噴気ガスの化学組成が火山活動の激しさの指標になる可能性が指摘されてきた。本研究では, 噴気ガスとともに放出される揮発性金属元素の量の変化と, 同位体組成の変動に相関が見られるか, 相関があれば脱ガスによるレイリーモデルにより同位体組成の変動が説明できるか検討する。三宅島の2000年噴火の火山灰試料とラバウルTavurvur火山の火山灰試料について検討する。

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本研究グループが国際熱核融合実験炉-FEATを対象として行った中性子発生量モニタリングシステムの詳細設計結果を基に、モンテカルロ中性子輸送計算(MCNP_-_4C2)を用いたシミュレーションにより、“その場較正”実験手法と期待される較正精度を明らかにした。今後、その場較正結果のプラズマプロファイルや高中性子発生量モニターへのつなぎの影響を評価するとともに、できるだけ効率的に高い較正精度を得るための較正点についての検討を行う予定である。

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The design and development of the International Thermonuclear Experiment Reactor () have been under way since 1992, based on the international agreement of the Engineering Design Activities (EDA). The design of the main machine and the development of key components were successfully completed by July 2001. The technical specifications of the machine are being prepared for the order of its fabrications in the Coordinated Technical Activities (CTA). The construction of is expected to start by 2005. The machine uses superconducting coils to confine and shape the plasma, and the coil system must be reliably operated to perform the plasma experiment. The system accounts for 28% of the direct capital costs and requires a long manufacturing period (6 years and 6 months). The strand for the superconducting conductor is the first procurement component in the . This paper explains the present design and status of the project and its superconducting coils.

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The Project has been significantly developed in the past years in preparation for its construction. The Negotiators have developed a draft Joint Implementation Agreement (JIA), ready for completion following the nomination of the Project’s Director General (DG). The International Team and Participant Teams have continued technical and organizational preparations. The actual construction will be able to start immediately after the international organization will be established, following signature of the JIA. The Project is now strongly supported by all the participants as well as by the scientific community with the final high-level negotiations, focused on siting and the concluding details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. The extent to which they both wish to host the facility is such that large contributions to a broader collaboration among the Parties are also proposed by them. This covers complementary activities to help accelerate fusion development towards a viable power source, as well as may allow the Participants to reach a conclusion on siting.

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The

project ^{1, 2)} was established in November 2006 by the Agreement involving seven Members (China, the European Union including Switzerland, India, Japan, Korea, the Russian Federation and the United States of America). is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. At the core of the facility, the tokamak will confine a plasma heated to temperatures in the region of 1–2×10⁸ K, in which deuterium-tritium fusion reactions will produce up to 500 MW of fusion power for periods ranging from several hundred to several thousand seconds. Extensive progress has been made in the on-site construction, the production of components for the tokamak, plant and auxiliary systems, and in the preparations for on-site installation. Recently, a major update of the baseline schedule and resource estimate has been undertaken. The revised schedule foresees an earliest technically achievable date for First Plasma of December 2025 and a target date for the transition to D/DT operation of late 2035. This report outlines the project management and recent progress of tokamak components manufacturing and on-site construction activities of the facility.

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Fusion power output of is measured by a group of neutron flux monitors combined with a neutron activation system and neutron profile monitors. These systems should be absolutely calibrated by use of DD/DT generators moving inside the vacuum vessel (in-situ calibration). Each neutron monitor has a limited measurement range of emission rate, but the ranges are connected by cross-calibration using the plasma with at least one decade overlapping. The over all dynamic range covered by the group of neutron flux monitors is 10¹⁴ n/sec to 10²¹ n/sec. Effects of vertical/radial movement of plasma on the measurement accuracy were reviewed. It was found that cross-calibration using specially planned jog shots, and a vertical neutron camera is important to minimize the inaccuracy caused by the plasma movement.

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The designs and analyses of the diagnostic port plug for installing diagnostics have been carried out from various aspects in order to advance the port plug concept to the realistic design. Manufacturing processes have been studied for the alternative structure using the rolling plates and ribs, in addition to the reference structure using forging material. Electromagnetic loads onto the upper port plug during a vertical displacement event have been evaluated. It has been shown that maximum moments are about half of the reference ones. The design integration of three diagnostics into one of the upper port plugs has been performed by considering required aspects such as diagnostic functions, neutron shielding, maintenance of diagnostic components and cooling channels.

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Engineering Design Activity (EDA) will be organized in a near future and will perform the necessary R & D and engineering design for six years.

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is a long-pulse tokamak with elongated plasma. The nominal inductive operation produces a D-T fusion power of 500 MW for a burn length of 300-500 s, with the injection of 50 MW of auxiliary power. With non-inductive current drive from the H&CD systems, the burn duration is envisaged to be extended to 3000 s. The term Instrumentation & Control (I&C) includes every thing required to operate the facility. It comprises three vertical tiers; conventional control, interlock system and safety system, and two horizontal layers; central I&C systems and plant system I&C. CODAC (Control, Data Access and Communication) system forms the upper level of the hierarchy, and is the conventional central control system of architecture. CODAC system is responsible for integrating all plant system I&C and enable operation of as a single integrated plant. CODAC system provides overall plant systems coordination, supervision, plant status monitoring, alarm handling, data archiving, plant visualization (HMI) and remote experiment functions. CIS (Central Interlock System) and CSS (Central Safety System) also form the upper level of the hierarchy to supervising and integrating all plant system interlock and safety functions. Plant system I&C forms the lower level of the hierarchy, and provide dedicated plant data acquisition, plant status monitoring, plant control and plant protection functions to perform individual plant system operation under the supervision of central I&C systems.

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Based on the results of the sub- and full-scale trials, the toroidal field (TF) coil and TF coil structure manufacturing procedures were considered. The radial plate (RP) will be manufactured by assembling ten sets of segments using laser-beam welding. The cover plates (CPs) will be manufactured using three different methods, depending on their geometry. For the winding pack (WP), a winding system has been designed that enables measurement of the conductor length with an accuracy of 0.01% for serial production. The assembly procedure and groove types of the narrow-gap tungsten inert gas (TIG) welding for coil structures were determined. Hereafter, technical improvements will be considered, aiming to further optimize manufacturing.

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The superconducting magnet system consists of 18 Toroidal Field (TF) coils, six Central Solenoid (CS) modules, six Poloidal Field (PF) coils and 18 Correction Coils. The Japan Atomic Energy Agency (JAEA), serving as the Japanese Domestic Agency (JADA) in the project, is responsible for the procurement of nine TF coil winding packs (WP), structures for 19 TF coils (including one spare), and assembly of the WP and the coil structures for nine TF coils. JAEA signed the procurement arrangements for the TF conductor in November 2007 and the ones for the nine TF coils and their TF coil structures in November 2008. The manufacture of the TF conductor has already started and its progress is reported in this special issue. In addition, sub- and full-scale trials were performed to achieve compliance for the manufacturing procedure of the TF coil and its structure. These results are also reported in detail in this special issue. Based on these successful results, JAEA is planning to start manufacturing the first TF coil from 2012 to meet the required schedule and complete the 18^th TF coil at the end of 2017 in cooperation with the European Domestic Agency (F4E).

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