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The CALorimetric Electron Telescope (CALET) on the International Space Station (ISS) consists of a high-energy cosmic ray CALorimeter (CAL) and a lower-energy CALET Gamma ray Burst Monitor (CGBM). CAL is sensitive to electrons up to 20 TeV, cosmic ray nuclei from Z = 1 through Z $\sim$ 40, and gamma rays over the range 1 GeV - 10 TeV. CGBM observes gamma rays from 7 keV to 20 MeV. The combined CAL-CGBM instrument has conducted a search for gamma ray bursts (GRBs) since Oct. 2015. We report here on the results of a search for X-ray/gamma ray counterparts to gravitational wave events reported during the LIGO/Virgo observing run O3. No events have been detected that pass all acceptance criteria. We describe the components, performance, and triggering algorithms of the CGBM - the two Hard X-ray Monitors (HXM) consisting of LaBr$_{3}$(Ce) scintillators sensitive to 7 keV to 1 MeV gamma rays and a Soft Gamma ray Monitor (SGM) BGO scintillator sensitive to 40 keV to 20 MeV - and the high-energy CAL consisting of a CHarge-Detection module (CHD), IMaging Calorimeter (IMC), and fully active Total Absorption Calorimeter (TASC). The analysis procedure is described and upper limits to the time-averaged fluxes are presented., 27 pages, 10 figures. Accepted for publication in ApJ

We present the measurement of the energy dependence of the boron flux in cosmic rays and its ratio to the carbon flux in an energy interval from 8.4 GeV/n to 3.8 TeV/n based on the data collected by the Calorimetric Electron Telescope (CALET) during ∼6.4 yr of operation on the International Space Station. An update of the energy spectrum of carbon is also presented with an increase in statistics over our previous measurement. The observed boron flux shows a spectral hardening at the same transition energy E_{0}∼200 GeV/n of the C spectrum, though B and C fluxes have different energy dependences. The spectral index of the B spectrum is found to be γ=-3.047±0.024 in the interval 25E200 GeV/n. The B spectrum hardens by Δγ_{B}=0.25±0.12, while the best fit value for the spectral variation of C is Δγ_{C}=0.19±0.03. The B/C flux ratio is compatible with a hardening of 0.09±0.05, though a single power-law energy dependence cannot be ruled out given the current statistical uncertainties. A break in the B/C ratio energy dependence would support the recent AMS-02 observations that secondary cosmic rays exhibit a stronger hardening than primary ones. We also perform a fit to the B/C ratio with a leaky-box model of the cosmic-ray propagation in the Galaxy in order to probe a possible residual value λ_{0} of the mean escape path length λ at high energy. We find that our B/C data are compatible with a nonzero value of λ_{0}, which can be interpreted as the column density of matter that cosmic rays cross within the acceleration region.

Abrasion of the brake lining of automobiles is one of the main antimony (Sb) sources on the road. Therefore, the road effluent possibly supplies Sb to the combined sewer collection system. However, Sb in road-sewer systems has attracted little concern, although heavy metals such as copper (Cu), zinc (Zn), cadmium (Cd), and lead (Pb) in sewer water have been studied by many previous authors. We investigated the effect of road effluent on Sb in the combined sewer water by collecting road effluent, road dust, and sewer water under rainy and dry weather conditions. Sb in road effluent showed a significantly higher concentration than the other types of samples, and the Sb concentration in sewer during wet weather was significantly higher than that during dry weather. Furthermore, the Sb concentration in sewer water decreased with time during a wash-off event. Clear positive relationships between Sb and Cu and between Sb and Ba in both road effluent and road dust extract indicate the effect of brake abrasion because the brake lining contains Cu, Sb, and Ba in high concentrations. Approximately 42% of Sb load occurred during the wash-off event, while the loads of Cu and Ba were much less. Unlike Cu and Ba, we conclude that Sb in combined sewer water largely depends on road effluent in wet weather due to the wash-off of road dust, which is probably associated with brake lining abrasion.

