ASTRA Code Research Articles

2D–4D Simulations of Neutral Particles Penetration into Central Plasma Region in the L-2M Stellarator. Verification of Results Using ASTRA Code

2D–4D Simulations of Neutral Particles Penetration into Central Plasma Region in the L-2M Stellarator. Verification of Results Using ASTRA Code

Simulation Dosimetry Studies for FLASH Radiation Therapy (RT) with Ultra-High Dose Rate (UHDR) Electron Beam

FLASH-radiotherapy (RT) presents great potential as an alternative to conventional radiotherapy methods in cancer treatment. In this paper, we focus on simulation studies for a linear particle accelerator injector design using the ASTRA code, which permits beam generation and particle tracking through electromagnetic fields. Space charge-dominated beams were selected with the aim of providing an optimized generated beam profile and accelerator lattice with minimized emittance. The main results of the electron beam and ultra-high dose rate (UHDR) simulation dosimetry studies are reported for the FLASH mode radiobiological treatment. Results for the percentage depth dose (PDD) at electron beam energies of 5, 7, 15, 25, 50, 100 MeV and 1.2 GeV for Poly-methyl-methacrylate (PMMA) and water phantom vs. the penetration depth are presented. Additionally, the PDD transverse profile was simulated for the above energies, delivering the beam to the phantom. The simulation dosimetry results provide an UHDR electron beam under the conditions of the FLASH-RT. The performance of the beam inside the phantom and the dose depth depends on the linear accelerator beam’s energy and stability.

Transport simulations and electric field in T-10 and TJ-II

A modeling of the transport processes in T-10 tokamak using Astra code for two sets of discharges having ECRH auxiliary heating is presented. Electric potential measurements using HIBP show that radial profiles are negative for ohmic heating but become positive for large heating powers. We first obtain T α and n profiles by adjusting transport coefficients and then try to obtain the radial E r profiles by assuming three models for it. Results show good agreement for Ohmic and small ECRH power phases but not for large power where E r > 0 indicating that other processes have to be included.

Perspectives in linear accelerator for FLASH VHEE: Study of a compact C-band system

Purpose:In order to translate the FLASH effect in clinical use and to treat deep tumors, Very High Electron Energy irradiations could represent a valid technique. Here, we address the main issues in the design of a VHEE FLASH machine. We present preliminary results for a compact C-band system aiming to reach a high accelerating gradient and high current necessary to deliver a Ultra High Dose Rate with a beam pulse duration of 3μs. Methods:The proposed system is composed by low energy high current injector linac followed by a high acceleration gradient structure able to reach 60–160 MeV energy range. To obtain the maximum energy, an energy pulse compressor options is considered. CST code was used to define the specifications RF parameters of the linac. To optimize the accelerated current and therefore the delivered dose, beam dynamics simulations was performed using TSTEP and ASTRA codes. Results:The VHEE parameters Linac suitable to satisfy FLASH criteria were simulated. Preliminary results allow to obtain a maximum energy of 160 MeV, with a peak current of 200 mA, which corresponds to a charge of 600 nC. Conclusions:A promising preliminary design of VHEE linac for FLASH RT has been performed. Supplementary studies are on going to complete the characterization of the machine and to manufacture and test the RF prototypes.

Choice of Gas Isotope Composition for Neutral Beam Injectors of the FNS-ST Compact Fusion Neutron Source

