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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%).

This paper summarizes the evolution of Japanese DEMO design studies in a retrospective manner by highlighting efforts to resolve critical design issues on DEMO. Japan is currently working on the conceptual study of a steady state DEMO (JA DEMO) with a major radius of 8.5 m and fusion power of 1.5-2 GW based on water-cooled solid breeding (WCSB) blanket with PWR water condition (290-325ºC, 15.5 MPa). Such a lower Pfus allows to find realistic design solutions for divertor heat removal. Recognizing that divertor heat removal is one of the most challenging issues on DEMO, divertor design has been carried out in different approaches including numerical divertor plasma simulation, magnetic configurations, heat sink design, etc. It is noteworthy that the latest divertor simulation led to a design window allowing divertor heat removal of the peak heat flux of < 10 MW/m2. Breeding blanket (BB) design has been concentrated on simplification of the internal structure and pressure tightness of BB casing against in-box LOCA (Loss Of Coolant Accident). Due to a large amount of radioactive waste generated in periodic replacement of in-vessel components, downsizing of waste-related facilities came to be regarded as a significant design issue. A possible waste management for reducing temporary waste storage was proposed and its impact on the plant layout was assessed.

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 5.7g-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 94kg/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 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, confinement concept of tritium release from PHTS is discussed under the condition of an 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.

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 |Brn=1| and its time derivative d|Brn=1|/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|Brn=1|/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.

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 Li2TiO3 pebbles 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.

核融合原型炉の3つのミッションの観点から，はじめに，基本となる核融合出力のダイバータ熱処理の観点からの物理・工学設計，核融合出力を可能とする強磁場コイルの概要，NBIシステム設計，主熱輸送系の設計概要について報告する．次に，炉構造の検討概要，遠隔保守方式の概念，暫定的な保守期間の評価について報告する．最後に，燃料システムの基本概念，トリチウム取り扱いに関する課題，ブランケット概念/原型炉TBM等について報告する．, プラズマ・核融合学会第37回年会

核融合原型炉の一次燃料系において、プロトン導電体を用いた水素同位体移送によるダイレクトリサイクルを導入した場合の燃料系のインベントリーと応答への影響を評価した。ダイバータから水素同位体を選択的に気相で排気昇圧する結果、循環時定数が大幅に短縮され、インベントリー低減効果があることがわかった。, 日本原子力学会 2020年秋の大会

高温ガス炉を用いた初期装荷トリチウム製造法を検討している．高温ガス炉用トリチウム(T)製造Liロッドに対しZrを用いたT閉じ込め方法を検討している。炉内に温度分布が存在することを考慮して温度範囲でのZr水素吸収速度を測定し、T閉じ込め性能を評価した。, 日本原子力学会 2020年秋の大会

原型炉プラントの安全上の特徴に基づく合理的な安全確保方針案の策定に向けて、炉内トリチウムインベントリの同定と冷却水喪失事象等の安全解析を通して様々なディスラプションシナリオの影響度を分析した結果を報告する。また、検討した様々なシナリオに対して、放出されるプラズマ熱エネルギーとその到達位置を同定し、保護リミターの設置位置の検討結果も報告する。, 第36回 プラズマ・核融合学会 年会

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 the fundamental concept of DEMO and its key components with main arguments on DEMO design strategy. Superconducting magnet technology on toroidal field coils is based on the ITER scheme where a cable-in-conduit Nb3Sn conductor is inserted in the groove of a radial plate. Development of cryogenic steel with higher strength is a major challenge on the magnet. Divertor study has led to a baseline concept based on water-cooled single-null divertor assuming plasma detachment. Regarding breeding blanket, fundamental design study has been continued with focuses on tritium self-sufficiency, pressure tightness in case of in-box LOCA (loss of coolant accident) and material compatibility. An important finding on tritium permeation to the cooling water is also reported, indicating that the permeation to the cooling water is manageable with existing technology.

Erosion reduction of tungsten (W) divertor target is one of the most important research subjects for the DEMO fusion reactor design, because the divertor target has to sustain large fluence of incident particles, composed mainly of fuel ions and seeded impurities, during year-long operation period. Rate of net erosion and deposition on outer divertor target has been studied by using the integrated SOL/divertor plasma code SONIC and the kinetic full-orbit impurity transport code IMPGYRO. Two background plasmas have been used: one is lower density ni and higher temperature case and the other is higher ni and lower temperature case. Net erosion has been seen in the lower ni case. But in the higher ni case, the net erosion has been almost suppressed due to increased return rate and reduced self-sputtering yield. Following two factors play important roles for net erosion formation: (i) relation between the 1st ionization length of sputtered W atom and the Larmor gyro radius of W+ ion, (ii) balance between the friction force and the thermal force exerted on W ions. DEMO divertor design should take into account these factors to prevent target erosion.

