A. Pizzuto

Design of the JT-60SA Superconducting Toroidal Field Magnet

The JT-60SA is a fusion experiment designed to contribute to the early realization of fusion energy, by providing support to the operation of ITER, by addressing key physics issues for ITER and DEMO and by investigating how best to optimize the operation of the next fusion power plants that will be built after ITER. It is a combined project of the JA-EU Satellite Tokamak Program under the Broader Approach (BA) Program and JAEAs Program for National Use, and it is to be built in Naka, Japan, using the infrastructure of the existing JT-60U experiment. This paper describes in detail the design of the JT-60SA Toroidal Field magnet and shows the strong points of each foreseen solution. Additional information about manufacturing procedures is given and technological issues are reported and critically analysed.

Conceptual Design of Superconducting Magnet System for JT-60SA

The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) has been decided by both parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of 18 toroidal field (TF) coils, a Central Solenoid (CS) with four modules, seven Equilibrium Field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure. The seven EF coils are attached to the TF coil cases through supports which include flexible plates allowing radial displacements. The CS modules operate at high field and use Nb3 Sn type superconductor. The TF coils and EF coils use NbTi superconductor. The magnet system has a large heat load from nuclear heating from DD fusion and large AC loss. This paper describes the technical requirements, the operational interface and the outline of conceptual design of the superconducting magnet system for JT-60SA.

THE FRASCATI TOKAMAK UPGRADE: ELECTRICAL MECHANICAL, AND THERMAL FEATURES

The report presents a short overview of the work carried out during the assembly phase of the FTU machine and its major FTU subsystems: vacuum and gas injection, power supplies, additional heating (LHRH). The commissioning work, the first plasma shots with the machine at liquid nitrogen temperature, the major engineering tests, and the study and the development of some additional subsystems (pellet injectors, ECRH, IBW, glow discharge, pumped limiter, and toroidal limiter) are described. Finally, the near-term experimental program is outlined.

A New Design for JT-60SA Toroidal Field Coils Conductor and Joints

The upgrade of JT-60U to JT-60 Super Advanced (JT-60SA), a fully superconducting tokamak, will be performed in the framework of the Broader Approach (BA) agreement between Europe (EU) and Japan. In particular, the Toroidal Field (TF) system, which includes 18 coils, is foreseen to be procured by France, Italy and Germany. This work covers activities from design and manufacturing to shipping to Japan. The present paper is mainly devoted to the analyses that lead to the conductor design and to the technical specifications of the joints for the JT-60SA TF coils. The conductor geometry is described, which is derived from Cable-In-Conduit concept and adapted to the actual JT-60SA tokamak operating conditions, principally the ITER-like scenario. The reported simulations and calculations are particularly dealing with the stability analysis and the power deposition during normal and off-normal conditions (AC losses, nuclear heating). The final conductor solution was selected through a trade-off between scientific approach and industrial technical orientation. Besides, the TF system connections layout is shown, derived from the industrially assessed twin-box concept, together with the associated thermo-hydraulic calculations ensuring a proper temperature margin.

Neutronic analysis of FAST

As part of the Fusion Advanced Studies Torus (FAST) project, a neutronic analysis has been performed, aimed to design optimization and radiological safety assessment. The neutron emissivity source foreseen for various FAST scenarios has been calculated and used as input for Monte Carlo calculations. The shielding analysis and nuclear heating calculations have been carried out with MCNP5 using a detailed 3-D model of the machine. The energy and spatial distributions of neutron fluxes have been used to perform activation analysis by means of the FISPACT code for safety assessment. The implications of the results on the design of the machine and on safety issues are presented and discussed.

