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 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 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.
Important progress in the development of high-temperature superconductors (HTSC) of the second group made it possible to design the quasi-stationary tokamak with reactor technologies (TRT) with the ...high magnetic field (
B
t0
= 8 T). The high magnetic field will ensure the achievement of plasma fusion regimes in the tokamak with the fusion energy gain
Q
> 1 at the considerably reduced size of the facility (
R
0
= 2.15 m,
a
= 0.57 m), and, consequently, at its reduced cost. TRT will be capable of operating in the quasi-stationary regimes (≥100 s) with hydrogen, helium, and deuterium plasmas (with the densities
n
e
of up to 2 × 10
20
m
–3
) and in the regimes with short (duration Δ
t
< 10 s) deuterium–tritium plasma shots with the fusion energy gain
Q
> 1 limited by the radiation heating of toroidal coils. TRT is being designed as a plasma prototype for both the pure fusion reactor and the fusion neutron source for the hybrid (fusion–fission) reactor. The TRT missions are the development of the key fusion technologies and their integration in one facility. These technologies are as follows: the HTSC electromagnetic system operating at the extremely high magnetic fields; the metal and liquid-metal (lithium) first wall and innovative divertor; the unique advanced systems for the auxiliary plasma heating and non-inductive current drive, including the systems for atomic beam injection with energy of 0.5 MeV and power of several tens of megawatts, the electron cyclotron heating system based on the megawatt-power gyrotrons with a frequency of 230 GHz and a total power of ~10 MW, and the ion cyclotron heating system at frequencies of 60–80 MHz with a power of several megawatts; the tritium fuel cycle; the remote control technologies; the technologies for diagnostics capable of operating under the fusion reactor conditions; the technologies for maintaining quasi-stationary plasma discharges; and the technologies for the tokamak operation in the fusion ignition regime, in which the heating by alpha particles is the dominant heating mechanism at the axis of the plasma column, in the deuterium–tritium experiments limited by the radiation heating of the toroidal coils. The results are presented from the conceptual design of the basic TRT components, as well as the expected characteristics of its operation. It is shown that TRT has a wide window of working parameters suitable for studying the reactor operating regimes. The high magnetic field provides the necessary margins of the pressure, MHD stability, and plasma controllability variation. Implementation of the advanced divertor and first wall concepts, including those using the liquid-metal technologies, will provide the optimum choice of design options in order to reliably control the heat and particle fluxes under the reactor conditions. The advanced systems for the auxiliary heating and current drive will make it possible to implement both the pulsed and stationary regimes of the reactor operation. Calculations of the TRT discharge scenarios show that, for the DT mixture with equal content of components, the long discharges (with duration exceeding 100 s) can be realized with a neutron flux of more than 0.5 MW/m
2
onto the wall, as well as the stationary discharges with a flux of approximately 0.2 MW/m
2
. Thus, TRT can be a real prototype of the fusion neutron source for the hybrid reactor.
At the present time, the construction of the T-15MD Tokamak is being completed at the National Research Center Kurchatov Institute. The magnet system of the Tokamak T-15MD should provide generation ...and confinement of the hot plasma in the divertor configuration. The design plasma parameters are as follows: major radius of 1.48 m, minor radius of 0.67 m, elongation of 1.7–1.9 and triangularity of 0.3–0.4, plasma current of 2 MA, toroidal magnetic field on the plasma axis of 2 T. The electromagnetic system includes toroidal and poloidal coils. The installation will be equipped with the auxiliary plasma heating and current drive in quasi-stationary discharges lasting hundreds of seconds. Scenarios of quasi-stationary plasma discharges are determined by the capabilities of the electromagnetic system to create and maintain an equilibrium magnetic configuration of the plasma column. Trouble-free operation of the power supply is provided by the protection and blocking system. This work is devoted to calculations of the maximum duration of currents in toroidal and poloidal windings in order to avoid their overheating, as well as the duration and scenarios of quasi-stationary plasma discharges.
Current status of tokamak T-15MD Khvostenko, P.P.; Anashkin, I.O.; Bondarchuk, E.N. ...
Fusion engineering and design,
March 2021, 2021-03-00, 20210301, Letnik:
164
Journal Article
Recenzirano
•T-15MD project is aimed at obtaining a database for creating a thermonuclear neutron source for atomic energy needs.•Magnet system of T-15MD will confine the hot plasma in the divertor ...configuration.•Toroidal magnetic field at the plasma axis is 2 T, plasma current is 2 MA.•Preparation to physical start-up of tokamak T-15MD is completed.•T-15MD should begin operation in 2021.
At the present time, the preparation to physical start-up of tokamak T-15MD is completed in the National Research Center “Kurchatov Institute”. The main parameters of T-15MD are: R = 1.48 m, a = 0.67 m, B = 2.0 T, Ipl = 2.0 MA. The magnet system is capable to maintain without overheating (more 60 °C) the plasma current of 2 MA for 4 s, 1 MA for 20 s, 700 kA for 40 s, 500 kA for 80 s, 300 kA for 160 s and 250 kA for 400 s. Plasma current drive can be maintained either by injection of fast neutrals or by electron cyclotron (EC)-, ion cyclotron (IC)- and low hybrid (LH) - waves. In August 2019 the electromagnetic system, consisting of TF and PF coils, together with vacuum vessel have been assembled in experimental hall. Power supply system of Tokamak T-15MD includes: two substations 110/10 kV, two substations 10/0.83 kV, thyristor convertors and different equipment. Total power consumption during the pulse with plasma current 2 MA and additional plasma heating of 20 MW will consist of 300 MVA. Power supply system is in the commissioning. Tokamak T-15MD will be operate using the information and control system. All the information and control system equipment, required for the implementation of physical start-up of tokamak T-15MD, is available. For plasma control the 250 different electromagnetic probes are installed inside vacuum vessel. The gyrotron with frequency 82.6 GHz and power of 1 MW will be used for pre-ionization.