原型炉設計合同特別チームでは、文部科学省の原型炉開発総合戦略タスクフォースの策定したアクションプランに沿って、日本の原型炉概念設計を実施している。原型炉の目標は、数十万kWの安定した電気出力、実用に供し得る稼働率、燃料の自己充足性を満たすトリチウム増殖、を実現することである。本講演では、原型炉概念設計の全体の進捗とともに、目標を達成するために重要な役割を担う増殖ブランケット概念の現状と関連する課題に対する解決策について発表する。, 第36回 プラズマ・核融合学会 年会

文部科学省の核融合科学技術委員会の要請により、2015年6月に量子科学技術研究開発機構六ヶ所核融合研究所に設置された原型炉設計合同特別チームでは、原型炉開発総合戦略タスクフォースの策定したアクションプランに沿って日本の原型炉概念設計を実施中である。産学が連携した総勢101名の専門家で構成されたオールジャパン体制の設計活動を推進するため、大学等との共同研究や関連学協会への委託調査、個別課題を検討するためのワーキンググループ活動に加えて、全体会合等の技術会合を開催して、設計情報の共有や合意の形成を図りながら設計活動を推進している。 原型炉には、(1) 数十万キロワットの安定した電気出力、(2) 実用に供しうる稼働率、(3) 燃料の自給自足の達成、が求められる。特別チームでは、ITERで採用された技術（原型炉本体、ダイバータや超伝導コイル等）を最大限に活かすこと、ITERでは実証されない技術（遠隔保守や増殖ブランケット等）の検討を中心に取り組むこと、原型炉の施設全体を検討すること、を方針に検討を進めた。その結果、これまでの設計検討において、ITERの技術基盤に基づきつつ、産業界の発電プラント技術及び運転経験を取り入れることで、(1) 約64万キロワットの電気出力、(2) 新たな保守方式の考案による稼働率~70%、(3) 増殖ブランケット構造の改良による燃料生産性の向上、に見通しが得られている。 2020年頃に予定されている第１回中間C&Rでは達成目標として、原型炉概念の基本設計と炉心・炉工学への開発要請の提示が求められている。それに向けて、特別チームではC&R項目とアクションプランに対応した構成の「原型炉概念基本設計報告書」、設計を裏付ける「設計根拠集」、アクションプランに沿ったWBSに基づく「設計図書」の整備を進めている。, 第36回プラズマ・核融合学会年会

原型炉概念構築においては，実験炉ITERの技術基盤を有効利用することで，TFコイル，ダイバータ，真空容器等の概念構築を進めている．一方，原型炉のブランケット，遠隔保守等はITERとは異なる概念構築を進めている．それら原型炉の炉内機器，真空容器，超伝導コイル，遠隔保守方式，プラント設備といったシステム設計の進捗を報告すると共に，ITERと比較した原型炉の技術課題を明確化する．また原型炉の各機器概念のシステム統合から議論可能となる原型炉のミッション(電気出力，トリチウム充足性，稼働率)との関係を考察する．, 第36回 プラズマ・核融合学会 年会

原型炉において周辺プラズマの一部は熱流束として磁力線に沿って移動、第一壁へと到達し、第一壁を構成するブランケット筐体の表面熱負荷の要因となる。表面熱負荷は磁力線と筐体の形状に依存するが、原型炉におけるブランケット筐体は強磁性体であるため、磁力線の形状が変化する。その結果熱負荷分布が予期せぬ形になることが懸念される。本研究ではこのような磁性体による影響を加味した筐体表面への熱負荷解析を実施する。, 第36回 プラズマ・核融合学会 年会