The dependence of the neutron yield of the FNS-ST (spherical tokamak) fusion neutron source on the fraction of tritium in the core D+T plasma is analyzed for the case of using tritium neutral beam injectors with 200-keV energy and 6-MW power. The FNS-ST operating regimes are explored using the SOLPS4.3 and ASTRA codes for different values of core plasma density ne, T fraction in the plasma, and particle diffusivity. The FC-FNS code is used to estimate the fluxes of the fuel components in the fuel cycle (FC), which are produced by different injection systems: gas puffing, pellet injection, and neutral beam (T) injection. It is shown that in the case of the Т beam injection, in the operating range of parameters, the neutron yield can reach 6.0 × 1017 s−1, which is the value comparable to that obtained for the scenario of D-beam injection into the balanced D+T plasma. In the case of the T-beam injection, in the range of parameters, for which the neutron yield is close to its maximum, the amount of tritium in the FC is lower than in the case of the D-beam injection. The neutron yield can be increased to 6.5 × 1017 n/s if full separation of the D and T is introduced for the gas pumped out from the divertor and puffed back into the torus. With this approach, in the case of the tritium beam, the amount of tritium in the FC is Tinv of ~170 g. If this approach is used in the case of the deuterium beam, the neutron yield can reach 7.0 × 1017 n/s. However, in this case, the amount of tritium contained in the FC increases to 215 g. The results of the analysis performed are used for optimizing the FC of the FNS-C (compact) fusion neutron source, which is planned for construction in the framework of the comprehensive program of the State Corporation Rosatom “Development of Engineering, Technology and Scientific Research in the Field of Using Atomic Energy in the Russian Federation for the Time Period up to 2030.”

Analysis and modelling of momentum transport based on NBI modulation experiments at ASDEX Upgrade

The prediction of plasma rotation is of high interest for fusion research due to the effects of the rotation upon magnetohydrodynamic (MHD) instabilities, impurities, and turbulent transport in general. In this work, an analysis method was studied and validated to reliably extract momentum transport coefficients from neutral beam injection (NBI) modulation experiments. To this end, a set of discharges was created with similar background profiles for the ion and electron temperatures, the heat fluxes, the electron density, and the plasma rotation that, therefore, should exhibit similar momentum transport coefficients. In these discharges, a range of temporal perturbations were imposed by modulating and varying the power deposition of the NBI, electron-cyclotron-resonance heating (ECRH), and ion-cyclotron-resonance heating (ICRH). The transport model including diffusion, convection, and residual stress was implemented within the ASTRA code. The Prandtl number was assessed via the GKW code. A convective Coriolis pinch was fitted and the intrinsic torque from the residual stress was estimated. The obtained transport coefficients agree within error bars for sufficiently small imposed temperature perturbations, as would be expected, from the similar background profiles. This successful validation of the methodology opens the door to study the parametric dependence of the diffusive and convective momentum transport of the main ions of the plasma as well as the turbulent intrinsic torque in a future work.

Plasma Stability in a Tokamak with Reactor Technologies Taking into Account the Pressure Pedestal

Studying stationary regimes with high plasma confinement in a tokamak with reactor technologies (TRT) [1] involves calculating the plasma stability taking into account the influence of the current density profiles and pressure gradient in the pedestal near the boundary. At the same time, the operating limits should be determined by the parameters of the pedestal, which, in particular, are set by the stability limit of the peeling–ballooning modes that trigger the peripheral disruption of edge localized modes (ELM). Using simulation of the quasi-equilibrium evolution of the plasma by the ASTRA and DINA codes, as well as a simulator of magnetohydrodynamic (MHD) modes localized at the boundary of the plasma torus based on the KINX code, stability calculations are performed for different plasma scenarios in the TRT with varying plasma density and temperature profiles, as well as the corresponding bootstrap current density in the pedestal region. At the same time, experimental scalings for the width of the pedestal are used. The obtained pressure values are below the limits for an ITER-like plasma due to the lower triangularity and higher aspect ratio of TRT plasma. For the same reason, the reversal of magnetic field shear in the pedestal occurs at a lower current density, which causes the instability of modes with low toroidal wave numbers and reduces the effect of diamagnetic stabilization.