A Reliable Technology to Manufacture the ITER Inner Vertical Target

Abstract ENEA and Ansaldo Nucleare S.p.A. (EA) have been deeply involved in the European International Thermonuclear Experimental Reactor (ITER) research and development activities for the manufacturing of high-heat-flux plasma-facing components and in particular for the inner vertical target (IVT) of the ITER divertor. These components have to be manufactured by using both armor and structural materials whose properties are defined by ITER. Their physical properties prevent the use of standard joining techniques. The reference armor materials are tungsten and carbon/carbon fiber composite (CFC), and for the cooling pipe, the materials are a copper alloy (CuCrZr). During the last years EA have jointly manufactured several actively cooled mock-ups and prototypical components of different lengths, geometries, and materials by using innovative processes: hot radial pressing (HRP) and prebrazed casting (PBC). The HRP technique is based on radial diffusion bonding between the cooling tube and the armor material obtained by pressurizing only the cooling tube while the joining zone is kept in vacuum and at the required bonding temperature. The heating is obtained by a standard air furnace. The PBC process is used for the CFC armor tile preparation. A soft copper interlayer between the tube and armor is necessary to mitigate the stress at the joint interface, and it is obtained by pure copper casting that follows the activation of the CFC surface by a standard brazing alloy. The optimization of the processes started from the successful manufacturing of both tungsten and CFC small-scale mock-ups and successful testing under the worst ITER operating condition (20 MW/m2) through the achievement of record performances obtained from a medium-scale vertical target CFC and tungsten armored mock-up: After ITER-relevant heat flux fatigue testing (20 MW/m2 for 2000 cycles, CFC part, and 15 MW/m2 for 2000 cycles, tungsten part), it reached a critical heat flux of 35 MW/m2 at ITER-relevant thermal-hydraulic conditions. Based on these results EA participated in the European program for the qualification and manufacturing of the divertor IVT, according to the Fusion for Energy (F4E) specifications. A divertor IVT prototype (400-mm total length) with three plasma-facing-component units was successfully tested at ITER-relevant thermal heat fluxes (20 MW/m2 for 3000 cycles, CFC part, and 15 MW/m2 for 3000 cycles, tungsten part). Now, EA are ready to face the challenge of the ITER IVT production, transferring to an industrial production line the experience gained in the development, optimization, and qualification of the PBC and HRP processes.

Fabrication of mock-up with Be armour tiles diffusion bonded to the CuCrZr heat sink

The aim of this work is the manufacture of high heat flux mock-ups with Be armour tiles on a CuCrZr heat sink for fabricating the beryllium section of the divertor vertical target (DVT) in the ITER reactor. Diffusion bonding between the CuCrZr bar and the beryllium tiles was obtained by inserting an aluminium interlayer to accommodate surface irregularities as well as to provide a compliant layer for accommodating thermal mismatches during both manufacturing and operation and cycles.

The tubular separate first wall for ITER-EDA

Abstract The first wall (FW) is one of the most loaded plasma-facing components; the heat flux is such that the thermal stresses are the most important design concern. In addition, the FW must resist the eddy current induced during plasma disruption and the high pressure of the coolant and should maintain its properties under a fast neutron flux (dose up to 3 MW m−2). The tubular solution is the most suitable to cope with the thermal stresses; the use of a double wall reduces the risk of leaks inside the vacuum vessel by avoiding the growth of cracks through both walls: the soft brazing in between the walls stops the growth of cracks from one tube to the other. The eddy currents induced in the tubes are low and the halo current flowing poloidally in the tubes exerts a radial pressure which is supported by the blanket box via supporting points provided in between the FW and the blanket. The tubes are protected with a coating of beryllium or boron carbide against the radiation heat load during disruption, and with a coating of copper against runaway electrons. Fins attached to the tubes are provided to cope with the change in the toroidal width of the FW along the poloidal direction. The fins are also protected by coatings. The tubes can be made of steel to resist a heat flux of up to 1 MW m−2. For higher heat loads, copper or vanadium can be used. The tubular FW can be replaced independently of the blanket. The thermohydraulic, electromagnetic and dynamic analyses confirm the viability of the solution proposed.

ENGINEERING FEASIBILITY STUDY OF THE OMITRON DEVICE

The overall OMITRON design has been recently revisited and subjected to a feasibility study jointly by the ENEA Fusion Department and the CITIF Consortium (firms FIAT, ANSALDO and ABB). As a result, it has been confirmed that the concept is sound and viable solutions exist for all the main engineering aspects of the project.

Feasibility study of a toroidal limiter for the FTU machine

The Frascati Tokamak Upgrade (FTU) is now operating in the ohmic phase (I/sub p/ up to 1 MA B/sub t/=6 T) with a poloidal limiter. The large heat flux expected during the 8-MW, 8-GHz LH radio frequency heating experimental phase suggested the design and verification of the feasibility of a toroidal limiter. The analyses were carried out on both stainless steel and graphite. The results obtained demonstrate that the solution providing graphite tiles is much better in terms of mechanical performance and reliability.<<ETX>>