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The current status and the progress of R&D activities for a Fusion DEMO reactor in the QST Rokkasho Fusion Institute is reported. In order to advance the JA DEMO activity, not only Japanese domestic activity but also international collaborations of Broader Approach Activity and ITER related activities are conducted in the QST Rokkasho Fusion Institute. Activities for DEMO design and relevant R&D, design of a fusion neutron source and development of an accelerator, ITER-TBS, tritium handling technology and information technology infrastructures including a supercomputer system and a remote experimentation system are carried out for a Fusion DEMO reactor.

In the development of a DEMO fusion reactor, mitigating the heat load deposited on the breeding blanket at steady-state operation is one of the challenging issues because of the localized high heat flux brought by charged particles and low thermal conductivity of the Reduced Activation Ferritic/Martensite (RAFM) steel structural material. This study proposes a concept of limiter for JA DEMO; a horseshoe-shaped limiter which is continuous poloidally in the first wall except for the divertor area, and discretized toroidally. The limiter is protruded from the first wall to shade the blankets from the charged particles, which move along the magnetic field lines. To enhance the heat removal capability and neutron irradiation tolerance, a tungsten mono-block using RAFM heat sink was applied as the plasma-facing unit (PFU) on the limiters. The heat removal capability of the limiter’s PFU was evaluated as 4.6 MW/m2. The limiter shape satisfying this value was determined basing on the utilization of heat load evaluation with three-dimensional magnetic field line tracing. Three limiter shape parameters; protrusion height, protrusion width and the first wall set-back, were investigated to understand the variation of charged particle load distribution. Both cylindrical and box-shaped blanket concepts were considered, and design windows for the limiter shape were acquired to discuss the optimum parameter. The optimum limiter shape fulfilling the requirement of steady-state operation has been defined, and the limiter’s occupied area amounts to 1.2 % of the FW area for both blanket options.

JA DEMO selected the vertical maintenance scheme as the primary maintenance option. In order to improve reliability of Remote Handling (RH) and plant availability, update of in-vessel transferring mechanism of the segment, pipe unit structure on maintenance port were carried out. The RH equipment is composed of end-effectors (grippers) for the banana segment, a power manipulator, a telescopic guide and a carrier. The pipe structure in vertical maintenance port was updated considering the thermal expansion, easy handling (short maintenance time) and alignment of welding groove. Based on estimation of maintenance time, four works in parallel will be required for plant availability reachable for commercialization (>60%).

JA DEMO selected the vertical maintenance scheme as the primary maintenance option. In order to improve reliability of Remote Handling (RH) and plant availability, update of in-vessel transferring mechanism of the segment, pipe unit structure on maintenance port were carried out. The RH equipment is composed of end-effectors (grippers) for the banana segment, a power manipulator, a telescopic guide and a carrier. The pipe structure in vertical maintenance port was updated considering the thermal expansion, easy handling (short maintenance time) and alignment of welding groove. Based on estimation of maintenance time, four works in parallel will be required for plant availability reachable for commercialization (>60%).

A conceptual design of breeding blanket module with pressure tightness against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. The cooling water for DEMO is operated at the PWR water conditions of 15.5 MPa and 290 oC-325 oC. In this design, the breeding area of the module is divided into 0.1-m-squared cells with rib structure and has simple interior for mass production using a mixed pebbles bed of Li2TiO3pebbles and Be12Ti ones. As a result, a rib with the thickness of 0.015m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. The cooling system for the blanket module is designed by fluid dynamics analysis based on the PWR water conditions, and the outlet coolant temperature and the pressure drop are 321 oC and 0.32 MPa, respectively. It was found that the self-sufficient production of tritium is likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with an idea to improve TBR that the coolant is changed from existing light water to heavy water.

his paper analyzed Japan’s fusion energy development scenarios by using a global energy systems model considering a combination of uncertainties of future socioeconomic development, CO2 emission pathways corresponding to the long-term target of the Paris Agreement and future fusion power plants. Fusion power plants will be installed in Japan and could contribute about 20% on average in power sector in the latter half of the century because of the limited-potentials of affordable renewable energies and carbon capture and storage in Japan. The benefit of fusion introduction in Japan is estimated to be about 10 billion US$/yr on average in 2050-2100. Its value increases with reduction of the capital costs of future fusion power plants, lower penetration of fission power plants and deeper decarbonization pathways. When considering unique characteristics of fusion energy development, i.e., long-term and large-scale projects and DEMO as a single step between ITER and a first commercial plant, demonstration and prospect of economic viability under DEMO projects becomes critically important. This systems analysis also suggests that an alternative option of fusion energy which directly produces hydrogen by high-temperature steam electrolysis could become economically attractive.

