The GLOBSYS code was developed for analysis and prediction of parameters of the Globus-M2 tokamak and its modifications. In 1, preliminary selection of correlations which connect physical and ...technical parameters was made. In this paper, the verification of the code using the achieved and predicted data from the installations NSTX, NSTX-U, MAST, MAST-U, and ST40 is given. As a whole, there is good agreement between simulations and plasma parameters at the discharge plateau. The best agreement is observed if ITER confinement scaling is used for energy confinement time with the enhancement factor
H
y
, 2
= 1–1.2. Simulations with other confinement scalings (Globus-2021, NSTX scalings) give good agreement with plasma parameters for the toroidal field
B
t0
~ 0.5 T. For increasing
B
t0
, more optimistic predicted plasma parameters are obtained for the Globus-2021 and NSTX scalings in comparison with the ITER confinement scaling. The condition of reaching the plasma quasistationary regime (or the time of establishment of quasistationary plasma profiles τ
L
/
R
) is estimated for NSTX, NSTX-U, MAST, MAST-U and ST40 discharges. This time is compared with two technical restrictions, which are connected with the times of toroidal field coil heating and poloidal flux capacity. Verification of the GLOBSYS code using the data from the aforementioned spherical tokamaks is the basis for the prediction of parameters of the next step of Globus-M program.
The description of the zero-dimensional engineering-physical code GLOBSYS (Globus spherical tokamak system code), designed for parametric analysis of the next step of the program Globus-M, Globus-M2, ...is given. Within the framework of the zero-dimensional approximation, the definitions of the main scaling parameters of the plasma (poloidal beta, the fraction of bootstrap current, the energy lifetime of the plasma), as well as the specifics of calculating the inductance and resistance of the plasma in spherical tokamaks, are refined. The results of calculations of the plasma parameters by the code were compared with the experimental data of one of the Globus-M2 discharges (no. 38800) with neutral beam heating and showed good agreement. It is proposed to perform a comparison of calculations based on the code with the achieved and predicted parameters of the spherical tokamaks NSTX, NSTX-U, MAST, MAST-U, and ST40 in a separate paper. The goals of the next step (Globus-3) are formulated, the main ones of which are long pulse, high toroidal field, and powerful heating, which allow us to consider Globus-3 as a hydrogen prototype of a neutron source. The infrastructural restrictions on the Globus-3 parameters are given, which require further analysis of various versions of the electromagnetic system. Using the example of Globus-M2 discharge no. 38800, the effect of restrictions on the flow balance and heating of the elements of the electromagnetic system is shown.
The concept of next-generation spherical tokamak is being considered: the Globus-3 project, which, in its characteristics, is compatible with the infrastructure existing at the Ioffe Institute, but ...differs from the currently operating Globus-M2 tokamak in the stronger toroidal magnetic field (1.5–3.0 T) and increased duration of plasma discharge. The parametric analysis data are presented that determined the preliminary selection of the facility parameters. Three options for the electromagnetic system were considered: with the warm copper coils, with the pre-cooled copper coils and with the coils made of high-temperature superconductors. For the first option, the concept for designing the electromagnetic system and vacuum vessel of the facility has been developed. The basic shot scenario with duration of up to 3 s at the field of 1.5 T and plasma current of 0.8 MA is presented.
A concept of plasma-facing components (PFCs) of the TRT tokamak operating in the mode of long discharges and high average thermal load on the walls is presented. The PFC structure and problems and ...the relationship of the main parameters and processes that ensure the comfortable coexistence of the plasma and its surrounding walls are described. Construction solutions for the panels of the first wall and the divertor, which are performed on the basis of ITER technologies, are presented. The features of the use of liquid metals in PFC are discussed. Possibilities of alternative solutions for a divertor target are considered in detail.