Selection of Fuel Isotope Composition in Heating Injectors of the FNS-ST Compact Fusion Neutron Source

For the FNS-ST compact neutron source, the dependence of the neutron yield on the tritium content in the bulk plasma is analyzed for the operation of the heating injectors with different isotope compositions of the neutral beams. Self-consistent simulations of the FNS-ST operating regimes are performed using the SOLPS4.3 and ASTRA codes for different densities of the bulk plasma and diffusion coefficients. The FC-FNS code is used to calculate the required fluxes of the fuel components into the plasma provided by different injection systems: the pellet injectors and the neutral beams. In simulations, the plasma density is varied in the range ne = (7–10) × 1019 m−3, and the ratio of the diffusivity to the heat conductivity in the range D/χe = 0.2–0.6. For the scenarios with the D + T or D beams, in the window of the operating parameters, the maximum possible fractions of tritium in the bulk plasma are calculated, and the corresponding neutron yields are obtained. For the regimes with the maximum neutron yield (4.5–5.5) × 1017 s−1, the accumulation of tritium at the site (up to 550 g) is calculated for different heating beams.

First Heat and Particles Transport Study in the Globus-M2 Spherical Tokamak with Neutral Beam Injection at the Current Ramp-Up

The article presents the results of studying the transfer of heat and particles in the Globus-M2 spherical tokamak in discharges with neutral injection at the current ramp up. An atomic beam was injected into the tokamak plasma at a fixed toroidal magnetic field of 0.7 T. The plasma current on the plateau was varied in the range 0.2–0.3 MA. Based on the electron temperature and concentration spatial distributions measured by the Thomson scattering method, the transport of heat and particles in plasma was simulated using the ASTRA code. The energy confinement time of the plasma was determined, as well as estimates of the coefficients of thermal diffusivity and diffusion was made.

Первые результаты исследований ионного теплопереноса на сферическом токамаке Глобус-М2

The paper presents the first ion temperature profile measurements results in the new spherical tokamak Globus-M2, which were obtained in discharges with neutral beam injection using the Charge eXchange Recombination Spectroscopy (CXRS) and the Neutral Particle Analyzer (NPA). Ion heat transport modeling using ASTRA code was carried out based on the obtained experimental ion temperature. The results of the modelling indicate the predominantly neoclassical ion heat transport.

Integrated modelling of core and divertor plasmas for the DEMO Fusion Neutron Source hybrid facility

A steady state regime for the tokamak-based DEMO Fusion Neutron Source (DEMO-FNS) with parameters R/a = 3.2 m/1.0 m, B = 5 Т, Ipl = 4–5 MA, PNBI = 30 MW and РECR = 6 МW is studied using coupled simulations of the central and divertor plasma. In our analysis, the divertor plasma state is determined by the values of the heat flux PSOL, the pressure of the neutrals in the divertor p n and the total number of neon particles NNe outside the separatrix. As the boundary conditions for the core plasma, we use the values at the separatrix of the electron density, ne_sep, and temperatures of the ions, Ti_sep, and electrons, Te_sep, the concentration of the neon impurity normalized to the electron density at the separatrix, cNe = ∑ nNe/ne_sep and the hydrogen neutral influx into the core plasma, Γ0sep. In the divertor region, all the values are calculated using the SOLPS4.3 code for a number of operating points (~150 in our case) with different values of PSOL, p n and NNe. Then the calculation results are approximated by analytical formulas. Power balance in the core plasma is calculated using the ASTRA and NUBEAM codes. The hydrogen (deuterium and tritium) density is modelled taking into account the sources generated by the neutrals originating from the divertor region, as well as by injection of fast atoms and pellet injection. The neon density and radiation in the main plasma are simulated using the STRAHL code.As the result of the simulations, the operational regime of DEMO-FNS is determined, in which the heat loading onto the divertor targets remains at the acceptable level below 10 MW m−2, the divertor plasma does not transit into the ‘full detachment’ mode and the plasma in a double-null separatrix configuration is kept up–down symmetric. Variations of these conditions versus the impurity level and confinement parameters are investigated and discussed.