Disruption is a critical phenomenon in a tokamak reactor. Although disruption causes serious damage to the reactor, its physical mechanism remains unclear. To realize a tokamak reactor, it is necessary to understand and control the disruption phenomenon. The present research constructs a disruption predictor using experimental high-beta plasma data in the JT-60U tokamak. The predictor was constructed using a support vector machine as a linear discriminant, and we focus on a variable selection problem for the binary classification by sparse modeling, specifically, exhaustively searching the best combinations of variables which maximize the predictor performance. By the sparse modeling, we found that the six input parameters as the best combinations. The selected parameters were the n = 1 mode amplitude | B r n = 1 | and its time derivative d | B r n = 1 | / d t , the plasma density (relative to the Greenwald density limit) and its time derivative, and the time derivatives of the plasma internal inductance and plasma elongation. In particular, it was identified that the parameter d | B r n = 1 | / d t , plays a key role on plasma disruption. We should notice that the combination with other plasma parameters is indispensable and remarkably make it possible to improve the performance of disruption prediction.

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan-Europe Broader Approach DEMO Design Activity (BA DDA). A common critical issue is the large power exhaust in the main plasma and divertor by the radiative cooling. (i) Different exhaust concepts in the main plasma and divertor have been developed for JA and EU DEMOs, and both divertor operations were systematically simulated and compared. (ii) Integrated designs of the heat removal components and cassette structures also have been developed. (iii) Both divertor designs are adequate to handle the peak heat loads on the target (qtarget) of 10 MW/m2 level under neutron irradiation condition., 28th IAEA Fusion Energy Conference

原型炉設計合同特別チームのこれまでの成果である日本の原型炉概念(JA DEMO)の基本設計として、核融合科学技術委員会が示した原型炉の目標（① 数十万kWを超える定常かつ安定した電力、② 実用に供しうる稼働率,③ 燃料の自己充足性を満足する総合的なトリチウム増殖）を満足する定常炉概念と今後のR&D課題について報告した。特に主な成果として、発電プラント設備を含む原型炉全体像の構築とそれ基づく正味電気出力の明確化、耐圧性とトリチウム生産性を両立するトリチウム増殖ブランケットの概念設計、及び原型炉の安全性確保に向けた分析した想定起因事象とその緩和方策について報告した。, 第28回 国際原子力機関（IAEA）核融合エネルギー会議（IAEA-FEC2020）

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan-Europe Broader Approach DEMO Design Activity (BA DDA). A common critical issue is the large power exhaust in the main plasma and divertor by the radiative cooling. (i) Different exhaust concepts in the main plasma and divertor have been developed for JA and EU DEMOs, and both divertor operations were systematically simulated and compared. (ii) Integrated designs of the heat removal components and cassette structures also have been developed. (iii) Both divertor designs are adequate to handle the peak heat loads on the target (qtarget) of 10 MW/m2 level under neutron irradiation condition., 28th IAEA Fusion Energy Conference

The neutronics of a fusion reactor is crucial in the conceptual design of all in-vessel components and the overall reactor design. A successful DEMO reactor must demonstrate self-sufficient production of tritium to allow sufficient plant availability and electrical generation. The neutronics shows that the design constraint of the neutron wall load (NWL) is related to the tritium breeding ratio and nuclear heating, which determine the fusion power and plasma configuration and are essential for the blanket design in a fusion reactor. To align the outlet temperatures of the blanket modules, the coolant piping system must be arranged to match the distribution of the NWL. 3D neutronics analysis ensures that the blanket design allows a self-sufficient supply of tritium. The maintenance scheme for the remote handling equipment to operate Japan’s DEMO reactor should consider the shutdown dose rate and residual heat in the vacuum vessel. Base on neutronics analysis, it is important to classify radwaste disposal and to develop radwaste management scenarios. The analyses in this paper show that the radiation exposure caused by nuclides is as low as 3.4 μSv/y when the radwaste is disposed in a shallow concrete pit. This radiation exposure level is below the shallow-land burial limit in Japan (

