Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokkasho Fusion Institute is reported. In order to advance the Japanese 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 Test Blanket System; tritium handling technology; and information technology infrastructures, including a supercomputer system and a remote experimentation system, are carried out for a Fusion DEMO reactor.

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). Common critical issues are the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling ( P rad tot / P heat ⩾ 0.8). Different exhaust concepts in the main plasma and divertor have been developed for Japanese (JA) and European (EU) DEMOs. JA proposed a conventional closed divertor geometry to challenge large P sep / R p handling of 30–35 MW m −1 in order to maintain the radiation fraction in the main plasma at the ITER-level ( f rad main = P rad main / P heat ∼ 0.4) and higher plasma performance. EU challenged both increasing f rad main to ∼0.65 and handling the ITER-level P sep / R p in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding P sep = 250–300 MW and 150–200 MW, respectively. Both results showed that large divertor radiation fraction ( P rad div / P sep ⩾ 0.8) was required to reduce both peak q target (⩽10 MW m −2 ) and T e,i div . In addition, the JA divertor performance with EU-reference P sep of 150 MW showed benefit of the closed geometry to reduce the peak q target and T e,i div near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130 °C–200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs.

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.

Sparse modeling for a data-driven approach in Plasma Physics Yasuhiko Igarashi 1,2, Tatsuya Yokoyama 1, Yuya Miyoshi3, Ryoji Hiwatari3, Akihiko Isayama4, Go Matsunaga4, Naoyuki Oyama4, Yuichi Ogawa1, Masato Okada1 1 Graduate School of Frontier Science, The University of Tokyo, 2 Japan Science and Technology Agency, PRESTO, 3Rokkasho Fusion Institute, QST, 4Naka Fusion Institute e-mail (speaker): igayasu1219@mns.k.u-tokyo.ac.jp

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 vertical bar B-r(n=1)vertical bar and its time derivative d vertical bar B-r(n=1)vertical bar/dt, 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 vertical bar B-r(n=1)vertical bar/dt, 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.

The conceptual design of Japan's fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in-vessel components to the primary cooling system is evaluated to be 5.7 g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94 kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 11.77TBq/year/loop (318 Ci/year/loop), which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, a confinement concept of tritium release from PHTS is discussed under the condition of ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA.

Recent progress of Japan's DEMO design is presented. The key concept is a steady-state DEMO with a major radius of 8 m class and fusion power of 1.5 GW level, which is proposed based on ITER physics and technology bases, and characterized by operational flexibility from pulse to steady-state operations. Even in a steady-state DEMO, the pulse operation is required for the commissioning of plant systems and also suitable for early demonstration of fusion electricity by moderate plasma performance. Regarding the physics design, divertor plasma simulation clarifies that the lower density to be compatible with detached plasma, which is consistent with the operational density of JA DEMO. Vertical stability evaluation by 3D eddy current and plasma control model shows that plasma elongation of 1.75 is sustainable by applying the double-loop type shells. The development of plasma operation scenario indicates the importance of off-axis ECCD for controlling the internal transport barriers. In addition to physics design, engineering designs are performed in wide area. The divertor cassette design is developed for reducing the fast neutron flux to protect the vacuum vessel and for replacement of the power exhaust units. The breeding blanket concept based on JA ITER-TBM strategy is developed to increase the pressure-tightness of the modules by considering safety assessment of in-box LOCA. On the TF coil design, assessment of the error field indicates that the fabrication tolerance can be mitigated by ~2.5 times as large as ITER's with correction coil current of several 100 kAT/coil. The concept of remote maintenance for the blanket segments is developed such as the stable transfer mechanism in the vertical, radial and toroidal directions. The rad-wastes generated by the maintenance can be disposed of in shallow land burial after 10-year storage. The concept of primary cooling water system is developed for effective use of thermal power removed from not only blanket but also divertor.

Water-cooled pebble-bed (WCPB) blanket, in which beryllium/beryllide in a pebble form is used as neutron multiplier, is one of blanket concepts based on conventional or near-future technology for fusion DEMO. Combination of water, as coolant, and beryllium/beryllide, however, may pose a critical safety problem, i.e. the chemical reactivity of the beryllium/beryllide pebble and hydrogen generation. We present a new phenomenological model of the reaction behavior of the beryllium/beryllide pebble with the steam. The model consists of the equations of the transients of (i) the radius of the unreacted part and (ii) the temperature of the pebble. We have developed a code PSYCHE to numerically solve the model equations. It has been found that the amount of the reaction-produced hydrogen obtained by the numerical simulation agree well with the experimental observations. We also show an application of the code to safety analysis of the transient behaviors of the Be and beryllide Be12Ti pebbles in an in-box LOCA, i.e. loss-of-coolant accident in a blanket box. The model simulation presents the better thermal stability of the Be12Ti pebble, compared to the Be pebble, in the in-box LOCA condition expected in a WCPB DEMO blanket.

