Hiroyasu Utoh

Critical design factors for sector transport maintenance in DEMO

This paper mainly focuses on a sector transport maintenance scheme from the aspects of high plant availability. In this study, three different maintenance schemes are considered based on (1) the number of maintenance ports and (2) the insertion direction. The design study clarifies critical design factors and key engineering issues on the maintenance scheme: (1) how to support an enormous overturning force of the toroidal field coils in the large open port for sector transport and (2) define the transferring mechanism of sectors in the vacuum vessel. On reviewing these assessment factors, the sector transport using a limited number of horizontal maintenance ports is found to be a more realistic maintenance scheme. In addition, evaluating maintenance scenarios under high decay heat is proposed for the first time. The key design factors are the cool-down time in the reactor and the cooling method in the maintenance scheme to keep components under operational temperature. Based on one-dimensional heat conduction analysis, after one month cool-down time, each sector of SlimCS could be transported to the hot cell facility by gas cooling.

Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. As for the in-VV LOCA, we analysed the multiple double-ended break of the first wall cooling pipes around the outboard toroidal circumference. As for the ex-VV LOCA, we analysed the double-ended break of the primary cooling pipe. The thermohydraulic analysis results suggest that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. Mitigations of the loads to the confinement barriers are also discussed.

Management Strategy for Radioactive Waste in the Fusion DEMO Reactor

We have considered a strategy for reducing the radioactive waste generated by the replacement of in-vessel components, such as blanket segments and divertor cassettes, for the fusion DEMO reactor. In the basic case, the main parameters of the DEMO reactor are a major radius of 8.2 m and a fusion power of 1.35 GW. Blanket segments and divertor cassettes should be replaced independently, as their lifetimes differ. A blanket segment comprises several blanket modules mounted to a back-plate. The total weight of an in-vessel component is estimated to be about 6,648 ton (1,575, 3,777, 372, and 924 ton of blanket module, back-plate, conducting shell, and divertor cassette, respectively). The lifetimes of a blanket segment and a divertor cassette are assumed to be 2.2 years and 0.6 years, respectively, and 52,487 tons of waste is generated over a plant life of 20 years. Therefore, there is a concern that the contamination-control area for radioactive waste may need to increase due to the amount of waste generated from every replacement. This paper proposes a management scenario to reduce radioactive waste. When feasible and relevant, back-plates of blanket segment and divertor cassette bodies (628 ton) should be reused. Using the three-dimensional neutron transportation code MCNP, the displacement per atom (DPA) of the SUS316LN back-plates is 0.2 DPA/year and that of the F82H cassette bodies is 0.6 DPA/year. Therefore, the reuse of back-plates and cassette bodies would be possible if re-welding points are arranged under neutron shielding. We found that radioactive waste could be reduced to 20 % when tritium breeding materials are recycled. Finally, we propose a design for the DEMO building that uses a hot cell and temporary storage.

Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

Safety studies of a water-cooled fusion DEMO reactor have been performed. In the DEMO design, the blanket primary cooling system involves a large amount of energy due to pressurized water coolant (290-325 °C, 15.5 MPa). Moreover, it contains radioactive materials such as tritium and activated corrosion products. Therefore, in the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three options of confinement strategies. In each option, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to the environment were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.

Progress in thermohydraulic analysis of accident scenarios of a water-cooled fusion DEMO reactor

Thermohydraulic analysis of postulated accidents will identify systems responses to accident scenarios and aid to develop design of safety systems and strategies to prevent and mitigate accident propagation. This paper reports recent progress in analysis of some accident scenarios of a water-cooled fusion DEMO reactor. Four types of accidents are particularly considered: ex-vessel loss-of-coolant of the primary cooling system, in-vessel loss-of-coolant of the first wall cooling pipes, loss-of-coolant in blanket modules and loss-of-vacuum. The analysis has identified responses of the DEMO systems to these accidents, pressure loads to confinement barriers for radioactive materials and radiological consequences. Effectiveness of the safety systems and integrity of the confinement barriers against the accidents are discussed.

Conceptual design study of the moderate size superconducting spherical tokamak power plant

A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW−1 h−1 (

Simplification of blanket system for SlimCS fusion DEMO reactor

2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.