Likelihood of high-beta disruption has been discussed from feature extraction using exhaustive search. A support vector machine (SVM) [1] has been used to construct a disruption predictor model. Establishment of prediction and avoidance of disruption is inevitable for realization of a tokamak fusion reactor. Since disruption physical mechanism has not been clearly identified yet, there have been studies trying to predict occurrence of disruptions based on experiment data using machine learning. Here, it should be noted that input parameters for machine learning have been given by empirical presumption in these studies to date. In our previous study [2], the concept called "sparse modeling" was introduced to establish the method to select optimal input parameters. The sparse modeling exploits the inherent sparseness in all high-dimensional data to extract the maximum amount of information from the data [3]. It was shown that the performance of disruption prediction was improved by selecting appropriate input parameters. In the present study, a model of disruption predictor has been constructed based on high-beta plasma experiment data in JT-60U where the beta value was close or above the no-wall beta limit [4]. A linear SVM is used as a two-class classifier here. The dataset consists of 23 candidate plasma parameters, that is, 10 macro plasma parameters (plasma current I_p , normalized beta β_N , poloidal beta β_p, internal inductance l_i, safety factor at 95% poloidal flux ????_95, plasma triangularity δ, plasma elongation κ, amplitude of magnetic perturbation (n = 1) |B_r (n=1)| , the ratio of plasma density to the Greenwald density limit f_GW = n_e(avg) / n_GW, ratio of radiated power to total input power f_rad = P_rad / P_input, time derivative values for seven of macro parameters, and six local parameters (velocity of plasma rotation V_t and its radial gradient dV_t/dρ , ion temperature T_i and its radial gradient dT_i /dρ, normalized radial location of q = 2 rational surface rho/a, magnetic shear s). It is important here to consider not only individual distributions of each parameter but also combinational effect between parameters. Therefore, the sparse modeling method called exhaustive search (ES), which searches all possible combinations of the input parameters, has been used in order to select the optimal combination of input parameters. As a result of ES, several parameters have been extracted as the key parameters of disruption prediction, those are, β_p, q_95, κ, f_GW, and T_i. Then these five parameters are highlighted to express the linear decision boundary of the classifier, in an exponential function. The input of a linear SVM was modified to the logarithms of each parameter and the calculation was carried again. Consequently, exponential expression of the boundary has been obtained as 1 = e^{7.45}*β_p^{5.39}*q_95^{-8.28}*κ^{7.40}*f_GW^{4.5}*T_i^{0.120}. The likelihood of occurrence of disruption can be given by the obtained exponential expression of decision boundary. The expression of likelihood of disruption provides a hint of physical hypothesis and is applicable for design and development of a control system of a fusion reactor. [1] C. Cortes and V. Vapnik, Machine learning, 20, 273, (1995). [2] T. Yokoyama, et al., Fusion Eng. Design, 140, 67-80 (2019). [3] Y. Igarashi, et al., J. Phys. Soc. Jpn, 87, 044802 (2018). [4] G. Matsunaga, et al., Nucl. Fusion 50, 084003 (2010)., 3rd Asia-Pacific Conference on Plasma Physics

Power exhaust scenario for the feasible DEMO plasmas and the divertor design have been studied with a high priority in the steady-state Japanese (JA) DEMO with the fusion power of 1.5 GW-level and the major radius of 8 m-class. The power exhaust concept requires large power handling in the SOL and divertor, i.e. Psep~250 MW, and Psep/Rp~30 MWm-1 corresponds to 1.8 times larger than ITER. Long leg divertor (Ldiv = 1.6 m; 1.6 times longer than ITER) was proposed as a reference design. SONIC simulation demonstrated that the peak heat load on the target (qtarget) was reduced to less than 10 MWm-2 under the partially detachment with large radiation fraction of (Pradsol+Praddiv)/Psep ~0.8. A design concept of the monoblock target and cooling water pipes for the JA DEMO was proposed in 2016 [1]: two different water-cooling units, i.e. 200C, 4MPa pressurized water in CuCrZr pipe and 290C, 15MPa pressurized water in Reduced Activation Ferritic–Martensitic steel (F82H) pipe, are used. The heat exhaust unit with the CuCrZr pipe can be applied near the strike-point (0.8 m) for the high heat load region, while the replacement is expected every 1-2 years due to the maximal irradiation dose on the CuCrZr pipe of 2 displacement-per-atom (DPA). Recently, cassette structure for the DEMO divertor was designed to incorporate the heat exhaust units and coolant pipes. One cassette covers the toroidal area of 7.5, and 3 cassettes are replaced from a lower maintenance port (total 16 ports). The cassette design is consistent with reduction in the fast neutron flux to protect the vacuum vessel and replacement of the inner and outer heat exhaust units of the CuCrZr pipe. The cassette structure consists of F82H, and the total thickness of 25 cm can reduce the fast neutron flux efficiently by arranging two lines of puddles for the pressurized water with the path length ratio of 7:3 for F82H and water, respectively. The water flow (1m/s) in the puddles removes the total nuclear heat of 0.7 MW in one cassette (totally 32 MW for 48 cassettes). Heat transport analyses of the W-monoblock and CuCrZr-pipe was performed in the three-dimensional (3-D) modeling by using the ABAQUS finite element method (FEM) code, considering the monoblock geometry (shaped target surface is used to protect the leading edge) and the heat flux components (radiation power, neutral flux and nuclear heat as well as plasma heat load along the field line) given by SONIC and MCNP-R simulations. Maximum temperature on the W-surface appears near the downstream edge in the plasma-wetted area. The critical operation temperature of 1200C, i.e. W-recrystallization, corresponds to the total peak heat load of 13.5 MWm-2, which is 1.8 times higher than the result simulated by SONIC (7.5 MWm-2). The maximum temperature of the CuCrZr pipe is 351C, and mechanical toughness of the cooling pipe is also near critical against thermal fatigue. Elasto-plastic analysis of the displacement and thermal stress on the W-monoblock and CuCrZr pipe under the higher heat load have been performed, and the results are presented., Third Technical Meeting on Divertor Concepts

