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日本における核融合原型炉の概念設計では、プラントの設計活動と並行して保守保全シナリオの策定を進めている。特に、核融合炉の炉内機器は重水素と三重水素との核融合反応で発生する高エネルギー中性子（14 MeV）により高い中性子損傷を受けるため、数年毎の定期的な交換が必要になる。さらに、高線量環境の中で大型機器を遠隔制御で交換することも求められる。原型炉での炉内機器の設計方針は、所定の運転期間（中性子損傷量）までは機能を担保するように設計し、その期間を超えて健全性を確保するため定期交換を行うという考え方である。他方、高エネルギー中性子が発生する核融合原型炉では、冷却水の放射化に伴い主冷却系などが広範囲にわたって高線量率になることにも留意する必要があり、プラント主要機器の放射線防護が保全の重要な課題である。本記事では、核融合原型炉の保守保全技術の概念検討の状況と課題について記述する。

The deuterium (D) permeation behavior for 1 displacement per atom Fe2+ damaged tungsten (W) was studied by the gas-driven permeation method and compared with undamaged W. The results of thermal desorption spectroscopy showed that dislocation loops and voids were formed in damaged W. It was found that the D permeation behavior in W was affected by irradiation defects. The effective diffusivity and permeability in the damaged W were lower than that in undamaged W. However, the difference in effective diffusivity and permeability between the undamaged sample and the damaged sample was reduced with increasing the heating temperature. Under 965 K, which was enough for D detrapping from voids, the permeability for damaged W was consistent with that for undamaged W.

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%).

Tritium permeation through structure materials in fusion blanket systems is a critical issue from the perspectives of fuel loss and radiological hazard. In the previous studies, detailed hydrogen isotope permeation behaviors in reduced activation ferritic/martensitic steels have been investigated; however, oxidation of the steel surface is expected under an actual DEMO reactor condition, and then the tritium permeation behavior will be changed. In this study, deuterium permeation through the steels heat-treated under simulated environment conditions has been investigated for more precise predictions of tritium loss at DEMO reactor blankets. Reduced activation ferritic/martensitic steel F82H substrates were heat-treated in helium gas flow containing 1 vol% hydrogen at 300, 400 and 500 °C for 100 and 200 h to simulate a solid breeder DEMO reactor blanket condition. After surface observation and analysis for the heat-treated samples, gas-driven deuterium permeation measurements were performed. An iron oxide layer was formed on the sample surface, and the thickness of the layer was 50 nm‒12 μm. The oxide layer on the sample surface heat-treated at 500 °C for 100 h decreased deuterium permeation by a factor of 5. After the permeation tests, dissipation of the oxide layers was confirmed.

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.

The International Energy Agency (IEA) has been promoting the IEA Environment, Safety and Economic Aspects of Fusion Power program for many years. Among the tasks of this program, one task in particular deals with radioactive waste management in order to analyze the issue of the final destination of fusion activated and tritiated materials after their use in a fusion power reactor. A collaborative study on these aspects has been carried out in recent years. An optimized waste management strategy is proposed, with the goals of avoiding underground disposal as much as possible, maximizing recycling of activated materials within the nuclear industry, and/or clearance and release to commercial markets if materials contain only slight traces of radioactivity. Some technological problems and recent research advances in this field are summarized.

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

幅広いアプローチ活動だより（77）では、第23回幅広いアプローチ(BA)運営委員会の報告、ITER遠隔実験センターでの遠隔実験実証およびサテライト・トカマク（JT-60SA）計画の進捗に関して報告している。 BA運営委員会では、2018年12月5日にフランス・グルノーブルにおいて開催され、日欧の委員、専門家、事業長および事業チーム員の計30名が参加し、各事業の進展を確認するとともに2019年作業計画が承認された。 ITER遠隔実験センターでの遠隔実験の実証では、青森県の六ヶ所核融合研究所から、約10000キロ離れた欧州のトカマク装置WESTとの遠隔実験を、2018年11月28日に始めて実証した。今回の遠隔実験の成功は、実験炉ITERへの遠隔実験の実現に向けて、大きなステップとなった。来年度には欧州のトカマク装置であるJET装置との試験を予定している。 JT-60SA計画の進展に関しては、超電導コイルを極低温(4K)状態に維持するために設置するサーマルシールドについて報告がされた。当該シールドは80Kのヘリウムガスを循環させ冷却することで、超伝導コイルへの熱侵入が抑えられる。狭い空間に設置されるサーマルシールドは、運転に関連する変位や地震による変位に対応できる空間確保が必要であり、複数の基準点をレーザートラッカーで測定しながら、現場合わせして製作したカプラーを用いて，斜めポートサーマルシールドに対して要求される取付け精度（±10mm）を実現した。これまでに、全ての斜め及び水平ポートのサーマルシールド設置が完了した。 これら、BA活動の主要な出来事を国内コミュニティーに対して情報発信する。

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.