The Calorimetric Electron Telescope (CALET), in operation on the International Space Station since 2015, collected a large sample of cosmic-ray iron over a wide energy interval. In this Letter a measurement of the iron spectrum is presented in the range of kinetic energy per nucleon from 10 GeV$/n$ to 2.0 TeV$/n$ allowing the inclusion of iron in the list of elements studied with unprecedented precision by space-borne instruments. The measurement is based on observations carried out from January 2016 to May 2020. The CALET instrument can identify individual nuclear species via a measurement of their electric charge with a dynamic range extending far beyond iron (up to atomic number $Z$ = 40). The energy is measured by a homogeneous calorimeter with a total equivalent thickness of 1.2 proton interaction lengths preceded by a thin (3 radiation lengths) imaging section providing tracking and energy sampling. The analysis of the data and the detailed assessment of systematic uncertainties are described and results are compared with the findings of previous experiments. The observed differential spectrum is consistent within the errors with previous experiments. In the region from 50 GeV$/n$ to 2 TeV$/n$ our present data are compatible with a single power law with spectral index -2.60 $\pm$ 0.03., Comment: main text: 7 pages, 4 figures; supplemental material: 10 pages, 12 figures, 1 table. arXiv admin note: text overlap with arXiv:2012.10319

Abrasion of the brake lining of automobiles is one of the main antimony (Sb) sources on the road. Therefore, the road effluent possibly supplies Sb to the combined sewer collection system. However, Sb in road-sewer systems has attracted little concern, although heavy metals such as copper (Cu), zinc (Zn), cadmium (Cd), and lead (Pb) in sewer water have been studied by many previous authors. We investigated the effect of road effluent on Sb in the combined sewer water by collecting road effluent, road dust, and sewer water under rainy and dry weather conditions. Sb in road effluent showed a significantly higher concentration than the other types of samples, and the Sb concentration in sewer during wet weather was significantly higher than that during dry weather. Furthermore, the Sb concentration in sewer water decreased with time during a wash-off event. Clear positive relationships between Sb and Cu and between Sb and Ba in both road effluent and road dust extract indicate the effect of brake abrasion because the brake lining contains Cu, Sb, and Ba in high concentrations. Approximately 42% of Sb load occurred during the wash-off event, while the loads of Cu and Ba were much less. Unlike Cu and Ba, we conclude that Sb in combined sewer water largely depends on road effluent in wet weather due to the wash-off of road dust, which is probably associated with brake lining abrasion.

Hyper-Kamiokande is the next generation underground water Cherenkov detector that builds on the highly successful Super-Kamiokande experiment. The detector which has an 8.4~times larger effective volume than its predecessor will be located along the T2K neutrino beamline and utilize an upgraded J-PARC beam with 2.6~times beam power. Hyper-K's low energy threshold combined with the very large fiducial volume make the detector unique, that is expected to acquire an unprecedented exposure of 3.8~Mton$\cdot$year over a period of 20~years of operation. Hyper-Kamiokande combines an extremely diverse science program including nucleon decays, long-baseline neutrino oscillations, atmospheric neutrinos, and neutrinos from astrophysical origins. The scientific scope of this program is highly complementary to liquid-argon detectors for example in sensitivity to nucleon decay channels or supernova detection modes. Hyper-Kamiokande construction has started in early 2020 and the experiment is expected to start operations in 2027. The Hyper-Kamiokande collaboration is presently being formed amongst groups from 19 countries including the United States, whose community has a long history of making significant contributions to the neutrino physics program in Japan. US physicists have played leading roles in the Kamiokande, Super-Kamiokande, EGADS, K2K, and T2K programs., 6 pages, prepared as Snowmass2021 LOI

The final optics system of a fast ignition laser reactor concept with a Dry first-wall and a high repetition laser (FALCON-D) is discussed. A cartridge module is proposed for the final optics system. Preliminary neutronics analysis for the final optics systems concept is carried out with the three-dimensional Monte Carlo N-particle transport code MCNP-4C. The total neutron and gamma-ray fluxes attenuate by 4–5 orders of magnitude between inlet and outlet of the duct in the final optics system for FALCON-D. Both of the total neutron and gamma-ray fluxes at the outlet in the final optics system are less then 1.0 × 108 cm−2 s−1, which are about 10 times less than that of the total neutron gamma-ray fluxes at the position just after the shield blanket of the tokamak reactor concepts. Hence, we can qualitatively get the prospect that the cask and hot-cell maintenance, which are proposed for blanket replacement in the tokamak reactor concepts, are also applicable to FALCON-D. The maintenance approach of the final optics system is also proposed. The cartridge module built in the wall of the reactor room can be replaced from the outside of the reactor room. This built-in style cartridge module probably leads an efficient replacement method for the high plant availability.

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 (a=2, η c =0.05, η h =0.2). Assuming the pulse heat load limit of a dry wall to be 2J/cm2, chamber radius ofR=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.