Power exhaust scenario and divertor design for a steady-state Japan (JA) DEMO and a pulse Europe (EU) DEMO1 have been investigated as one of the most important common issues in Broader Approach DEMO Design Activity. Radiative cooling is a common approach for the power exhaust scenario. For the JA DEMO, development of the divertor design appropriate for high Psep/Rp ~30 MWm-1 is required, while radiation fraction in the main plasma (fradmain = Pradmain/Pheat) is ITER-level (0.40-0.45) and the exhaust power above the L to H-mode power threshold (fLH = Psep/PthLH) is large margin (~2). For the EU DEMO1, larger fradmain (= 0.67) and smaller fLH (=1.2) plasma is required, using higher-Z impurity seeding, in order to apply ITER-level devertor (Psep/Rp = 17 MWm-1). ITER technology, i.e. water cooling with W-monoblock and Cu-alloy (CuCrZr) heat sink, is a baseline for JA and EU to handle the peak heat load of 10 MWm-2-level, and neutron flux and dose are comparable. For the JA DEMO, two different water-cooling pipes, i.e. CuCrZr and F82H steel, are proposed. For the EU DEMO1, heat sink consists of all Cu-alloy pipe, and several choices of the heat sink components have been developed appropriate to the high neutron irradiation condition. These JA and EU approaches of the power exhaust scenario will provide important case-studies for the future decision of the DEMO divertor design.

The primary cooling water systems (CWS) of the blanket and the divertor is anticipated to have several amount of tritium because of permeation from the core plasma. In this research, the required performance of DEMO tritium removal facility to control the primary tritium concentration has been calculated and it is shown that existing devices can be applied to DEMO. The primary tritium is also anticipated to permeate the pipes to the downstream CWS and turbine system. Finally, tritium in turbine system permeates a condenser pipes and is leaked into seawater. The evaluation of tritium leakage rate into seawater has been done. Because of tritium transport model from water to water is not well known, gas to gas calculation model is used. The results show that a heat exchanger decreases the tritium permeation rate by 1 order of magnitude, and H2 addition to the upstream cooling water also decreases the tritium permeation rate by 3 orders of magnitude.

The requirements for Japanese DEMO are to demonstrate (1) steady and stable electrical power generation, (2) plant availability on a level relevant for commercialization, and (3) self-sufficiency of fuels. The DEMO design activity has been mainly focused on challenging DEMO design issues such as divertor heat removal, superconducting magnets, breeding blankets, remote maintenance and the water cooling system, in order to reach the systemic consistency toward a promising DEMO concept. The current status of the DEMO conceptual design activity is presented.

Tritium is required for research and development activities for the deuterium–tritium (DT) fusion reactor and fueling the DEMOnstration Power Station (DEMO). However, tritium is a very rare nuclide and must be produced artificially. Tritium production by loading Li compounds (Li rods) into burnable poison holes of a high-temperature gas-cooled reactor (HTGR) has been proposed (H. Matsuura, et al., Nucl. Eng. Des. 243 (2012) 95–101.). Al2O3 and Zr are used to prevent tritium leaks. Nuclear reaction heat caused by the nuclear reaction (e.g., 6Li(n,α)T reaction) can cause a spatial temperature profile in the Li rods and may change its tritium containment performance, because Al2O3 and Zr performance strongly depend on these temperatures. The effect of nuclear reaction heat by the 6Li(n,α)T reaction on the tritium containment performance of the Li rods was evaluated by simulation. The temperatures of the Li rods for the high-temperature engineering test reactor (HTTR) and gas turbine high-temperature reactor 300 (GTHTR300) increased by 36 K and 46 K, and the leaked tritium decreased by 32% and 37% via nuclear reaction heat, respectively.

Considering the water cooling system in fusion reactors, there are several fusion-specific challenges in designs of in-vessel components and their cooling water systems (CWS). In this research, solutions of the challenges have been discussed and indicate the CWS concept of Japan's DEMO. The CWS transfers thermal energy of blanket and a part of divertor to a generator system, and thermal energy of the other part of divertor is used as heat utilization in plant. The required performance of system devices such as pumps is expected to be achieved by proven technologies and its extensions. It is also indicated that blanket can work as the low-pass filter, and can suppress the effect of fusion output fluctuation. The designed cooling water system requires 70 MW electric power and generates 620 MW power with a 1.5 GW fusion power plasma. This research indicates the basic concept of Japan's DEMO CWS that can be practically achieved.