Recent progress in the technology of experimentation and measurement makes it possible to obtain a huge amount of high-dimensional data. Needless to say, this trend has forced scientists to change their attitudes toward data. Effective use of high-dimensional data involves a strong framework and a flexible methodology to make the tight connection of information science to the original purpose of data analysis derived from various scientific disciplines. This framework is called data-driven science, which is extending from biology and medical science to physics, including plasma physics [1-3]. We noticed that sparse modeling (SpM) is a key technology of data-driven science. SpM is a generic term for techniques that exploit the inherent sparseness that is common to all high-dimensional data, enabling the efficient extraction of the maximum amount of information from high-dimensional data in real. In concrete terms, the basic notions of SpM are as follows. First, high-dimensional data are sparse and can be essentially expressed by a small number of variables. Second, the number of explanatory variables should be reduced without loss of accuracy. Finally, explanatory variables are selected objectively, and models of target phenomena are constructed automatically. In the field of information science, SpM has recently received much attention. So far, the data-driven approach based on SpM has achieved promising results in various fields such as medical science, astronomy and material science. If this technology can be strengthened by clarifying the common principles that apply in the background of each case, it will lead to innovative developments in all the natural sciences [1,2]. In this talk, to explain the effectivity of SpM in plasma physics, we show our study case applying SpM for prediction of high-beta disruptions, which is a critical phenomenon in a tokamak reactor [3]. 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 SpM, specifically, exhaustively searching the best combinations of variables which maximize the predictor performance. We introduced 17 ~ 23 input parameters into SpM, including time derivatives of plasma parameters. By SpM, we have succeeded to find out several important parameters for predicting plasma disruption, e.g., poloidal beta, safety factor at 95% of poloidal flux, plasma elongation, the plasma density (relative to the Greenwald density limit) and ion temperature at around q=2 surface. We should notice that the combination with these plasma parameters is indispensable and remarkably make it possible to improve the performance of disruption prediction. [1] Igarashi, Y., Nagata, K., Kuwatani, T., Omori, T., Nakanishi-Ohno, Y., & Okada, M. (2016, March). Three levels of data-driven science. In Journal of Physics: Conference Series (Vol. 699, No. 1, p. 012001). IOP Publishing. [2] Igarashi, Y., Ichikawa, H., Nakanishi-Ohno, Y., Takenaka, H., Kawabata, D., Eifuku, S., ... & Okada, M. (2018, June). ES-DoS: Exhaustive search and density-ofstates estimation as a general framework for sparse variable selection. In Journal of Physics: Conference Series (Vol. 1036, No. 1, p. 012001). IOP Publishing. [3] Yokoyama, T., Miyoshi, Y., Hiwatari, R., Isayama, A., Matsunaga, G., Oyama, N., Igarashi, Y., Okada, M. & Ogawa, Y. (2019). Prediction of high-beta disruptions in JT-60U based on sparse modeling using exhaustive search. Fusion Engineering and Design, 140, 67-80., 3rd Asia-Pacific Conference on Plasma Physics