Plasma disruption is one of crucial phenomena in a tokamak fusion reactor. To realize nuclear fusion reactor, it is necessary to elucidate and control it. However, its physical mechanism is not clearly identified yet, so there are some studies trying to predict occurrence of disruptions based on experimental data. In this research, we constructed disruption predictor using a support vector machine(SVM) based on the large experimental data in JT-60U and feature extraction by sparse modeling was carried out. The concept of sparse modeling exploits the inherent sparseness that is common to all high-dimensional data and enables us to efficiently extract the maximum amount of information from data. For the sparse modeling, we used exhaustive search with SVM, assuming that the optimal combination of explanatory variables is K-sparse [1]. We have obtained some results showing that feature extraction can contribute to improvement of disruption prediction performance and understanding of the physical background of disruption. As a variable before narrowing down, we chose 17 parameters from physical knowledge. We selected normalized beta and plasma internal inductance because we use results of highbeta experiment. We also selected safety factor 95% of minor radius, and these parameters are obtained from MHD equilibrium calculation. We use not only those parameters’ value, but also time derivative value. In our results, 6 parameters including mode lock amplitude and its time derivative are extracted as the optimal combination of parameters. We will try to specify dangerous parameter area where disruption is likely to occur using sparse modeling.

全状態探索によるデータ駆動アプローチを用いた JT-60Uの高ベータディスラプション予知と物理背景の抽出 Data-driven approach on high-beta disruption in JT-60U using exhaustive search 横山達也 ,三善悠矢 ,日渡良爾 ,諫山明彦 ,松永剛 ,大山直幸 , 五十嵐康彦 ,岡田真人 ,今川直人 ,小川雄一 1 Tatsuya YOKOYAMA, Yuya MIYOSHI, Ryoji HIWATARI, Akihiko ISAYAMA, Go MATSUNAGA, Naoyuki OYAMA, Yasuhiko IGARASHI, Masato OKADA, Naoto IMAGAWA, Yuichi OGAWA 東大新領域 ,QST六ヶ所 ,QST那珂 ,科学技術振興機構さきがけ ,東大工学部 5 Graduate School of Frontier Science, The Univ. of Tokyo, Rokkasho Fusion Institute, QST, Naka Fusion Institute, QST, PRESTO, Faculty of Engineering, The Univ. of Tokyo

Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of P-sep = 205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of Pout = 250 MW and the total radiation fraction at the edge, SOL and divertor (P-rad/P-out = 0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load (q(target)) at the attached region was reduced to similar to 5 MW m(-2) with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak qtarget was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak qtarget of 10 MWm(-2) and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 degrees C. Heat flux of 16 MW m(-2) was distributed in the major part of the coolant pipe. These results were acceptable for the plasma facing and structural materials.

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.

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 (R-p=5.5m) and Demo-CREST (R-p=7.3m), have been proposed in Japan as possible DEMO designs which will fit in the policy

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 similar to 110 days for the net TBR(DT)=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.

We have developed the road traffic simulator to analyze the location of 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 EV running out of electricity 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 EV running out of electricity, and it does not depend on the initial input of the location of the charging ST and the complexity of the road network.

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. (C) 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

This paper discusses the plasma commissioning scenario and the relationship between the initial tritium inventory and the commissioning period for a demonstration reactor concept Demo-CREST. The tritium density ratio (T-ratio) control is applied to keep the high density operation preferable for divertor heat-handling during gradual increase of the fusion power in the commissioning phase. It is found that Demo-CREST can start from zero fusion power operation with T-ratio fnT~0%, in which the divertor heat-handling condition on the SOL density and the radiation power required for divertor heat load less than 10MW/m is similar to that of the ITER steady state operation. An operation route keeping high density by the T-ratio control is also proposed for the commissioning period. This proposed operation route has a consistency with the start-up scenario without the initial tritium loading, and the relationship between the initial tritium inventory and the commissioning period is also evaluated for Demo-CREST

A simple core-SOL-divertor model (CSD model) was developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. In the CSD model, the core plasma model of ITER physics guidelines and the two-point SOL-divertor model are applied. This CSD model is validated by the two dimensional divertor transport code (B2-EIRINE) and by the JT-60U divertor recycling database, and this model is applicable to the low- and high-recycling state of the divertor plasma. The CSD model is applied to the study of the EAST operational space with lower hybrid current drive under various kinds of trade-off for the basic plasma parameters, and the relationship between the operational space and the plasma discharge duration is also discussed.