The production of tritium (T) using high-temperature gas-cooled reactors (HTGRs) has been studied for a prior engineering research with T handling and initial T possession in demonstration fusion reactors. Stable containment of T in Li-loading rods during HTGR operation is a critical issue. This study investigates this for an irradiation test to examine T-containment performance in Li-loading rods and develops an analytical model of evaluating the amount of T outflow to a He coolant. The hydrogen absorption characteristics, including the deterioration of the hydrogen absorption speed after Zr has sufficiently absorbed the hydrogen, is experimentally measured assuming an HTGR setting. We present an analytical model of evaluating the T outflow from a Li rod and, on the basis of this model, estimate the total T outflow, assuming the presence of a gas-turbine high-temperature reactor of 300 MWe with a nominal capacity and a high-temperature engineering test reactor. It is demonstrated that, by loading a sufficient amount of Zr into the Li rod, the T outflow can be suppressed to less than a small percent of the total T produced during 360 days of reactor operation.

The estimation of tritium permeation rate through the plasma facing wall into coolant is required to discuss tritium balance in a D-T fusion plant, to design tritium recovery system and to perform safety assessments. In this work, tritium permeation rates in the blanket first wall and the divertor were estimated by numerical analysis for simplified multi-layer structures with considering the temperature distribution in recent JA DEMO condition. The permeation rate in the blanket first wall, which was a double layer consisting of tungsten and F82H, was estimated to be 0.69 g/day. The permeation rate in the divertor, which was a triple layer consisting of tungsten, copper and copper alloy or F82H, was estimated to be 0.013 g/day. When the permeation rate in tritium breeding region in the blanket can be reduced by three orders of magnitude due to a permeation barrier, total tritium permeation rate in the blanket and the divertor was estimated to be 0.71 g/day.

The shutdown dose rate at Japan’s demonstration nuclear fusion (DEMO) plant was analyzed using MCNP-5 and DCHAIN-SP2001 to assess its significance in the development of a maintenance program. It was assumed that the blanket segments were integrated with a shielding plug (SP) and replaced through vertical upper ports, whereas divertor cassettes, also integrated with the SP, were replaced through bottom ports. To minimize the dose rate, remote handling equipment should approach from behind the SP and be fixed with an attachment to it. The estimated dose rates during manifold cutting and rewelding in the maintenance ports were as low as 0.01 Gy/h for the blanket segment and 0.1 Gy/h for the divertor cassette. When the outboard blanket segments and divertor cassettes were removed along with the SP, a spatial dose rate of 100 Gy/h was found at the maintenance ports. The spatial dose rate of the in-vacuum vessel at ITER during maintenance using remote handling equipment is limited to 250 Gy/h. The study confirmed that a maintenance scheme involving pipe cutting or rewelding would be applicable to ITER at the estimated spatial dose rates.

Scrape Off Layer (SOL) plasma circumnavigates the torus along magnetic field lines, and some part of it encounters a leading edge of the first wall (FW). This leads the concentration of the plasma energy in a small area, causing a hot spot. In a DEMO reactor that deals with a few hundreds of MW of plasma energy in the SOL, the leading edge problem is a critical issue in the FW design. For the design purposes in DEMO and power reactors, a new analysis method is introduced, where a flux tube is poloidally divided into the Apple Peel Like Elements (APPLEs) and each APPLE is enclosed by adjacent 4 magnetic field lines. Considering the contact area of the APPLEs on the FW and the radial energy transport between the APPLEs, FW heat load profile can be analyzed. The result of the case that the fusion power is 1.5 GW, and the major radius is 8.2 m, shows that the heat load is peaked to about 1 MW/m 2 near the inboard midplane and the baffle plate. This method is expected to be useful for the DEMO design.

