The ratio of energy stored in the magnet to the mass of the structure required to withstand the electromagnetic load is known to be one of the most important characteristics of a system used as a superconducting magnetic energy storage (SMES).The concept of quasi-force-free winding, when applied to the design of the SMES magnet system, shows the way to increase the energy to mass ratio and allows to develop new SMES systems with technical and economical parameters better than the existing ones. This paper presents some progress in development of the concept for the SMES with stored energy up to 10 GJ and describes the steps of scaling the "large SMES project" towards a representative model for lab testing. The lab model has been manufactured and tested. The results of the magnetic measurements performed on the lab model confirm the correctness of the main theoretical assumptions which the concept is based upon and, in some way, a possibility to produce the SMES with a high energy-to-mass ratio.

Одной из интереснейших и амбициозных отечественных разработок в области термоядерных технологий является эскизное проектирование токамака ТРТ [1]. Компактность и высокие рабочие параметры установки приводят к необходимости конструировать основные обмотки электромагнитной системы с использованием сильноточных проводов на базе высокотемпературных сверхпроводников (ВТСП). Важным элементом обмотки крупной электромагнитной системы (ЭМС), от которой во многом зависит надёжность её работы, является контактное соединение (КС). Среди требований, предъявляемых к конструкции КС, главными являются низкое омическое сопротивление, минимальное влияние на распределение тока в проводе, компактность и технологичность в изготовлении. Вопросам анализа отечественного и зарубежного опыта разработки КС сильноточных ВТСП-проводов, а также обоснованию выбора концепции КС для обмоток ЭМС ТРТ была посвящена [2]. Там же кратко излагались некоторые аспекты экспериментальной программы исследования образцов и макетов КС. В настоящей статье экспериментальная программа рассмотрена более подробно: освещаются вопросы выбора конструкции макетов КС, описан процесс изготовления образцов, представлено стендовое испытательное оборудование, приводятся детали проведения тестов и результаты измерений величины омического сопротивления макетов КС трёх типов. На основании проведённых исследований делается вывод о соответствии выбранных инженерных решений целям и задачам проекта ЭМС ТРТ. Conceptual design of TRT – the tokamak with reactor technology [1] is one of the most interesting and ambitious projects in the RF national thermonuclear program. To ensure high performance and compact design of the machine, designers of the electromagnetic system (EMS) have to consider HTS as the only option for high-current winding conductor for the major coils. The electrical joint of high-current cables should be treated as an important element of winding upon which the reliability of a large EMS depends. The major requirements for a high-current electrical joint of HTS cables have been formulated as follows: low ohmic resistance, minimal effect on the current distribution among cable elements, compact design, consistent and relatively simple technology of joint manufacturing and assembly. The analysis of the international projects concerning HTS join design, the experience gained in the framework of the domestic HTS programs, as well as the main reasons for selection of the present TRT joint concept were presented in paper [2]. Some aspects of the TRT joint test program were also mentioned there. In this paper, the test program is presented with more details: the design of the joint prototypes and the process of sample manufacturing, the test facility and some aspects of the experiments, as well as the test results are given and discussed. The results of sample manufacturing and the tests performed allow us to conclude that the basic concept and the engineering solutions selected for the HTS joints answer the purposes of the TRT EMS project.

High field magnets are required for fusion, biochemical and other scientific applications. Utilization of second generation high temperature superconductors (2G HTS) in such magnets is beneficial for achieving record magnetic fields. 2G HTS-based magnet was successfully designed, manufactured and tested. The magnet consists of two coaxial coils: a partially insulated 40 mm cold bore layer-wound inner section and an outer section with eight double pancakes (100 mm cold bore). The special high-field YBCO conductors were developed and produced: 12 mm mechanically reinforced tape for double pancakes and 4 mm two-tape stack for layer windings. Both HTS conductors are partially or fully insulated by UV cure urethane varnish. The double pancake pack were additionally reinforced by Hastelloy wrapping and aluminum casing. The superconducting magnet with HTS bus-bars and HTS current leads was installed into a helium cryostat. Two cryocoolers were used, one for helium re-condensation and another for current leads cooling. Inner and outer sections of the magnet operated separately by different power sources. During the tests was achieved magnetic field 20.1 T.

