Ryouji Hiwatari

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.

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 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.

Conceptual design study and strategy toward fusion demonstration plants

A report in 2005 by the Atomic Energy Commission of Japan has stated an expectation to secure the prospect of putting fusion into practical use by the middle of 21st century. A roadmap based on this policy was developed in 2008. The roadmap consists of a breakdown list of works which has shown and categorized the R&D issues required to construct the DEMO plants. Two tokamak DEMO concepts, SlimCS (Rp=5.5m) and Demo-CREST (Rp=7.3m), have been proposed in Japan as possible DEMO designs which will fit in the policy.

A Design system for layout of Charging infrastructure for Electric Vechicle

We introduce a design system a layout of the charging infrastructures for an electric vehicle (EV).The design system consists of a traffic simulator for EVs and charging infrastructures, and a pre-post tool, which produces the input files into and the resultant figures from the traffic simulator. The traffic simulator can analyze the location of the dead EV(which means the EV running out of electricity) and the number of charging EV at each charging station(ST).We also have proposed the search algorithm for the effective layout of charging STs based on the location of the dead EV by the road traffic simulator. That algorithm has been installed into the traffic simulator. The layout of charging STs is successfully determined to reduce the number of the dead EV.

Thermal-hydraulic design of direct superheated steam blanket for the compact reversed shear tokamak reactor (CREST)

The compact reversed shear tokamak, CREST, is a cost competitive power reactor concept based on the reversed shear high beta plasma. As the cooling system of CREST blanket, the direct superheated steam cycle is adopted to simplify the heat transport/power generation system and to achieve high thermal efficiency with a water cooled concept. Thermal-hydraulic design of the blanket was performed for the nominal operation, the start-up operation and the partial load operation. As a result of the study, it was confirmed that the direct steam cycle blanket is promising for a power reactor and is feasible not only at the nominal operation condition but also at the reactor start-up and the partial load conditions.

Commissioning scenario without initial tritium inventory for a demonstration reactor demo-crest

Abstract We discuss the applicability of the commissioning scenario without the initial tritium inventory to Demo-CREST. Analysis on MHD stability and current drive property (i.e., NBI injection power, its injection region, the driven current profile, etc.) makes clear the potential to start up the plasma operation without the initial tritium inventory. The critical issue on the core plasma operation is the high confinement of HH=1.57. We also discuss the tritium dead inventory in the plasma area. The key for the commissioning period without the initial tritium inventory is found to be the increment of the dead inventory. Finally, the required commissioning period is estimated at 75˜110 days for the net TBRDT=1.05. That possibility strongly depends on the increment of the dead inventory, and understanding the tritium behavior not only in the plasma region but also in other tritium subsystem is important.

Maintenance method and its critical issues for a fast-ignition laser fusion reactor based on a dry wall chamber

The neutronics analysis has been carried out for feasibility study of the FALCON-D concept by Monte Carlo N-paticle transport code (MCNP), in order to inspect the cooling performance of in-vessel and ex-vessel components, and a connection pipe between Vacuum Vessel and reactor room. The nuclear heating rate in the Vacuum Vessel was at the same level as that of NBI duct of the ITER. The temperature of the connection pipe was found to be 345, which was smaller than the melting point of structure materials (F82H). Moreover, the radiation damage of the final optics was also investigated. We propose a sliding changer concept for replacement. This method could be adapted for the replacement of one FPY cycle in the final optics system.