Abstract
The 1 MW experimental stand was modernized with a scroll swirler and a crushed fuel supply system. Comparative data on combustion and gasification of coal fuel crushed in high-stress mills - ...disintegrator, vibrocentrifugal and hammer mill - at a stand with a thermal power of 1 MW were obtained. The experiments used coal of the Kuznetsk Basin, grade D, with technical characteristics: W
r
, % = 5.4; A
r
, %=22.3; V
r
, % = 32.3; Q
sr
, MJ/kg = 20.0. Elemental analysis showed that: C
r
, %=54.6; H
r
, % = 4.1; N
r
, % = 1.3; S
r
, % = 0.5; O
r
= 11.8. In experiments with grinding coal on a disintegrator mill, the value of H
2
= 4.5 vol.% and CO = 9.4 vol.%, when grinding in a vibro-centrifugal mill, the values of H
2
= 0.6 vol.% and CO = 5.8 vol.%, when grinding in a hammer mill, the values of H
2
= 0.3 vol.% and CO = 2.8 vol.%. When studying the combustion of mechanochemically treated coal samples, it was found that, all other things being equal, the gasification parameters, namely, the gas concentration and the distribution of temperature zones, depend strongly on the type of equipment used for processing. In particular, processing to approximately the same degree of fineness in mechanical mills-activators with constrained impact and in free impact mills (disintegrators) resulted in different flame parameters.
At present, the T-15MD tokamak is being built and its supporting technological systems are being modernized at the National Research Center Kurchatov Institute in the framework of the federal ...target-oriented program “New Generation Nuclear Power Technologies in 2010–2015 and in Prospect to 2020.” The T-15MD magnet system is to ensure producing and confinement of hot plasma in the divertor configuration. The T-5MD plasma parameters are as follows: major radius
R
0
= 1.48 m, minor radius
a
= 0.67 m, elongation
k
95
= 1.7–1.9, triangularity δ
95
= 0.3–0.4, plasma current
I
p
= 2 MA, and toroidal magnetic field at the plasma axis
B
TO
= 2 T. The tokamak will be equipped with an auxiliary plasma heating and current drive system (
P
aux
= 15–20 MW), which will enable achieving a high plasma temperature (
T
i
~
T
e
of 5–9 keV) and a plasma density (
n
e
≈ 10
20
m
–3
) during a discharge with a pulse duration of up to 30 s. The magnet system includes a toroidal winding and a poloidal magnet system. The poloidal magnet system may generate the one- or two-null divertor magnetic configuration. The power supply system provides the necessary current scenarios in the magnet system windings. The T-15MD magnet together with the vacuum vessel was pre-assembled at the NPO GKMP Ltd. plant in Bryansk. All elements of the magnet system and the vacuum vessel are delivered to the National Research Center Kurchatov Institute in Moscow, where the T-15MD tokamak will be assembled. The technological systems are to be reconstructed by the end of 2019. The physical start-up of the T-15MD is scheduled for December 2020.
Data on plasma disruption processes in the modernized Globus-M2 spherical tokamak are presented. Electron temperature and density profiles before the disruption, immediately after thermal quench and ...in the stage of plasma current quench are measured using the diagnostics of Thomson scattering of laser radiation. The dependence of the plasma current decay time during disruption on the pre-disruption current value is determined. The distribution of the toroidal current, which is induced during disruption, in the shell of the vessel is determined on the basis of magnetic measurements. Electromagnetic loads on the vessel are calculated.
Display omitted
•T-15MD project is the initial technical base for creating fusion neutron source for atomic energy needs.•The preassembly of the tokamak T-15MD magnet system together with the vacuum ...vessel was completed.•Most of tokamak systems were manufactured and preliminary tested before the final assembly of tokamak.•All the diagnostic equipment is available and part of it was used in experiments on tokamak T-10.•Physical start-up T-15MD is scheduled for December 2020 year.
At the present time, in the NRC Kurchatov Institute under the auspices of the Federal Target Program “Nuclear energy-technologies of new generation for period 2010–2015 and to the prospect until 2020” the tokamak T-15MD and supporting facilities are being built. The preassembly of the tokamak T-15MD magnet system together with the vacuum vessel was completed at a plant in Bryansk. All elements of the magnet system and vacuum vessel have been delivered to the NRC “Kurchatov Institute” in Moscow for the tokamak T-15MD assembly. It is expected that the T-15MD assembly will be completed in March of 2019. The reconstruction of the sub-station 110/10/04 kV for own needs was completed in 2017 and the reconstruction of the main sub-station 110/10/1 kV, 300 MW was completed in 2018. Twenty- two of the new transformers 10/1 kV and 20 new thyristor convertors will be installed during 2018–2019 period. One gyrotron with output power 1 MW for pre-ionization should be installed in 2019. Tokamak T-15MD connection to water and electrical communication and also the adjustment of control system will be completed in the middle of 2020. Physical start-up T-15MD is scheduled for December 2020 year.
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