Heat and Particle Transport in the Initial Phase of Ohmic Discharges in the Globus-M Spherical Tokamak

The results of experiments and simulations of the transport of heat and particles in the Globus-M spherical tokamak are presented. Investigations were carried out in the ohmic mode in hydrogen and deuterium plasma. It is shown that in the phase of current growth under the same initial conditions, as long as the condition is met at which the minimum of the safety factor exceeds unity, two scenarios of discharge development are possible: with an electron temperature dive or with an electron density peak. After the safety factor reaches the value of 1, the differences are no longer observed. Modeling of transport processes using the ASTRA code showed that the dive of the electron temperature and density is a consequence of the decrease in the transport of heat and particles in the plasma core (i.e., the presence of internal transport barriers).

Beam Ion–Driven Alfvén Instabilities in T-15 Baseline Scenario

Properties of toroidal Alfvén eigenmodes (TAEs), driven by neutral beam injection (NBI) hot ions, are described for the tokamak T-15 under construction in the Kurchatov Institute to test a possible influence on the beam and plasma particle losses. The T-15 baseline scenario with a 10-s flat-top 2 MA current stage, 6-MW NBI plus 6 MW of electron cyclotron resonance (ECR) heating is computed with the ASTRA code. The spatial structure and the frequencies of different TAE modes with the toroidal indexes n = 2 to 8 have been obtained with the ideal magnetohydrodynamic KINX code. The bulk plasma Landau damping, linear growth rates, and nonlinear evolution of the TAE mode amplitudes driven by the NBI ions have been computed with the VENUS code. Our numerical estimations for the T-15 TAE modes are compared with experimental and theoretical results for the DIII-D and NSTX tokamaks.

Перенос тепла и частиц в начальной фазе омических разрядов сферического токамака Глобус-М

AbstractThe results of experiments and simulations of the transport of heat and particles in the Globus-M spherical tokamak are presented. Investigations were carried out in the ohmic mode in hydrogen and deuterium plasma. It is shown that in the phase of current growth under the same initial conditions, as long as the condition is met at which the minimum of the safety factor exceeds unity, two scenarios of discharge development are possible: with an electron temperature dive or with an electron density peak. After the safety factor reaches the value of 1, the differences are no longer observed. Modeling of transport processes using the ASTRA code showed that the dive of the electron temperature and density is a consequence of the decrease in the transport of heat and particles in the plasma core (i.e., the presence of internal transport barriers).

Impurity transport in T-10 plasmas with ohmic heating

Impurity transport in the T-10 tokamak plasma with ohmic heating is studied in this paper. The values of various impurities densities, measured with the use of passive spectral diagnostics in the visible (Zeff), active charge exchange measurements (He, C, O), and integral bolometric measurements with absolute extreme ultraviolet detectors (Fe, W) are shown. The experimental data show that accumulation level is growing with impurity nuclear charge and determined by the parameter , which is common for all sorts of impurities. Accumulation process is determined by neoclassical processes and begins with the increase of impurity content in the plasma and ends with the formation of density profiles more peaked than the ne(r). In discharges with low γ anomalous transport completely dominates. So it prevents the impurity accumulation and flattens their density profiles down to the ne(r). These observations correlates with measured negative (positive) plasma potential in discharges with high γ (low γ). 1D modelling using ASTRA and STRAHL transport codes is performed to describe the behaviour of impurities in a wide range of T-10 ohmic regimes. It is shown that the coefficients of anomalous transport Dan and Van established in Krupin et al (1983 Sov. J. Plasma Phys. 9 529–36) and Krupin et al (1985 12th EPS Conf. on Plasma Physics) by describing the density dynamics of injected argon and potassium ions are applicable for the modelling of the He, C, O, W impurity density profiles and their sources. The analysis of the obtained results allows us to state the existence of a common dependence of the anomalous transport for all ions (impurities and deuterons) on the discharge parameters in the T-10 ohmic regimes.

Business is Unusual: Discussing the Philosophy of Astra Code of Ethics

.