Preliminary assessment for two options of remote maintenance, the banana shape segment transport using all vertical maintenance ports (BSAV) and saddle-shaped segment transport using limited number of horizontal maintenance ports (SSLH), is carried out to establish a feasible maintenance scheme on the DEMO design, focusing the plasma vertical stability, in-vessel transferring mechanism of blanket segment and divertor, and pipe connection. Analysis on the plasma vertical stability indicates that the BSAV scheme with double-loop shell is the most effective on plasma vertical stability. Design and preliminary assessment of in-vessel transferring mechanism and pipe connection were investigated. Although the BSAV has more pipe connection, shutdown dose rate at connection operation area is lower than that on the SSLH scheme. Therefore, the BSAV maintenance scheme would be more feasible scheme for DEMO.

The Joint Special Design Team for Fusion DEMO was organized in 2015 to enhance Japan’s DEMO design activity and coordinate relevant research and development (R&D) toward DEMO. This paper presents t...

Prediction and likelihood identification of high-beta disruption in JT-60U has been discussed by means of feature extraction based on sparse modeling. In disruption prediction studies using machine learning, the selection of input parameters is an essential issue. A disruption predictor has been developed by using a linear support vector machine with input parameters selected through an exhaustive search, which is one idea of sparse modeling. The investigated dataset includes not only global plasma parameters but also local parameters such as ion temperature and plasma rotation. As a result of the exhaustive search, five physical parameters, i.e., normalized beta βN, plasma elongation κ, ion temperature Ti and magnetic shear s at the q = 2 rational surface, have been extracted as key parameters of high-beta disruption. The boundary between the disruptive and the non-disruptive zones in multidimensional space has been defined as the power law expression with these key parameters. Consequently, the disruption likelihood has been quantified in terms of probability based on this boundary expression. Careful deliberation of the expression of the disruption likelihood, which is derived with machine learning, could lead to the elucidation of the underlying physics behind disruptions.

This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

Generally, DEMO requires larger toroidal field (TF) coils than ITER, resulting in one of the major difficulties, the tolerance in TF coil fabrication. This paper presents the possible solutions based on the design study on Japan’s DEMO (JA DEMO). It was confirmed that, in the case of adopting a mitigated tolerance by a factor of 2.5–5 compared with that of ITER, the resulting error field of TF coils is correctable to an acceptable level in terms of locked mode avoidance. In addition, the design of the error field correction coil (EFCC) on JA DEMO was investigated.

The blanket modules in DEMO are made of reduced-activation ferritic martensitic (RAFM) steel F82H. This material is ferromagnetic and it drags the magnetic field lines into the first wall (FW). Because of this, the heat load by the plasma heat flux, which goes along the magnetic field line will become higher. In this research, the first analysis of such effect has been done. The extra magnetic field Bm made by RAFM wall becomes higher at inner midplane, and the heat load at the module front surface becomes 1.3 MW/m2 to 5 MW/m2. Additionally, near the toroidal gaps, BM becomes high. Thus, at the top of the FW, magnetic field lines are dragged into the toroidal gaps directly because, the magnetic flux surface is not closed. This makes high (about 10MW/m2) heat load concentration at the moduel edge. The effect of the NBI port is also analyzed. Also near the port, Bm becomes high and the orbit of the magnetic field lines are changed. The effect of this doesn't occur near the port, but far region such as inner midplane or top of the FW. The heat load becomes 6 MW/m2 at inner midplane. These results indicate that the effect of RAFM steel on the FW heat load is not negligible, and more detailed analysis is necessary.

The blanket modules in DEMO are made of reduced-activation ferritic martensitic (RAFM) steel F82H. This material is ferromagnetic and it drags the magnetic field lines into the first wall (FW). Because of this, the heat load by the plasma heat flux, which goes along the magnetic field line will become higher. In this research, the first analysis of such effect has been done. The extra magnetic field Bm made by RAFM wall becomes higher at inner midplane, and the heat load at the module front surface becomes 1.3 MW/m2 to 5 MW/m2. Additionally, near the toroidal gaps, BM becomes high. Thus, at the top of the FW, magnetic field lines are dragged into the toroidal gaps directly because, the magnetic flux surface is not closed. This makes high (about 10MW/m2) heat load concentration at the moduel edge. The effect of the NBI port is also analyzed. Also near the port, Bm becomes high and the orbit of the magnetic field lines are changed. The effect of this doesn't occur near the port, but far region such as inner midplane or top of the FW. The heat load becomes 6 MW/m2 at inner midplane. These results indicate that the effect of RAFM steel on the FW heat load is not negligible, and more detailed analysis is necessary.

