Ken Tomabechi

Study of a compact reversed shear Tokamak reactor

Abstract A reversed shear configuration, which was observed recently in some Tokamak experiments, might have a possibility to realize compact and cost-competitive Tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the Compact Reversed Shear Tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio ( R/a =3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and rf driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact Tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project.

Generation of net electric power with a tokamak reactor under foreseeable physical and engineering conditions

This study reveals for the first time the plasma performance required for a tokamak reactor to generate net electric power under foreseeable engineering conditions. It was found that the reference plasma performance of the ITER inductive operation mode with βN = 1.8, HH = 1.0, and fnGW = 0.85 had sufficient potential to achieve the electric break-even condition (net electric power ) under the following engineering conditions: machine major radius 6.5 m ≤ Rp ≤ 8.5 m, the maximum magnetic field on TF coils Btmax = 16 T, thermal efficiency ηe = 30%, and NBI system efficiency ηNBI = 50%. The key parameters used in demonstrating net electric power generation in tokamak reactors are βN and fnGW. βN ≥ 3.0 is required for with fusion power Pf ~ 3000 MW. On the other hand, fnGW ≥ 1.0 is inevitable to demonstrate net electric power generation, if high temperatures, such as average temperatures of Tave > 16 keV, cannot be selected for the reactor design. To apply these results to the design of a tokamak reactor for demonstrating net electric power generation, the plasma performance diagrams on the Q vs Pf (energy multiplication factor vs fusion power) space for several major radii (i.e. 6.5, 7.5, and 8.5 m) were depicted. From these figures, we see that a design with a major radius Rp ~ 7.5 m seems preferable for demonstrating net electric power generation when one aims at early realization of fusion energy.

Conceptual design of a breeding blanket with super-heated steam cycle for CREST-1

Abstract Conceptual design of a tritium breeding blanket for CREST-1 was conducted. CREST-1 (Compact REversed Shear Tokamak), a conceptual design of water-cooled commercial reactor based on a reversed shear high beta equilibrium, has been proposed as a possible scenario to an economical and feasible reactor succeeding the ITER project. In the present design study, a possibility of cost competitive fusion power plants with water-cooled concept, which has much experience in nuclear power plants, was examined. The new blanket design is based on the low activation ferritic steel components and an advanced super-heated steam cycle, which is used to realize a high thermal efficiency. High value of the thermal efficiency is very effective for reduction of the cost of electricity. On designing the blanket, allowable temperature range of the structure material, low activation ferritic steel is assumed to be 350–900 K with an expectation of the material research and development. Mixture of lithium oxide pebbles and beryllium pebbles is installed in the breeding zone for high tritium breeding ratio and high thermal conductivity of the breeding zone. Mixture ratio of beryllium and lithium-6 enrichment were optimized from viewpoints of temperature distribution in the breeding zone, achievable tritium breeding ratio and its reduction due to burn up. The designed blanket system has approximately 1.4 of local tritium breeding ratio with a 1.0 mm thickness of zirconium plate which is placed at 24 cm from the plasma surface as a conducting shell for kink stabilization. Arrangements of cooling channels and breeding zones and flow rate and inlet temperature of the coolant were also optimized to keep the temperatures of structure materials, breeding materials and coolant in the allowable range. The first wall is cooled by pressurized water at about 570 K. The coolant out of cooling channels of the first wall is lead to those of breeding zone and starts partially boiling. The steam is super-heated up to 750 K in the blanket. This high temperature raises the thermal efficiency of turbine to 41%. Our cost assessment has shown that CREST-1 generates about 1.16 GWe electric power within a competitive COE range.

Design of a steady-state tokamak device with superconducting coils for a volumetric neutron source

Abstract We designed a volumetric neutron source for testing large-scale blanket components, based on a steady-state tokamak device with superconducting coils. It is found that a neutron flux of approximately 1.0 MW m −2 is available in the medium-size device ( R =4.5 m, a =1.0 m, κ =1.8, I p =5.6 MA) under the conditions of H ∼2 and β N ∼3 with a neutral beam injection (NBI) power of about 60 MW. We demonstrate the controllability of the current profiles required for high-beta plasma up to β N =3–3.8 with the combination of bootstrap current and NB-driven current ( E b =1.0 MeV). If an advanced performance scenario such as a reversed shear configuration is available, a neutron flux of 1.4 MW m −2 is achievable. We install the breeding blanket of Li–Pb only at outboard and upper regions, and find that a local tritium breeding ratio (TBR) of 1.5 is achievable and a net TBR of 0.8 could be available. The analysis of shielding materials at the inboard region shows that the proper combination of tungsten, steel and boric water yields a reduction of the nuclear irradiation of TF coil by a factor of approximately 10.

Plasma core for commercially feasible reactor

Feasibility of commercial tokamak reactors is discussed. The range of cost of electricity (COE) by competitors of fusion is estimated and the target range of COE for the fusion reactors is proposed. It is shown that the high β advanced plasma mode will break through this COE target. As a possible path toward the cost competitive tokamak concepts, the reactor compact reversed shear tokamak (CREST), which is based on the high β plasma and super-heated steam cycle, is briefly reviewed.

Maintenance and radiation protection issues of the CREST reactor

A maintenance approach of the Compact Reversed Shear Tokamak (CREST) for high availability is proposed. Full sector removal through horizontal ports for easy maintenance is adopted in order to increase availability. Cask type sector removal machines are used to transfer the blanket/divertor sectors. Achievable availability is estimated with a maintenance period and its interval. The proposed maintenance approach allows, at least, the similar availability to that of the present nuclear plants. The estimated availability is acceptable from an economical viewpoint. Safety concerns related to the super-heated direct steam cycle are also examined. One of the unique concerns is due to nitrogen-16 production in the coolant. The super-heated steam must flow through the turbines within several 10 s after leaving the blanket. The requirements for shielding on the heat transfer systems are declared. Another concern is the confinement and recovery of tritium in the coolant, because a high temperature blanket may have a large tritium permeation rate. Water detritiation system proposed here would be able to control the tritium concentration in the coolant within an allowable range.

Design of a tritium breeding blanket for volumetric neutron source

Abstract A water-cooled and austenitic stainless-steel-structured breeding blanket system is designed for a volumetric neutron source (VNS), based on a steady-state tokamak device. The designed VNS with super-conducting coils has a 4.5 m main radius and a total of 300 MW in fusion power. It yields the tritium consumption of approximately 10 kg per year with 50% availability. It is unrealistic to assume that this amount of tritium could be supplied by other nuclear facilities, therefore, tritium breeding with the VNS device itself should be accommodated. The capability of tritium breeding for the Li-Pb breeder is examined. Compared to ceramic breeding materials, in general, the Li17Pb83 breeder has the potential for a higher tritium breeding ratio. In addition, the compatibility of Li-Pb with water coolant and with structure material at low temperature and low flow velocity may be acceptable. The calculated local TBR in the present design reached 1.4 or more in the Li-Pb breeder with a sufficient thickness of the breeding zone. The temperature distribution in the breeding region of the Li-Pb blanket was controlled to avoid corrosion of the structure material, using a double-walled cooling panel. The net TBR might attain the acceptable value when only the outboard Li-Pb blanket is used, since the contribution of outboard blanket to tritium breeding is higher than its coverage.