Plasma potential and electron temperature evaluated by ball-pen and Langmuir probes in the COMPASS tokamak

The radial distributions of the main plasma parameters in the scrape-off-layer of the COMPASS tokamak are measured during L-mode and H-mode regimes by using both Langmuir and ball-pen probes mounted on a horizontal reciprocating manipulator. The radial profile of the plasma potential derived previously from Langmuir probes data by using the first derivative probe technique is compared with data derived using ball-pen probes. A good agreement can be seen between the data acquired by the two techniques during the L-mode discharge and during the H-mode regime within the inter-ELM periods. In contrast with the first derivative probe technique, the ball-pen probe technique does not require a swept voltage and, therefore, the temporal resolution is only limited by the data acquisition system. In the electron temperature evaluation, in the far scrape-off layer and in the limiter shadow, where the electron energy distribution is Maxwellian, the results from both techniques match well. In the vicinity of the last closed flux surface, where the electron energy distribution function is bi-Maxwellian, the ball-pen probe technique results are in agreement with the high-temperature components of the electron distribution only. We also discuss the application of relatively large Langmuir probes placed in parallel and perpendicularly to the magnetic field lines to studying the main plasma parameters. The results obtained by the two types of the large probes agree well. They are compared with Thomson scattering data for electron temperatures and densities. The results for the electron densities are compared also with the results from ASTRA code calculation of the electron source due to the ionization of the neutrals by fast electrons and the origin of the bi-Maxwellian electron energy distribution function is briefly discussed.

Study of thermal quench of discharge using a massive gas injection into the plasma of T-10 tokamak

The report presents the results of simulations of plasma disruption in the T-10 tokamak using the ASTRA code, including discharges with a disruption being initiated by the massive gas injection (MGI). This study focuses on the development phase of thermal quench, and the effect of different channels of heat loss during quench evolution is examined. It is shown that the initial phase of the slow phase of thermal quench can be described by the selection of sources and transport coefficients.

Transition into the improved core confinement mode as a possible mechanism for additional electron heating observed in the lower hybrid current drive experiments at the FT-2 tokamak

In experiments on lower hybrid current drive (LHCD) carried out at the FT-2 tokamak, a substantial increase in the central electron temperature T e (r = 0 cm) from 550 to 700 eV was observed. A complex simulation procedure is used to explain a fairly high LHCD efficiency and the observed additional heating, which can be attributed to a transition into the improved core confinement (ICC) mode. For numerical simulations, data obtained in experiments with deuterium plasma at 〈n e 〉 = 1.6 × 1019 m–3 were used. Simulations by the GRILL3D, FRTC, and ASTRA codes have shown that the increase in the density and central temperature is apparently caused by a significant suppression of heat transport in the electron component. The mechanism for transition into the improved confinement mode at r < 3 cm can be associated with the broadening of the plasma current channel due to the lower hybrid drive of the current carried by superthermal and runaway electrons. In this case, the magnetic shear s = (r/q)(dq/dr) in the axial region of the plasma column almost vanishes during the RF pulse. In this study, the effect of lower hybrid waves on the plasma parameters, resulting in a transition into the ICC mode, is considered. New experimental and calculated data are presented that evidence in favor of such a transition. Special attention is paid to the existence of a threshold for the transition into the ICC mode in deuterium plasma.

A stepladder approach to a tokamak fusion power plant

We present an approach to design in a consistent way a stepladder connecting ITER, DEMO and an FPP, starting from an attractive FPP and then locating DEMO such that main similarity parameters for the core scenario are constant. The approach presented suggests how to use ITER such that DEMO can be extrapolated with maximum confidence and a development path for plasma scenarios in ITER follows from our approach, moving from low β N and q typical for the present Q = 10 scenario to higher values needed for steady state. A numerical example is given, indicative of the feasibility of the approach, and it is backed up by more detailed 1.5-D calculation using the ASTRA code. We note that ideal MHD stability analysis of the DEMO operating point indicates that it is located between the no-wall and the ideal wall β-limit, which may require active stabilization. The DEMO design could also be a pulsed fallback solution should a stationary operation turn out to be impossible.