V. A. Krylov

Design Improvements and R&D Achievements for Vacuum Vessel and in-Vessel Components Towards ITER Construction

During the preparation of the procurement specifications of ITER for long lead-time items, several detailed vacuum vessel (VV) design improvements are being pursued, such as elimination of the inboard triangular support, adding a separate interspace between inner and outer shells for independent leak detection of field joints, and revising the VV support system to gain more structural performance margin. Improvements to the blanket design are also under investigation, an inter-modular key instead of two prismatic keys and a co-axial inlet?outlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R&D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and non-destructive tests for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. In FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block, and the divertor components, have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods.

Vacuum vessel port structures for ITER-FEAT

The equatorial and the upper port structures are the most loaded among those of the ITER-FEAT vacuum vessel (VV). For all of these ports, the VV closure plate and the in-port components are integrated into the port plug. The plugs/port structures are affected by plasma events and must withstand high mechanical loads. Based on typical port plugs, this paper presents the conceptual design of the port structures (with emphasis on the supporting system), and the results of analyses performed.

Kazakhstan tokamak for material testing conceptual design and basic parameters

Abstract The construction of a special machine for plasma facing material testing under powerful and particle and heat flux deposition is necessary for progress of researches in the field of controlled fusion to industrial application. Kazakhstan tokamak for material testing (KTM) is planned as spherical tokamak with moderate-to-low aspect ratio (A=2) and high plasma and vacuum vessel elongation, that allows to reach high plasma parameters, large power-intensity at a compact arrangement of design elements and low requirements to a toroidal magnetic field. KTM tokamak is planned in order to investigate the following issues: (1) Plasma confinement in tokamak with A=2, plasma parameters and configurations working window; (2) Differed kinds of divertor plates under power flux of plasma to divertor volume; (3) Plasma-wall interaction (different materials and coating) and plasma-limiter configurations. In the paper the basic parameters of the machine are given. The design of magnet system with poloidal field coils, vacuum vessel and divertor are submitted.

KTM Experimental Complex Project Status

Abstract A review of KTM experimental complex project status, which is aimed the creation of a Kazakhstani spherical tokamak for study and tests materials and components of future fusion reactors. Revised basic parameters of the KTM facility and ground of the changes taking into account new plasma core geometry, new design of vacuum chamber and modified magnetic system, transport sluice and movable divertor devices, and additional RF-heating system are presented here.

GLOBUS-M. From concept to engineering

GLOBUS-M spherical tokamak is to be designed and constructed at A.F. Ioffe Institute, St. Petersburg. Basic machine parameters, operational regimes, expected plasma parameters, as well, as necessary configuration of magnetic field coils were simulated and corresponding results are presented. Structural analysis of the machine was performed that gives the possibility to identify critical components of GLOBUS-M tokamak with detail analysis and tests of their behavior (central solenoid, conductor for central solenoid and toroidal field coil joints). Project expertise results in revision of some tokamak parts design which improves the machine performance (toroidal and partly poloidal field coils, vacuum vessel). The revision approach is outlined.

Just: Concept development and status

Abstract Concept of the Joint Upgraded Spherical Tokamak (JUST) – a multifunctional facility for working out of physical regimes (discharge scenarios, mode of buming, limits of working plasma parameters), promising divertor devices and testing of reactor materials is presented. The proposed JUST parameters are: R = 1.8 m; a = 1.2 m; A = 1.5; k = 2.3; Bto = 2.1 T; Ip = 10-14 MA, P AUX = 15-20 MW; P FUS∼ 50 MW, tburn ∼ 10 s. Results of the preliminary study of the tokamak design are presented. Wide use of Russian industrial experience in creation of super-powerful electro-generators and advanced technologies of airspace complexes is planned.