E. A. Azizov

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.

Project ‘Baikal’ – testing the scheme for electric pulse generation

The scheme of a multi-objective installation ‘Baikal’, a multi-megajoule source of X-ray emission with a temperature of 250–300 eV and a power of 500–1000 TW, is described in this article. The Project of the ‘Baikal’ installation is based on an inductive energy store, which in the process of step-by-step energy transmission with increasing power generates an electric pulse with parameters required for compression of liners. Electric parameters of the ‘Baikal’ installation are: current −50 MA, voltage 8–10 MV, pulse duration −150 ns, pulse energy 30 MJ. An installation ‘MOL’ is intended for testing the scheme of one of ‘Baikal’ modules. It consists of an inductive store IN-1 with accumulated energy of 12,5 MJ, capacitor banks, a magnetic amplifier, a magnetic compressor with a capacitor bank for generation of initial magnetic flow, a plasma opening switch (POS) and a load simulator.

Tokamak DEMO-FNS: Concept of magnet system and vacuum chamber

The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion—fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet system and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k95 = 2. The fusion power is PFUS = 40 MW. The toroidal magnetic field on the plasma-filament axis is Bt0 = 5 T. The plasma current is Ip = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb3Sn, NbTi and Nb3Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron—water radiation shield (70% of stainless steel and 30% of water).

Pulse power system development for megajoule X-ray facility BAIKAL

TRINITI develops a project of multiterawatt generator «BAIKAL» to produce powerful pulses of soft X-rays, using electric pulse power 500 – 1000 TW. Parameters of proposed X-ray generator are: X-ray pulse energy − 10 MJ, X-ray pulse duration − 10 ns, load current amplitude − 50 MA. The methods of pulse power increasing proposed for Baikal project are studies and tested on 12 MJ inductive store installation MOL located at TRINITI Institute.

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.

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233U fuel production: 1. Concept of a stationary Tokamak as a neutron source

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

Assessment of restrictions imposed by neutron irradiation on the workability of a copper conductor in volumetric neutron source coils

The paper presents the results of the analysis of the irradiation conditions and the quantitative estimation of factors that control the lifetime of copper alloys as materials for a conductor in volumetric neutron source coils, as well the results of the analysis of the effect of neutron irradiation at T irr = 100°C on the mechanical properties and electrical conductivity of copper alloys of Cu–Cr–Zr and GlidCop type and of pure copper.

Tokamak tma concept

Abstract Conceptual design and construction of the “TMA” tokamak is presented. TMA is a Tokamak of Minor Aspect ratio (A = 2) and elongated cross section k ≤ 2. It is planned to be created on the basis of existing TSP tokamak-complex at TRINITI, Troitsk, on a short time. Main parameters of installation are: toroidal field Bto≤2T, major plasma radius R = 0.7 m, minor plasma radius a = 0.35 m, plasma current Ip ≤ IMA, plasma elongation k ≤ 2, triangularity δ ≤ 0.7. Inductor flux of a value Δφ = ±0.5 V.s, plasma RF heating power PRF ≤ 9 MW. Duration of plateau of toroidal field is expected to be Δt = 2 s. Main goals of TMA are: study of moderate aspect ratio of elongated shaped plasma with divertor, achievement of a high bootstrap current, study of active controllable SOL plasma.