The engineering part of the GLOBSYS code is presented, and the parameters of the Globus-3 facility, which is a development of the Globus program, are analyzed. The facility is primarily designed to provide a long pulse, a large toroidal magnetic field and strong heating. The concepts of searching for Globus-3 parameters under physical and engineering limitations are described. Obviously that reliable confinement and a large part of noninductive current are necessary to ensure existence of a plasma for a long time. Engineering constraints are involved in the choice of parameters in a more complex way: in some cases, it is overheating of the coils, in other cases, it is the total power supply, or the limit on the flux provided by the ohmic solenoid, or the strength of the constructions. The parameters of the Globus-3 spherical tokamak were preliminarily selected for the cases of a “warm” copper EMS (Electromagnetic system) and the EMS precooled to liquid nitrogen temperature. The exceeding of the duration of the plasma current plateau Δtplateau over the characteristic settling time of the plasma profiles τL/R was chosen as the key condition. At values of the toroidal magnetic field Bt0 = 3 T, the condition Δtplateau > τL/R cannot be attained even for precooled EMS. At Bt0 = 2 T, only options with precooled EMS can be considered acceptable, but the facility dimensions are fairly large. For the field Bt0 = 1.5 T, the options with “warm” EMS correspond to the duration of the plasma current plateau 3 s (Δtplateau/τL/R 1–1.5). In the case of precooled EMS, the duration of the plateau can increase to 12–13 s (Δtplateau/τL/R 5). In the latter case, as a basis for further development of the Globus-3 facility, options with the following geometric dimensions are reasonable: R0 0.6–0.7 m, a 0.35–0.4 m, А ≤ 1.7–1.8, k95 1.7–1.8. The minimum allowable value of the plasma current I_p,min under the condition of effective absorption of the input power of neutral injection has been calculated. In the Globus-3 facility, Ip ≈ 0.8 MA was chosen as the base value.

Further development of a fusion-fission hybrid facility based on a superconducting tokamak, DEMO-FNS, continues in Russia for integrated commissioning of steady-state and nuclear fusion technologies at a power level up to 40 MW fusion and 400 MW fission reactions. This facility is considered as the main source of technological and nuclear scientific information in the RF National program on controlled fusion and plasma technologies that is currently being developed and submitted to the authorities for approval. In this work the progress over the last two years is reported in core plasma modelling, design and integration of enabling systems, including the first wall, divertor, neutron beam heating and current drive, as well as the choice of materials. Use of supercritical CO2 coolant was reviewed from an activation point of view. Interaction of fusion-fission hybrid systems with the nuclear fuel cycle of Russia's nuclear power industry was shown. onstruction of the device is expected in 2033.

New fusion devices are being discussed in Russia. One item is a superconducting magnet development for the demonstration hybrid facility-fusion neutron source based on a tokamak concept with the conventional aspect ratio ~3. The magnetic parameters of this device are planned to be: 5 Ton the plasma axis and about 12 Ton the high field side of the toroidal field coils. Because of a high neutron flow and a thick shielding, the space remained for the inner legs of the toroidal field coil is very tight. Therefore, a very high current density is needed for the windings. Several magnet designs have been addressed up to now with required engineering critical current density sim 1000 A/mm 2 at least. In this review, we present conceptual design of the magnet system, necessary strand parameters, and their irradiation properties. Few possible designs of the toroidal field, conductors are discussed. The R&D of efficient HTS current leads to be used for fusion machines are discussed as well.

