K. Tokimatsu

Role of nuclear fusion in future energy systems and the environment under future uncertainties

Abstract Debates about whether or not to invest heavily in nuclear fusion as a future innovative energy option have been made within the context of energy technology development strategies. This is because the prospects for nuclear fusion are quite uncertain and the investments therefore carry the risk of quite large regrets, even though investment is needed in order to develop the technology. The timeframe by which nuclear fusion could become competitive in the energy market has not been adequately studied, nor has roles of the nuclear fusion in energy systems and the environment. The present study has two objectives. One is to reveal the conditions under which nuclear fusion could be introduced economically (hereafter, we refer to such introductory conditions as breakeven prices) in future energy systems. The other objective is to evaluate the future roles of nuclear fusion in energy systems and in the environment. Here we identify three roles that nuclear fusion will take on when breakeven prices are achieved: (i) a portion of the electricity market in 2100, (ii) reduction of annual global total energy systems cost, and (iii) mitigation of carbon tax (shadow price of carbon) under CO2 constraints. Future uncertainties are key issues in evaluating nuclear fusion. Here we treated the following uncertainties: energy demand scenarios, introduction timeframe for nuclear fusion, capacity projections of nuclear fusion, CO2 target in 2100, capacity utilization ratio of options in energy/environment technologies, and utility discount rates. From our investigations, we conclude that the presently designed nuclear fusion reactors may be ready for economical introduction into energy systems beginning around 2050–2060, and we can confirm that the favorable introduction of the reactors would reduce both the annual energy systems cost and the carbon tax (the shadow price of carbon) under a CO2 concentration constraint.

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.

Study of design parameters for minimizing the cost of electricity of tokamak fusion power reactors

The impact of the design parameters on the cost of electricity (COE) is studied through a parameter survey in order to minimize the COE. Three kinds of operating modes are considered; first stability (FS), second stability (SS) and reversed shear (RS). The COE is calculated by a coupled physics-engineering-cost computer system code. Deuterium-tritium type, 1000 MW(e) at electric bus bar, steady state tokamak reactors with aspect ratios A from 3 to 4.5 are assumed. Several criteria are used for the parameter survey; for example, (a) the thermal to electrical conversion efficiency is assumed to be 34.5% using water as a coolant; (b) the average neutron wall load must not exceed 5 MW/m2 for plasma major radius Rp > 5 m; (c) a 2 MeV neutral beam injector (NBI) is applied. It is found that the RS operating mode most minimizes the COE among the three operating modes by reducing the cost of the current drive and the coils and structures. The cost-minimized RS reactor can attain high fbs, high βN and low q95 at the same time, which results in a short Rp of 5.1 m, a low Bmax of the maximum magnetic toroidal field (TF) of the TF coils of 13 T and a low A of 3.0. It can be concluded that this cost-minimized RS reactor is the most cost-minimized within the frameworks of this study. This cost-minimized RS reactor has two advantages: one is that a Bmax = 13 T TF coil can be made by use of ITER coil technology and the other is that the same cooling technology as that of ITER (water cooling) can be used.

Energy analysis and carbon dioxide emission of tokamak fusion power reactors

Energy gain and carbon dioxide (CO 2 ) emission of tokamak fusion power reactors are evaluated in this study compared with other reactor types, structural materials, and other Japanese energy sources currently in use. The reactors treated in this study are (1) a conventional physics performance international thermonuclear experimental reactor (ITER), like a reactor based upon the ITER engineering design activity (ITER-EDA), (2) a RS (reversed shear) reactor using the reversed shear safety-factor/plasma current profile, and (3) a ST (spherical torus) reactor based upon the final version of the advanced reactor innovative engineering study ST (ARIES-ST). The input energy and CO 2 emission from these reactors are calculated by multiplying the weight or cost of the fusion reactor components by the energy intensity and/or with the CO 2 intensity data, which are updated as often as possible. The ITER cost estimation is estimated based on the component unit costs. The following results were obtained: (1) The RS and the ST reactor can double the energy gain and reduce CO 2 emission by one-half compared with the ITER-like reactor. (2) Silicon carbide (SiC) used as the structural material of inner vessel components is best for energy gain and CO 2 emission reduction. (3) The ITER-like reactor is slightly superior to a photovoltaic (PV) with regard to CO 2 emission. (4) The energy gain and CO 2 emission intensity of the RS reactor and the ST reactor are as excellent as those of a fission reactor and a hydro-powered generator. These results indicate that a tokamak fusion power reactor can be one of the most effective power-generating technologies both in high-energy payback gains and reduction of CO 2 .