CREST is a conceptual tokamak reactor design with high β plasma, high thermal efficiency, competitive cost and water-cooled ferritic steel components. Some of its parameters are similar to those of the ITER advanced mode plasma. In this manuscript, the specific issues and analysis on damage to TF coils of CREST were carried out based on the three-dimensional model of the CREST with the widely used code MCNP/4C and the IAEA latest released FENDL/2.1 data library. Damage to some specific regions of the TF coils near large openings and at the inboard mid-plane are calculated and analyzed. Parameters such as the distributions of nuclear heat density, fast neutron flux, dose rate to the epoxy insulator, and peak displacement dose to Cu conductor for the TF coil near these regions were calculated and analyzed. The shield thicknesses at these regions are optimized.

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.

A system code for designing laser fusion reactors has been developed, and the design possibility of the laser fusion commercial plant with a single dry wall chamber and a high repetition laser has been examined. In a fast ignition concept, it is confirmed that the target gain G ≥ 100 is achievable with the fusion pulse energy of 40 MJ. If we assume the laser repetition rate of 30 Hz and the pulse heat load limit of 2 J/cm2, the plant which has moderate electric power ( ∼ 400 MWe) can be designed with a dry wall chamber of which radius R = 5.6 m. Although there are several technological issues needed to be solved, this design point can be characterized as a first-step reactor of the laser fusion study because of its relatively low construction cost.

The compact reversed shear tokamak CREST is a conceptual tokamak reactor design with high β plasma, high thermal efficiency, competitive cost and water-cooled ferritic steel components. In this manuscript neutronics analysis on CREST is presented based on the three-dimensional (3D) model and calculations by MCNP/4C code and FENDL/2.0 data library. The comparison among the results of tritium breeding ratio (TBR) for the one-dimensional (1D), two-dimensional (2D) and 3D calculations shows that the results are consistent with each other. The system would be tritium self-sufficient when considering the blanket coverage fraction, etc. and even some uncertainty. It has an energy multiplication of 1.38. The maximum neutron wall loading and maximum damage rate at the first wall (FW) are 6.30 MW/m2 and 51.9 dpa/FPY, respectively. The FW material needs to be replaced every 3.8 years of operation with an availability of 75% if the neutron damage of ferritic steel is assumed to limit to 150 dpa. The total nuclear heat for TF coils is under the limit for TF coils in ITER case. The nuclear heat density for TF coils at the mid-plane is roughly the same as the limit for ITER.

We develop a simple Core-SOL-Divertor (C-S-D) model to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. To construct the simple C-S-D model, a simple core plasma model of ITER physics guidelines and a two-point SOL-divertor model are used. The simple C-S-D model is applied to the study of the HT-7U operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters. Effective methods for extending the operation space are also presented. From this study for the HT-7U operation space, it is shown that the C-S-D model is a useful tool to understand qualitatively the overall features of the plasma operation space.

A demonstration tokamak fusion power plant Demo-CREST is proposed as the device for early realization of net electric power generation by fusion energy. The plasma configuration for Demo-CREST is optimized to satisfy the electric breakeven condition (the condition for net electric power, ) with the plasma performance of the ITER reference operation mode. This optimization method is considered to be suitable for the design of a demonstration power plant for early realization of net electric power generation, because the demonstration power plant has to ensure the net electric generation. Plasma performance should also be more reliably achieved than in past design studies. For the plasma performance planned in the present ITER programme, net electric power from 0 to 500?MW is possible with Demo-CREST under the following engineering conditions: maximum magnetic field 16?T, thermal efficiency 30%, NBI system efficiency 50% and NBI current drive power restricted to 200?MW. By replacing the blanket system with one of higher thermal efficiency, a net electric power of about 1000?MW is also possible so that the performance of the commercial plant with Demo-CREST can also be studied from the economic point of view. The development path from the experimental reactor 'ITER' to the commercial plant 'CREST' through the demonstration power plant 'Demo-CREST' is proposed as an example of the fast track concept.