The conceptual design of the breeding blanket with a honeycomb structure has been created with pressure tightness against in-box Loss-of-coolant accident based on a water-cooled solid breeder. In the previous design, the breeding area of the module was divided into 0.1-m-squared cells with rib structure. As a honeycomb structure is higher in pressure tightness than a square prism structure, the area for filling the mixed pebbles breeder of Li2TiO3 pebbles and Be12Ti ones can be enlarged. Then, the overall TBR is improved to increase the packing ratio of the tritium breeding material. In the created blanket, the capabilities of the pressure tightness, tritium breeding and heat removal are studied using interaction analyses of the neutronics, stresses and fluid dynamics analysis. As a result, a rib with the thickness of 0.015 m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. The packing factor of the mixed pebbles breeder increase to 77 % from 68 % by changing the rib structure from a square prism structure to a honeycomb structure. From the 3D neutronics analysis results, the target of the overall TBR (>1.05) is achievable. The cooling system for the created blanket is designed by fluid dynamics analysis based on the PWR water conditions which are the coolant temperature of 290 - 325 ºC and the operation pressure of 15.5 MPa, respectively. In addition, the tritium extraction system in the created blanket is proposed together with the purge gas system which does not clog the holes. The saturated time of the tritium extraction is also estimated to grasp the tritium inventory in the breeding area., 14th International Symposium on Fusion Nuclear Technology (ISFNT-14)

In order to proceed the integrated core plasma scenario modeling of JA DEMO, (i) core transport simulation, (ii) vertical stability and (iii) MHD stability analyses have been performed. On the 1.5-D time-dependent core transport simulation by integrated code TOPICS, the steady-state operation condition with HH98y2 = 1.41, betaN = 3.6, fBS = 0.69 is demonstrated by optimizing the heating scenario, where CDBM transport model is used. It is also demonstrated that off-axis ECCD (30MW, O-mode, 190GHz) has important roles for maintaining the internal transport barriers (ITBs) for steady-state condition and for controlling the fusion power by control of ITB location. On the vertical stability analysis, the ramp-up scenario of high elongated plasma has been developed by using the plasma equilibrium simulator MECS with 3D eddy current effects. The temporal evolutions of the poloidal beta and internal inductance are evaluated using TOPICS. The result indicates that plasma elongation at 95% of poloidal flux of 1.75 is achievable in JA DEMO. Regarding the MHD stability analysis, the beta limit of JA DEMO plasma has been evaluated by using the linear ideal MHD stability code MARG2D. The beta limit without conducting wall is betaN ~ 2.6, while that with conducting wall can be improved to ~3.5 at the wall radius of rW/a = 1.35. Further improvements are observed with decreasing the wall radius, for example betaN ~ 3.9 at rW/a = 1.30., 14th International Symposium on Fusion Nuclear Technology

Japan’s demonstration plant (JA DEMO) missions are defined as follows: (1) steady and stable power generation beyond several hundred megawatts, (2) availability prospect for commercialization, and (3) overall tritium breeding to fulfill self-sufficiency of fuels. To realize those missions, the basic concept of JA DEMO has been developed based on the feasible technology as applied in the ITER design. First, the major radius R=8.5m of JA DEMO enables to accommodate the CS coil large enough for both pulse and steady-state discharge, to bridge the gap between ITER and DEMO. The fusion power Pf=1.5GW of JA DEMO is determined by the heat-handling capability of the divertor investigated by 2D divertor transport code. Because of neutron irradiation by longer operation than ITER, not only W-mono-block/Cu-alloy-pipe (for the strike-point region) but also W-mono-block/F82H-pipe (for the baffle and dome region) are applied for divertor target. This divertor enables longer operation period than about 1 year, and it contributes to increase the plant availability. The vertical maintenance method is applied for blanket replacement, and the replacement maneuver is originated using firm grip method by both up and down supports of the end-effector. The blanket module of honeycomb structure has pressure tightness against the pressurized water condition to avoid the in-vessel loss of coolant accident. The advanced functional materials (Li2TO3 and Be12Ti) developed in the BA activity are applied to avoid hydrogen generation. This honeycomb blanket module achieves tritium breeding ratio TBR=1.07, and its critical R&D issue is manufacture method by hot isostatic pressing (HIP). As for the toroidal field coil (TFC), the design stress is improved from 667MPa of the ITER condition to 800MPa based on the existing cryogenic steel for high pressure gaseous hydrogen, while TFC follows the radial plate winding method. An operation plan of JA DEMO is also prepared to show the strategy to realize the JA DEMO missions. The operation plan is recently applied to discuss commissioning method, and it reveals requirement of the initial tritium loading. Furthermore, to improve plant availability, the operation plan proposes the high core radiation discharge and the second divertor concept without Cu-alloy-pipe. Those basic concept and operation plan of JA DEMO are the starting point to discuss the shift to the DEMO construction phase in Japan., 14th International Symposium on Fusion Nuclear Technology

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 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 effect of this is analyzed. 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 10 MW/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., 14th International Symposium on Fusion Nuclear Technology