This article describes the change in geometric parameters of the quasi-periodic structure of the micro cavities inside the optical fiber. The quasi-periodic structure is a chain of localized cavities ...filled with oxygen. This structure is produced by the melting of the core of the fiber with a moving plasma spark, which appeared because of optical breakdown. While the quasi-periodic structure is being recorded, the fiber sample is placed between the poles of a permanent magnet with a magnetic field induction of 11.7 T. As a result of the effect of the magnetic field on the moving plasma, there is a change in the parameters of the quasi-periodic structure of the micro cavities. In the region with the highest magnetic field gradient, a 7% increase in the transverse dimensions of the cavities and a 9% increase in the longitudinal dimensions were observed. It was depending on the direction of plasma propagation. In the magnetic field, the effect of dividing the cavities into two unequal parts was observed. The main cavity was located at the axis of the fiber, an additional smaller cavity-satellite appeared at some distance from it in the radial direction.
•The periodic structure in a fiber resulting from the plasma spark propagation.•The magnetic field affects the periodic structure of the micro-cavities.•The geometric parameters of the internal structure were measured.•The spatial instability of the plasma spark is observed.•The magnetic field affects the shape and size of the cavities.
An experimental set-up was constructed based on an ytterbium pulsed laser with a passive holmium fiber shutter, which provides time-domain reflectometry research the length of active fiber samples ...after drawing and during storage. The possibility of studying non-uniformities in the fiber and quantifying the attenuation coefficient of the optical signal at the source wavelength is discussed. The results indicate the possibility of conducting studies of this type, as well as the potential development of reflectometers for the needs of laboratories of enterprises that produce active optical fibers commercially.
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 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 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 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.
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