This paper gives an overview of the new facility for MHD and heat transfer (HT) tests of liquid metal breeder blanket mock-ups in high magnetic field. The facility named LIMITEF5 is under construction now in JSC "NIIEFA" (D.V. Efremov Institute). The facility includes the lead-lithium (LL) loop passing through the warm aperture of the superconducting magnet. Superconducting magnet is planned to be put in operation in 2018 with the following characteristics: - magnetic field induction is up to 5.5 T; - dimensions of the magnet "warm" zone: diameter - 900 mm, length - 1600 mm; - winding design: low temperature superconducting split solenoid. LL loop consists of melting and feeding tanks; main loop with electromagnetic pump (EMP), electromagnetic flow meter, calibration nozzle, heat exchanger and blanket mock-up; LL impurities control and purification system containing oxygen sensor, plug indicator, cold trap, LL sampler. The details of the lead-lithium ceramic breeder test blanket module (LLCB TBM) mock-up for MHD/HT tests in magnetic field of similar to 5 T which is under conceptual design are also given. (C) 2017 Elsevier B.V. All rights reserved.

The possibility to design, manufacture and test the dipole type magnets from the second generation hightemperature superconductors (HTS-II like YBCO and ReBCO) was demonstrated at the Efremov Institute. The paper describes available computation techniques, design approaches and manufacturing equipment, which could be used to meet the modern requirements for the magnets of accelerators, research equipment, magnet levitation systems etc. The manufacturing equipment comprises the winding lines and insulating devices to provide different configurations and insulating schemes of coils. Additionally, an equipment to produce the Roebel-cable for high current applications was procured and put in operation. As an example, the results of development of the HTC-II dipole type magnets for the different kind dummies of maglev systems are presented. The ReBCO tapes produced by JSC “SuperOx” (Moscow) were used. Up to 0.5 T magnets cooled by liquid nitrogen were designed as a part of levitation system consisting of permanent, HTS-II and normal conductive magnets. Comprehensive tests verified the computation results and demonstrated the readiness to develop HTS-II dipole magnets under the customer requirements.

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 k (95) = 2. The fusion power is P (FUS) = 40 MW. The toroidal magnetic field on the plasma-filament axis is B (t0) = 5 T. The plasma current is I (p) = 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).

In recent years increasingly discusses the prospects of application of high-temperature superconductors (HTS) as the winding current-carrying elements of magnetic systems for various purposes. It seems particularly attractive possibility of such systems at liquid nitrogen temperature. The article describes the prototype of module of the magnetic system which is made on the basis of high-temperature superconducting tapes, designed for the installation and testing on a working model of a static levitation. In the working model levitation of the platform carried by the interaction of the magnetic field of the assembly of permanent magnets mounted on the platform with a field similar to assemblies located in the track structure. Compact HTS module replaces the two assemblies of permanent magnets mounted on the platform. Each block of the module represents HTS racetrack coil with current inputs, power structure, positioning system and bracing which is placed in a cryostat, providing at minimum wall thickness of the required mechanical strength and thermal insulation at liquid nitrogen temperature. The prototype of unified superconducting module successfully passed preliminary tests.

FGUP «Efremov's NIIEFA» has more than 50 years of experience for creation of superconducting magnets and devices. Among its well-known projects there are magnetic frames UNK, TSP, T-15, Globus-M, ITER, etc. In addition to conventional low-temperature superconductors today the VTSP technology is developing rapidly. NIIEFA possesses the technology and has all necessary experience for design, manufacturing and testing the superconducting coils (unified module) of magnetic levitation transport system.

A computational technique is proposed for the analysis and design optimization of inductive-type superconducting fault current limiters (SFCL) with variable impedance for power systems application. The technique is applicable for a wide range of magnet systems. The paper is focused on the analysis of electromagnetic transients. A methodological example of a model SFCL is presented. Numerical experiments are described to study nonlinear effects at saturated and unsaturated states. Results demonstrate that the nonlinear magnetic behavior of the SFCL components affects noticeably the accuracy of predictions.