Plasma heat flux in the peripheral plasma reaches the first wall (FW) along a magnetic field line, and it sometimes causes several MW/m2 orders of magnitude high heat flux concentration at narrow region, such as the edge of the blanket module. Thus, to assess and to reduce the heat load is key issue in the DEMO design activity. In this research, a new heat load analysis code is introduced based on the e-folding model. In this code, the decay length λ is changed depending on wall connection length of magnetic field line (defined as the length of magnetic field line from the wall to the wall), and parallel heat flux q / / is calculated in each flux tube. This code can calculate the FW heat load simulating actual blanket module shapes. The 0.23 MW/m2 peak heat load at inner midplane in the case of the ideal FW (without gaps between blanket modules) is increased to 22 MW/m2 at toroidal module edge in the case of box shaped module. To shadow the edge and reduce such peak heat load, toroidal, and poloidal roof shaping is applied. Required roof height is analyzed from this code calculation. After shaping, peak heat load is reduced to 1.2 MW/m2. This value is under the allowable value 1.5 MW/m2, and in this case, surface temperature is also less than allowable temperature.

Plasma heat flux in the peripheral plasma reaches the first wall (FW) along a magnetic field line, and it sometimes causes several MW/m2 orders of magnitude high heat flux concentration at narrow region, such as the edge of the blanket module. Thus, to assess and to reduce the heat load is key issue in the DEMO design activity. In this research, a new heat load analysis code is introduced based on the e-folding model. In this code, the decay length λ is changed depending on wall connection length of magnetic field line (defined as the length of magnetic field line from the wall to the wall), and parallel heat flux q// is calculated in each flux tube. This code can calculate the FW heat load simulating actual blanket module shapes. The 0.23 MW/m2 peak heat load at inner midplane in the case of the ideal FW (without gaps between blanket modules) is increased to 22 MW/m2 at toroidal module edge in the case of box shaped module. To shadow the edge and reduce such peak heat load, toroidal, and poloidal roof shaping is applied. Required roof height is analyzed from this code calculation. After shaping, peak heat load is reduced to 1.2 MW/m2. This value is under the allowable value 1.5 MW/m2, and in this case, surface temperature is also less than allowable temperature.

Vertical instability of an elongated plasma and its stabilized concepts by in-vessel components and vacuum vessel (VV) design have been studied intensely in JA and EU Broader Approach, DEMO Design Activity. The vertical stabilization of the plasma represents one of the key issues for EU and JA DEMO, due to the large distance of the active control coils for the presence of thick breeding blanket system. A feasible DEMO reactor that maintains plasma vertical stability was proposed from an engineering viewpoint. The vertical stability performances are acceptable, without considering a passive stabilizer, if a maximum elongation of κ95 = 1.6 is chosen. For the higher-elongated plasmas (κ95 > 1.70), additional inboard passive stabilizer is effective.

Fusion energy has abundant resources, does not exhibit nuclear runaway, and it has zero-carbon emissions. Long-term strategies for fusion energy development will become critically important to promote future DEMOnstration Fusion Reactor (DEMO) projects by another largescale investment and gain social acceptance. This paper assessed the potential contribution of fusion power generation to low-carbon development, which is prescribed in the Paris Agreement, under the combination of different uncertainties of future socioeconomic development, probability of the 2 °C target, and development of commercial fusion power plants. Global negative CO2 emission in 2100 by drastic decarbonization of energy systems was required to achieve the 2 °C target, and fusion power plants were expected to be installed in the latter half of the 21st century mainly in countries with limited potentials of zero-emission energy sources, such as Japan, Korea, and Turkey, for cost-efficient climate change mitigation. If inexpensive power plants could be developed by enhanced R&D and advanced design in DEMO projects or if the establishment of fission plants in the future is low, fusion power generation will also be deployed in the EU28, India, and China. Further cost reduction by innovative design and alternative concepts are also essential to diffuse fusion plants in zero-emission resource-rich countries.

