Yu. Krasikov

The thermal shields for the ITER magnet system: thermal, structural and assembly aspects

Abstract The thermal shield system is a continuous barrier between the magnet system operating at 4.5 K and the warm tokamak components. It provides a substantial reduction of both total and local thermal loads to the cold structures to achieve the limits required for normal operation of the superconducting magnet system and maximum heat load of the cryogenic plant. This paper describes details of the design of the different types of thermal shield and presents some results of thermal–hydraulic and structural analyses and some aspects of the assembly procedure for the vacuum vessel thermal shield, which is a challenging task considering the rather small dimensional tolerances that have to be obtained on the fully assembled shield.

The ITER thermal shields for the magnet system: specific design, assembly and structural issues☆

Abstract The thermal shield system provides the required reduction of both total and local thermal loads to the cold structures operating at 4.5 K. The most complex thermal shield component is the vacuum vessel (VV) thermal shield (VVTS) located in a narrow gap between the toroidal field coil system and the VV. The paper concentrates on specific thermal shield issues, such as the VVTS overall design including inboard support system and neutral beam ports, details of the VVTS sub-assembly and assembly procedures, section joints, design and efficiency of thermal shield component interfaces, etc. It also presents results of cryovacuum, thermal-hydraulic, seismic and structural analyses.

Selected Design Solutions for the Integration of the CXRS Diagnostic in to ITER Upper Port Plug No. 3

Diagnostic plug for the ITER core charge exchange recombination spectroscopy (core CXRS) is located in the upper port 3. It transfers the light emitted by interaction of plasma ions with the diagnostic neutral beam (DNB). Conceptual design study of the core CXRS port plug has indicated several challenging technical problems: (1) likely too short lifetime of the first mirror, (2) quite contradictory requirements to the first mirror holder, (3) harsh environmental conditions for the “shutter”, that is a movable element protecting the first mirror, (4) a task to combine a sufficient structural integrity and nuclear shielding capability of the plug with a wide enough optical path, (5) excessive electromagnetic loads caused by the halo current and applied at the plug as a whole. This paper describes possible design solutions for the listed technical problems.

The tokamak TSP-AST concept

Abstract The TSP tokamak—Russian fusion project based on very powerful adiabatic heating by compression (900 MJ power supply per pulse)—was entered into exploitation in 1987. The study program with circular cross-section plasma was carried out on the machine from 1988 to 1992. The TSP tokamak program object lost urgency for today. So, modernization of TSP tokamak is planned as creation of new version of machine, which will allow from one hand to use the basic infrastructure of existing machine and unique systems of the pulse power supplies and on another hand to execute greatest possible amount of researches on the fusion basic problems. This modernization was named TSP-AST tokamak (TSP in the mode of advanced or adiabatical spherical tokamak). In the paper the purposes and tasks of the machine, basic parameters of operations regimes and plasma discharge scenarios are submitted. The design version of the basic units of machine (magnet, vacuum vessel) are shown. The calculations results of tensile strength of the machine basic units are given.

Toroidal field coil and central solenoid modernization for Globus-M tokamak

The Globus-M spherical tokamak placed in A.F. Ioffe Institute is in operation since March 1999. The Globus-M tokamak peculiarity is its small aspect ratio plasma properties (R/a=1.5). Major parameters of the Globus-M tokamak are: plasma major radius 0.36 m, plasma minor radius 0.24 m, toroidal magnetic field 0.5 T, plasma current 0.3 MA and pulse length 0.3 s [1]. Manufacturing of toroidal field coils required high accuracy of processing and high quality of contact surfaces. Winding uniformity of the conductor, both between turns, and on layers in the radial direction was required for the central solenoid. The ability to change a toroidal field coil together with providing access to the vacuum vessel manholes without disassembly of the machine is discussed. Calculations indicating dependence of magnetic flux distribution from central solenoid winding quality are presented.