thus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW (electric) hybrids. Typical plasma parameters for 1000-MW (electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of Ip = 15 MA, and toroidal field on the axis of B0 = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-produ...

Analysis and experimental results on ingress of coolant event in vacuum vessel

Experiments on the ingress of coolant event (ICE) in the vacuum vessel of a fusion reactor have been carried out in the Japan Atomic Energy Research Institute (JAERI) as one of the research and development tasks for the International Thermonuclear Experimental Reactor (ITER) to obtain the thermofluid data for validation of safety analysis codes. The ICE experiment is numerically analyzed using the transient reactor analysis code (TRAC) which is one of the codes preparing for the safety analysis of the ITER. The TRAC has been modified so as to analyze the ICE phenomena in the vacuum vessel of a fusion reactor. Several ICE experiments have been carried out as benchmark tests for the safety analysis codes. We have analyzed those experiments by using the TRAC, and considered the difference between the analysis and experimental results. Analysis results of the temperature in the vacuum vessel show a tendency completely different from the experimental result. It is clarified that the present TRAC has not been verified on the scattering behavior of water droplets.

Proposal of Integrated Test Facility for in-Vessel Thermofluid Safety of Fusion Reactors

A vacuum vessel (VV) of a tokamak fusion reactor like the International Thermonuclear Experimental Reactor (ITER) consists the first confinement barrier that includes the largest amount of radioactive materials such as tritium and activation products. The ingress of coolant event (ICE) is a design basis event in the ITER where water is used as the coolant. The loss of vacuum event (LOVA) is also considered as an independent design basis event. Based on the results of ICE and LOVA preliminary experiments, an integrated in-vessel thermofluid test is being planned and conceptual design of the facility is in progress. The main objectives of the integrated test are to investigate the consequences of possible interaction of the ICE and the LOVA and to validate the analytical model of thermofluid events in the VV of the fusion reactor. This paper introduces a conceptual design of the integrated test facility and a testing plan.

Radiation Streaming Calculations for Intor-J

The effects of radiation streaming through the neutral beam injector (NBI) port and divertor throat of a tokamak fusion reactor, the INTOR-J, was evaluated using Monte Carlo and discrete ordinates methods. Radiation streaming through the NBI port is found to be tolerable when a thick drift tube support acts as an effective shield. Neutron streaming through the divertor throat, however, makes the shutdown dose too high for personnel access into the reactor room. The radiation levels in the reactor room resulting from leakage through the NBI room walls are far smaller than that from leakage through the bulk shield, except behind the NBI room. The Monte Carlo-Monte Carlo and discrete ordinatesMonte Carlo coupling techniques used in the present study are shown to be very effective for the radiation streaming calculations.

Parametric study on blanket neutronics and economics of fusion—fission hybrid reactors

Abstract A parametric study has been performed for the blanket neutronics and economics of fusion—fission hybrid reactors. Optimum blanket designs were sought for both the electricity-producing hybrid reactor and the fissile fuel-producing hybrid reactor. Comparisons of the nuclear performances and economics between these two hybrid reactors were also conducted. A tokamak-type fusion reactor was selected as the base reactor. It was assumed that natural uranium oxide is fed to the blanket as the fertile material with lithium oxide as the tritium breeding material. In the neutronics design of the electricity-producing hybrid reactor, blanket fuel composition and arrangement were surveyed to attain high energy multiplication in the blanket. Moreover, efforts were made to suppress the blanket thermal power increase due to plutonium buildup, because it determines the frequency of blanket fuel exchange and limits the reactor load factor. On the other hand, the blanket for the fissile fuel-producing hybrid reactor was designed so that the amount of the fissile fuel bred per unit blanket thermal power is maximized. Economic analysis shows that the electricity from the symbiotic system composed of the fissile fuel-producing hybrid reactor and light water reactors costs about 30% less than that from the electricity-producing hybrid reactor. However, the large energy multiplication in the blanket of the electricity-producing hybrid reactor makes the neutron wall loading 40% lower than that of the fissile fuel-producing hybrid reactor. Therefore, the electricity-producing hybrid reactor can make its requirement for fusion technology less demanding than the other .