The present study reveals forthcoming break-even conditions of tokamak plasma performance for the fusion energy development. The first condition is the electric break-even condition, which means that the gross electric power generation is equal to the circulating power in a power plant. This is required for fusion energy to be recognized as a suitable candidate for an alternative energy source. As for the plasma performance (normalized beta value ΒN), confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density fnGW), the electric break-even condition requires the simultaneous achievement of 1.2 < ΒN < 2.7, 0.8 < HH, and 0.3 < fnGW < 1.1 under the conditions of a maximum magnetic field on the TF coil Btmax = 16 T, thermal efficiency ηe = 30 %, and current drive power PNBI < 200 MW. It should be noted that the relatively moderate conditions of ΒN ˜ 1.8, HH ˜ 1.0, and fnGW ˜ 0.9, which correspond to the ITER reference operation parameters, have a strong potential to achieve the electric break-even condition. The second condition is the economic break-even condition, which is required for fusion energy to be selected as an alternative energy source in the energy market. By using a long-term world energy scenario, a break-even price for introduction of fusion energy in the year 2050 is estimated to lie between 65 mill/kWh and 135 mill/kWh under the constraint of 550 ppm CO2 concentration in the atmosphere. In the present study, this break-even price is applied to the economic break-even condition. However, because this break-even price is based on the present energy scenario including uncertainties, the economic break-even condition discussed here should not be considered the sufficient condition, but a necessary condition. Under the conditions of Btmax = 16 T, ηe = 40 %, plant availability 60 %, and a radial build with/without CS coil, the economic break-even condition requires ΒN ˜ 5.0 for 65 mill/kWh of lower break-even price case. Finally, the present study reveals that the demonstration of steady-state operation with ΒN ˜ 3.0 in the ITER project leads to the upper region of the break-even price in the present world energy scenario, which implies that it is necessary to improve the plasma performance beyond that of the ITER advanced plasma operation.

Beam driven stable equilibria for a conceptual reactor, Demo-CREST, which was designed for early demonstration of electric power generation, has been investigated. Considering current profiles driven by neutral beams, the attainable normalized beta βN with a stabilization wall is about 3.4 with a normal shear (NS). With reversed shear (RS), a higher βN is attainable. The stable equilibria up to 4.0 can be sustained by a couple of On-and Off-axis beams. In the range of 1.9 20 keV is allowable.

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.

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.

Roles and requirements of the blanket system of the fusion power reactors are discussed from viewpoints of economics, fuel supply, generation system, maintenance, radioactive waste, and interaction with the plasma core. As the blanket system influences the cost of the fusion energy, the blanket system must be designed to minimize the fusion energy cost. Tritium breeding performance of the blanket is indispensable role to show the advantage of fusion energy on energy security. Material development for high temperature use under high neutron flux is one of the key issues of the generation system because the thermal efficiency depends on the coolant temperature of the blanket. Innovative maintenance technologies such as dividable superconducting coil system are very effective to make the fusion power reactor attractive. From viewpoints of natural resources and waste management, materials used in the fusion reactors should be recycled. Material selection is also of a large importance on the cost of radioactive waste disposal. Finally, it must be paid a careful attention that the design of the blanket system is inseparable from the achievement of a high performance plasma core.

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.

Supply of tritium for initial loading was concerned to be a limit for fusion power plant to increase its number in the future. In order to consider the implication of the possible tritium self-production, the potential of fusion energy in the future electricity supply market was estimated. Future energy market is analyzed with world energy and environment model that describes composition of supply-side energy system structures under economical and environmental constraint to meet the world energy demand. In the model, composition of supply-side energy system structures is determined to limit the CO 2 concentration of 550ppm in 2100 with minimal energy system cost. The result revealed that after introduction to the market, share of the fusion energy is strongly restricted by the initial tritium supply. Capability to produce initial loading of tritium removes this limitation, and future fusion share could be doubled.

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.

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.

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.

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.

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.

The compact reversed shear tokamak CREST is a cost competitive reactor concept based on a reversed shear high β plasma and water cooled ferritic steel components. The moderate aspect ratio A = 3.4 and the elongation κ = 2.0 of CREST are very similar to the case of the ITER advanced mode plasma. Presentation of such a concept based on the ITER project should be worth while for formulating a fusion development strategy. The achievement of a competitive cost of electricity (COE) is the first priority for electric power industries. High β and high thermal efficiency are the most effective parameters for achieving a competitive COE. In order to achieve a high efficiency power plant, a superheated steam cycle has been adopted which permits a high thermal efficiency (η = 41%). Current profile control and high speed plasma rotation by neutral beam current drive stabilize the ideal MHD activity up to the Troyon coefficient βN = 5.5. A cost assessment has shown that CREST could generate about 1.16 GW(e) electric power at a competitive cost.

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.