Socio-economic study of fusion energy at the Japan Atomic Energy Research Institute

Abstract The socio-economic studies of fusion from various aspects are performed by Japan Atomic Energy Research Institute (JAERI) and its collaborators in Japan, where steady-state tokamaks have been designed as a promising candidate. Energy model analysis estimates the significant contribution of fusion in the latter half of the century under the global environment constraints if it will successfully developed and introduced into the market. Application to fuel production process such as hydrogen by utilizing heat is designed to increase possible contribution for energy supply further. Operation without initial tritium loading liberates fusion from fuel supply constraints, and has an effect to rapidly increase fusion share. Externality of fusion energy is analyzed by CO 2 emission, impact pathway of tritium, rad-waste, and some social impacts. These results are expected to answer the questions on significance of fusion development, and improve the plant design and development strategy to maximize the benefit of fusion research.

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.

A novel method of TVTS in the TS-3 device and the proposal of its application to a large device

Abstract A novel method of Television Thomson Scattering (TVTS) has been developed in the TS-3 device. In this system, a framing camera is located between a spectroscope and a close coupled device camera. This framing camera can take two successive frames from the exit of the spectroscope. The time interval is of 500 ns between those two frames. Because of the shortness of this time interval, the background light is negligible; moreover TVTS is applicable to the TS-3 device whose plasma lifetime is about 150 gs. This method indicates the possibility of not only high spatial resolution but also time repetition in a simple system.

A cost minimized tokamak reactor using conservative coils/cooling technology

Abstract One attractive concept for future commercial reactors is one of low cost with minimum advanced technology. Cost-minimized reactors with different operating modes are compared to reveal the most cost-minimized reactor with impact of deign parameters on the cost of electricity (COE). Three kinds of operating modes are considered; i.e. first stability region (FS), second stability region (SS) and reversed shear (RS). Deuterium–tritium fueled, 1000 MW(e) reactors with conventional aspect ratio are assumed. Several criteria for the parameter survey are used; for example, (1) thermal-to-electricity conversion efficiency is assumed to be 34.5% with water as a coolant, (2) average neutron wall load must not exceed 5 MW m −2 with a plasma major radius >5 m, (3) 2 MeV neutral beam injector is applied. It is concluded that the RS operating mode is the most cost-minimized. The results indicated that attaining high β N (the Troyon beta coefficient), low q 95 (safety-factor on the 95% flux surface), and high f bs (fraction of bootstrap current) is the best way to obtain a cost-minimized reactor. This cost-minimized RS reactor is advantageous because it can be designed without using advanced coil technology nor an advanced cooling system.

Feasibility study of a volumetric neutron source based on a superconducting tokamak device

We design a volumetric neutron source based on a steady-state tokamak device with superconducting coils. It is found that the neutron flux of 1 MW/m/sup 2/ is available in the medium-sized device (R=4.5 m, a=1.0 m, /spl kappa/=1.8 and I/sub p/=5.6 MA) under the conditions of H=2 and /spl beta//sub N/=3 with a NBI power of 60 MW. We demonstrate the controllability of the current profiles required for high beta plasma up to /spl beta//sub N/=3 with the combination of the bootstrap current and NB-driven current. The Li-Pb blanket is adopted with water-cooled and austenitic stainless steel-structured system. The calculated local TBR reaches 1.4 or more in Li-Pb breeder at a sufficient thickness of breeding zone. The net TBR might be available only with the outboard Li-Pb blanket.