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 Δ
t
plateau
over the characteristic settling time of the plasma profiles τ
L
/
R
was chosen as the key condition. At values of the toroidal magnetic field
B
t
0
= 3 T, the condition Δ
t
plateau
> τ
L
/
R
cannot be attained even for precooled EMS. At
B
t
0
= 2 T, only options with precooled EMS can be considered acceptable, but the facility dimensions are fairly large. For the field
B
t
0
= 1.5 T, the options with “warm” EMS correspond to the duration of the plasma current plateau ~3 s (Δ
t
plateau
/τ
L
/
R
~ 1–1.5). In the case of precooled EMS, the duration of the plateau can increase to 12–13 s (Δ
t
plateau
/τ
L
/
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:
R
0
~ 0.6–0.7 m,
a
~ 0.35–0.4 m,
А
≤ 1.7–1.8,
k
95
~ 1.7–1.8. The minimum allowable value of the plasma current
under the condition of effective absorption of the input power of neutral injection has been calculated. In the Globus-3 facility,
I
p
≈ 0.8 MA was chosen as the base value.
The engineering and technical aspects of the creation of the electromagnetic system (EMS) developed using high-temperature superconductors (HTSCs) for the TRT project (Tokamak with Reactor ...Technologies) are considered. Some proposals on the design of superconducting the wire and EMS windings are presented, taking into account a number of restrictions imposed by equipment and operating conditions of the TRT. A brief comparison of the TRT EMS with a number of existing projects is given.
•The purpose of prebaking of the vessel is to check the quality of the numerous welds.•The maximal baking power was 65 kW.•The maximal temperature of vacuum shell was 187 °C.•The difference between ...maximal and minimal temperatures of vessel shell was 60 °C.•After the end of baking, no leaks were found in shell welds.
Presently, the Tokamak T-15MD is being built in the NRC “Kurchatov Institute”. The vacuum vessel was manufactured and passed the preliminary vacuum tests at the plant in St. Petersburg (Efremov Institute) in 2016. Vacuum vessel consists of toroidal shell made from 321stainless steel of 5 mm and 8 mm thickness, horizontal and vertical ports (152 in total), in-vessel elements. The vessel volume is 47 m3 and the surface square faced to plasma is ˜200 m2. The purpose of prebaking of the vacuum vessel is to check the quality of the numerous welds. To bake the vacuum vessel up to 220 °C at the plant in Bryansk, the Ohmic heaters have been laid on vessel shell surface both outside and inside. The thermal insulation (basalt wool) was applied on the outer surface of the vessel. The vessel’s surface temperature was controlled by thermocouples. The currents in heaters was regulated by four thyristor regulators. The temperature data were processed and stored using the data acquisition system.
A neutron collimator is developed to attenuate the neutron flux and reduce the residual induced activity in the interportal space of the diagnostic system of neutral particle analyzers of the ITER ...tokamak reactor. The collimator is installed in the port plug of the ITER vacuum vessel in front of an inlet to the vacuum pipeline of the diagnostic system. The collimator design has a cellular structure with 80% transparency for the neutral atom beam that goes out of the plasma and is recorded by the analyzers. However, because of increased scattering of neutrons in the collimator, their flux in the interportal room in the service zone of diagnostic systems of equatorial port no. 11 is significantly reduced and allows the equivalent dose rate in this zone to be decreased by several times. Thermal analysis showed that, during the reactor operation in modes with generation of the maximum power of 500 MW, the plasma radiation will cause the heating of the collimating grid to a temperature not exceeding 250°C, which makes it possible to select the stainless steel (316L(N)-ITER grade) as a material for manufacturing the collimator. In this case the cyclic strength of the collimator meets the ITER requirements, and it can be used without replacement during the entire deuterium-tritium experiment of the tokamak reactor.
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, Nb
3
Sn, NbTi and Nb
3
Sn, 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).
•T-15U project is the initial technical base for creating fusion neutron sources.•Magnet system of T-15U will confine the hot plasma in the divertor configuration.•Toroidal magnetic field at the ...plasma axis is 2T.•T-15U should begin operations in 2016.
Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48m, a minor radius of 0.67m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.