Goals of Japan’s fusion demonstration (DEMO) reactor are to demonstrate (1) steady and stable electric power generation in a power plant scale, (2) self-sufficient production of fuel (tritium), and (3) reasonable availability using a remote maintenance scheme anticipated in a commercial plant. Main design parameters of JA DEMO are a plasma major radius of 8.5 m, fusion output of 1.5-2 GW, magnetic field on the plasma axis of 5.94 T. The superconducting coil system of the reactor consists of a central solenoid (CS), 7 poloidal field (PF) coils and 16 toroidal field (TF) coils. Regarding CS and PF coils, superconducting coil technology on DEMO is basically the same as that on the world largest fusion experimental reactor called ITER. In contrast, TF coils have a technology gap on magnetic energy and the resulting stress between ITER and DEMO due to their size and magnetic field. In particular, the necessity of higher design stress is critical for TF coils, requiring the development of high strength cryogenic steels surpassing the existing ones. The fundamental design strategy to mitigate tolerances in TF coil fabrication is also presented.

This paper describes an emergency fast discharge scheme of toroidal field (TF) coils of fusion demo reactors for reduction of induced voltage applied to turn insulations of conductors. TF coils are divided into serially connected segments that are electrically isolated from each other and only the coil segment having a failed coil is rapidly discharged. It was found from a circuit current analysis that this discharge scheme enables to reduce the insulation voltage with a factor of ~0.6 or less and which would contribute to ensure reliability of the turn insulations.

The Japanese government has set a power sector goal for photovoltaic (PV) power usage to reach 53 million kW by 2030. To achieve the large-scale introduction of PV, a large storage capacity, in the form of pumped storage systems or batteries, is needed to store surplus electricity from PV plants. At the same time, in the transport sector, the electric vehicle (EV) is being developed as an environmentally friendly vehicle. To promote the diffusion of EVs, we need infrastructure that can charge EVs in a short time; a battery-switch station is one solution to this problem. This study 1) proposes the use of the station batteries as a countermeasure for surplus electricity from PVs and 2) conducts two relevant analyses. In the first analysis, we calculate the marginal value of a battery and an inverter using the Optimal Generation Mix Model (OPTIGEN). In the second analysis, we set the annual lease fee for the inverter and the battery, and calculate the optimum installed capacity of these devices. The results showed that the marginal value of the inverter/battery decreases with increasing inverter/battery capacity. The optimum installed capacity of the inverter/battery is derived from the intersection of the line of marginal value with the line of the annual lease fee. The stations gain an additional profit by leasing batteries to utilities.

We propose a “multi-layer (ML)” 1D model for understanding behaviors of partially detached divertor plasmas. The basic idea is to analyze interaction among the adjacent flux tubes for the attached and detached states in the divertor region. Cross-field transport terms in both SOL and divertor regions are considered in this model. We show some preliminary simulations which indicate necessity of the ML 1D model. According to the simulations, cross-field energy flow in the divertor region can affect stability of a detachment front of partially detached plasmas, i.e. such flow can decrease the front speed towards the X point.

The numerical analysis of the demonstration fusion reactor Demo-CREST has been carried out; this analysis focuses on impurity seeding. Several design activities for DEMO have been carried out; however, its detailed divertor plasma analysis remains to be carried out. Therefore, in this study, we discuss the possibility of neon puffing in Demo-CREST to decrease the power load to the divertor plate by using the B2-EIRENE code. It has been shown that the radiation power loss by neon increases with upstream plasma density and that the peak power load to the divertor plate comes close to the allowable level by using the preliminary divertor configuration (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

A Plug-in hybrid electric vehicle (PHEV) is a kind of hybrid electric vehicles, which can be charged from the electric grid and can provide several ten kilometers of all-electric driving range. Due to recent introduction of hybrid electric vehicles and development of advanced batteries such as the lithium-ion battery, PHEV is expected as a good candidate to reduce oil consumption and CO2 emission in Japan. The purpose of this study is to analysis the economic advantage and ecological effects of PHEV in Japan considering utilization patterns and electricity supply mixture. The results are as follows: (1) Considering the utilization patterns of passenger vehicles in Japan, PHEV with 35 kilometers of all-electric driving range could reduce oil consumption by 58 % compared with HEV. (2) When the total of vehicle cost and discounted fuel cost is minimized, the PHEV will become widespread after 2010, and it will account for 80% of all passenger vehicle market after 2030. (3)In the case of (2), it will reduce CO2 emission by about 57 million t-CO2 in 2050 compared with that of all gasoline vehicles in 2050.