Yoshiyuki Asaoka

Design and technology development of solid breeder blanket cooled by supercritical water in Japan

This paper presents results of conceptual design activities and associated RD neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was successfully fabricated. It withstood the high heat flux test at 2.7 MW m−2. Also, a correlation parameter of the Li2TiO3 pebble bed made by the sol–gel method was verified by measurement of the thermal conductivity of the breeder pebble bed, which is one of the most important design data.

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.

Studies of breakeven prices and electricity supply potentials of nuclear fusion by a long-term world energy and environment model

In response to social demand, this paper investigates the breakeven price (BP) and potential electricity supply of nuclear fusion energy in the 21st century by means of a world energy and environment model. We set the following objectives in this paper: (i) to reveal the economics of the introduction conditions of nuclear fusion; (ii) to know when tokamak-type nuclear fusion reactors are expected to be introduced cost-effectively into future energy systems; (iii) to estimate the share in 2100 of electricity produced by the presently designed reactors that could be economically selected in the year. The model can give in detail the energy and environment technologies and price-induced energy saving, and can illustrate optimal energy supply structures by minimizing the costs of total discounted energy systems at a discount rate of 5%. The following parameters of nuclear fusion were considered: cost of electricity (COE) in the nuclear fusion introduction year, annual COE reduction rates, regional introduction year, and regional nuclear fusion capacity projection. The investigations are carried out for three nuclear fusion projections one of which includes tritium breeding constraints, four future CO2 concentration constraints, and technological assumptions on fossil fuels, nuclear fission, CO2 sequestration, and anonymous innovative technologies. It is concluded that: (1) the BPs are from 65 to 125 mill kW−1 h−1 depending on the introduction year of nuclear fusion under the 550 ppmv CO2 concentration constraints; those of a business-as-usual (BAU) case are from 51 to 68 mill kW−1h−1. Uncertainties resulting from the CO2 concentration constraints and the technological options influenced the BPs by plus/minus some 10–30 mill kW−1h−1, (2) tokamak-type nuclear fusion reactors (as presently designed, with a COE range around 70–130 mill kW−1h−1) would be favourably introduced into energy systems after 2060 based on the economic criteria under the 450 and 550 ppmv CO2 concentration constraint, but not selected under the BAU case and 650 ppmv CO2 concentration constraint, and (3) the share of electricity in 2100 produced by the presently designed tokamak-type nuclear fusion reactors (introduced after 2060) is well below 30%. It should be noted that these conclusions are based upon varieties of uncertainties in scenarios and data assumptions on nuclear fusion as well as technological options.

Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features ofthe prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900°C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal conductivities of 15 and 60 W/m per K are assumed in the prototype and the commercial type, respectively.

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.

Conceptual design of fast-ignition laser fusion reactor FALCON-D

A new conceptual design of the laser fusion power plant FALCON-D (Fast-ignition Advanced Laser fusion reactor CONcept with a Dry wall chamber) has been proposed. The fast-ignition method can achieve sufficient fusion gain for a commercial operation (~100) with about 10 times smaller fusion yield than the conventional central ignition method. FALCON-D makes full use of this property and aims at designing with a compact dry wall chamber (5–6 m radius). 1D/2D simulations by hydrodynamic codes showed a possibility of achieving sufficient gain with a laser energy of 400 kJ, i.e. a 40 MJ target yield. The design feasibility of the compact dry wall chamber and the solid breeder blanket system was shown through thermomechanical analysis of the dry wall and neutronics analysis of the blanket system. Moderate electric output (~400 MWe) can be achieved with a high repetition (30 Hz) laser. This dry wall reactor concept not only reduces several difficulties associated with a liquid wall system but also enables a simple cask maintenance method for the replacement of the blanket system, which can shorten the maintenance period. The basic idea of the maintenance method for the final optics system has also been proposed. Some critical R&D issues required for this design are also discussed.

Analysis of critical development issues towards advanced tokamak power plant CREST

A development scenario of the tokamak reactor in three stages (i.e. the experimental reactor ITER, a demonstration reactor and a commercial reactor) has been recently discussed. In order to construct a feasible development strategy, it is necessary to evaluate which component of reactor technologies and to what extent should be developed. From the viewpoint of the future electric supplier, we have proposed the conceptual design of a commercial power plant, compact reversed shear tokamak (CREST), and a demonstration power plant, Demo-CREST. On the other hand, the project of the experimental reactor ITER is underway, and its experimental plan and R&D activities are almost completed. Hence, it is most important and reasonable to investigate the demonstration power plant on the track of ITER in order to show a specific development scenario of the tokamak reactor. In this report, we discuss the engineering aspect in the Demo-CREST design and analyse the critical development issues towards an advanced tokamak CREST. The power flow and power plant system for Demo-CREST are investigated for improvement in the thermal efficiency of a single device, and the development goals for each reactor component and for each development step are quantitatively analysed.

Preliminary consideration on maintenance approach for a fast ignition ICF reactor with a dry wall chamber and a high repetition laser

Abstract The fast ignition method enables a reduction of the laser power required to achieve a large energy gain. This suggests consideration of a new inertial confinement fusion power plant concept, which has a small fusion pulse and a high repetition laser with a dry wall chamber. To establish the potential of the fast ignition method and to make clear the critical issues, a Fast Ignition ICF reactor concept with a Dry Wall chamber and a High Repetition Laser (FI-DWHRL concept) was previously proposed. The maintenance approach for this Fast Ignition ICF reactor concept is preliminary considered and its critical issues are described in this paper. The large cask and the large maintenance port for replacing the blanket sectors are applied to this Fast Ignition ICF reactor concept. The first wall and blanket system is divided into 20 sectors and all beam lines go between blanket sectors. The vacuum vessel is located outside the blanket system and this vacuum vessel also serves as the tritium boundary. To replace the final optical device, 6 access corridors are placed along the reactor room. Finally, critical issues on this maintenance approach are listed.

Design study of dry wall fast ignition laser fusion reactor with high repetition laser

Abstract The design of a laser fusion reactor with a dry wall chamber has been carried out. According to a simple point model calculation, sufficient pellet gain (G > 100) can be achieved with the injection energy of 400kJ under relatively conservative parameters (α = 2, ηc = 0.05, ηh = 0.2). Assuming the pulse heat load limit of a dry wall to be 2J/cm2, chamber radius of R = 5.64m is achievable. 1-D thermal analysis also supports the feasibility of this design. Then a medium scale plant (400MWe electric output) can be designed with moderate construction cost, which suits for the first-step reactor, if the laser repetition rate can be increased to 30 Hz. Since laser fusion reactors have flexibility in changing its output, this design enables them to be in flexible use according to the time-varying electric demand as the present fossil fuel power plants. This design is remarkable because it gives a new property to the fusion reactors.