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

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

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

In the conceptual design activity of advanced fusion neutron source A-FNS, a variety of test modules for fusion DEMO reactor are planned. In these modules, progresses of design activities on Blanket Nuclear Property Test Module (BNPTM) and Diagnostic and Control Device Test Module (DCDTM) were reported. The BNPTM is a module in order to evaluate accuracies of nuclear analyses of the DEMO blanket such as tritium production rate. The influences of the cooling water and test cell wall were evaluated in order to decide its design. The DCDTM is one to achieve irradiation data of functional materials on diagnostic and control devices such as mirror and window. It was clarified from its nuclear analysis that the neutron fluence obtained in the DCDTM was enough to investigate the accumulated irradiation effects on the materials. In addition to the irradiation tests on the fusion reactor materials, various neutron applications are planned at A-FNS. One of the applications is medical isotope production. We clarified that enough amount of the medical isotope molybdenum-99 could be produced in comparison with the demand in Japan by using Radio-Isotope Production Module (RIPM). The RIPM affects little influence on the fusion reactor material irradiation tests., 第14回核融合核技術国際シンポジウム(ISFNT-14)

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

In the conceptual design activity of advanced fusion neutron source A-FNS, a variety of test modules for fusion DEMO reactor are planned. In these modules, progresses of design activities on Blanket Nuclear Property Test Module (BNPTM) and Diagnostic and Control Device Test Module (DCDTM) were reported. The BNPTM is a module in order to evaluate accuracies of nuclear analyses of the DEMO blanket such as tritium production rate. The influences of the cooling water and test cell wall were evaluated in order to decide its design. The DCDTM is one to achieve irradiation data of functional materials on diagnostic and control devices such as mirror and window. It was clarified from its nuclear analysis that the neutron fluence obtained in the DCDTM was enough to investigate the accumulated irradiation effects on the materials. In addition to the irradiation tests on the fusion reactor materials, various neutron applications are planned at A-FNS. One of the applications is medical isotope production. We clarified that enough amount of the medical isotope molybdenum-99 could be produced in comparison with the demand in Japan by using Radio-Isotope Production Module (RIPM). The RIPM affects little influence on the fusion reactor material irradiation tests.

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.

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.

合同特別チームでは、ITERの1.8倍程度のダイバータ熱処理パラメータ（Psep/R~30MW/m）を目標として原型炉のダイバータ設計を進めている。中性子照射のため使用寿命が最も制限されると思われる銅合金配管は、高熱負荷ターゲット部のみで使用し1-2年での交換を考えている。その遠隔保守と真空容器への中性子遮蔽に適したカセット設計を進めている。また、10MWm-2以上の熱負荷を受けた際のモノブロックや冷却配管の温度や応力の弾塑性解析を進めている。本発表ではダイバータ設計の現状と課題を発表する。, 第12回核融合エネルギー連合講演会

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 the 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 divertor (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 irradiation dose are comparable. For the JA DEMO, two different water-cooling pipes, i.e. CuCrZr and F82H steel, are proposed. For the EU DEMO1, the heat sink consists of all Cu-alloy pipe, and the divertor size is reduced with replacing the baffles by the breeding blankets. Choices of the heat sink components have been developed appropriate to the high irradiation dose 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 13the International Symposium on Fusion Nuclear Technology(ISFNT-13)

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.

Power handling and the divertor design has been investigated in steady-state Japan DEMO concept of 1.5 GW-level fusion power (Pfus) and the major radius of Rp = 8.5m. System code survey suggested to increase the plasma elongation (k95) from 1.65 to 1.75, loading to increasing Ip and ne, in order to obtain Pfus larger than 1.5 GW in the impurity seeding scenario, where the radiation power in the main plasma (Pradmain) is 180 MW (Pradmain/Pheat = 0.41, where Pheat= 430 MW), impurity concentration in the main plasma, (nAr/ne)core, of 0.5-0.75%. Design of the divertor size and geometry for the power exhaust parameter (Psep/Rp) of 29 MWm-1 was investigated by the divertor simulation (SONIC), where the outer leg lengths of 1.6 and 2.0 m were compared for a high radiation fraction (Praddiv/Pheat = 0.44). For the shorter leg divertor, the peak qtarget at the attached region (Tediv ~20 eV, Tidiv ~30eV) in the partial detached divertor was ~5 MWm-2 and it was also ~5 MWm-2 mostly due to the surface-recombination in the inner divertor (Tediv = Tidiv ~1eV). The longer leg divertor was preferable to reduce qtarget since the partial detachment extended to the outer flux surface, while the size of the vessel and TFC become larger. It was found that Cu-ally (CuCrZr) cooling pipe is applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom (DPA) rate on Cu-alloy was estimated to be 1-2 per year from neutronics calculation. Coolant rooting for Cu-alloy and RAFM steel pipes and the flow velocities were determined to handle the peak qtarget of 10 MWm-2 level and the total thermal and nuclear heat removal of 300 MW and 118 MW, respectively. Heat transport and thermomechanical analyses of the W-monoblock and Cu-alloy pipes were performed. The maximum temperature of the Cu-ally pipe was 331C at the side surface. Heat flux of 16 MWm-2 is distributed in the major part of the interlayer side, while the maximum heat flux of 25 MWm-2 was localized near the poloidal side surface of the monoblock, which was acceptable level. The physics and engineering results were consistent for an integrated DEMO divertor design, which can handle the peak qtaret of 10 MWm-2 level. The integrated design of the two different water-cooling heat sinks for the divertor with Ldiv = 1.6m was shown., 26th IAEA Fusion Energy Conference

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.