Electron Cyclotron Heating Power Research Articles

Experimental research of ECW pre-ionization and assisted startup in EAST

Experimental research on the electron cyclotron wave (ECW) pre-ionization and assisted start-up was carried out systematically for the first time in EAST tokamak, which is a superconducting device with ITER-like full metal wall. Breakdown and plasma initiation at low toroidal electric fields (<0.3 V m−1) with ECW pre-ionization and startup assistance has been demonstrated. Also, the parameter domain of breakdown is significantly extended towards higher prefill gas pressure. The effect of ECW injection timing, power, toroidal injection angle on breakdown were also investigated. Injecting ECW earlier leads to an earlier breakdown and a higher plasma current ramp rate. The electron cyclotron heating (ECH) power threshold for breakdown in EAST is approximately 0.4 MW. In the range of ECH power tested in this work, higher ECH power is advantageous for achieving earlier and faster breakdown. Furthermore, the breakdown with radial ECW injection occurs earlier compared with oblique injections (co-current and counter-current). During the ECW-assisted startup, the process of burn-through is prolonged by the higher pre-filled gas pressure even though it enhances the ease of breakdown. In addition, compared to the low hybrid wave assistance, the ECW assistance has an effect in averting the generation of runaway electrons and improving the safety of device during startup. Moreover, the ECW assistance exhibits a high tolerance to the impurity and thus ensures a high ramp rate of plasma current even with a high impurity level.

First application of data assimilation-based control to fusion plasma

Magnetic fusion plasmas, which are complex systems comprising numerous interacting elements, have large uncertainties. Therefore, future fusion reactors require prediction-based advanced control systems with an adaptive system model and control estimation robust to uncertainties in the model and observations. To address this challenge, we introduced a control approach based on data assimilation (DA), which describes the system model adaptation and control estimation based on the state probability distribution. The first implementation of a DA-based control system was achieved at the Large Helical Device to control the high temperature plasma. The experimental results indicate that the control system enhanced the predictive capability using real-time observations and adjusted the electron cyclotron heating power for a target temperature. The DA-based control system provides a flexible platform for advanced control in future fusion reactors.

Determining the access to H–mode in the ITER pre–fusion and fusion power operation phases at low plasma current with full–radius TGLF–SAT2 simulations of L–mode plasmas

The pre–fusion power operation 1 phase of ITER is planned to be characterized by electron cyclotron resonance heating only. Under the assumption that the access to H–mode is determined by a critical ion heat flux at the plasma edge, full–radius ASTRA simulations with the TGLF–SAT2 transport model are performed in order to compute the ion heat flux produced by the thermal exchange between electrons and ions in different operational conditions. Both hydrogen and deuterium plasmas at 5 MA are considered, respectively at 1.8 T and 2.65 T, corresponding to one third and half of the nominal maximum magnetic field. Different levels of electron cyclotron heating power are considered in sets of simulations with increasing values of the electron line averaged density. The predictions are compared with the currently available scaling of the critical ion heat flux. In hydrogen, 20 MW of electron heating power are predicted to allow H–mode access in a vanishingly small density window, whereas 30 MW and 40 MW would allow more substantial H–mode operational windows. Despite the fact that in deuterium plasmas the thermal exchange between electrons and ions is smaller by the hydrogen to deuterium mass ratio compared to hydrogen plasmas, the lower H–mode power threshold in deuterium leads to the prediction that an even broader and more robust domain to access H–mode is obtained at half field at 40 MW in deuterium as compared to operation in hydrogen at one third of the maximum magnetic field, even at the same power.

Analysing the effects of heating and gas puffing in Proto-MPEX helicon and auxiliary heated plasmas *

The Material Plasma Exposure Experiment (MPEX) is being constructed at Oak Ridge National Laboratory to investigate critical fusion reactor issues, such as plasma–material interactions (PMIs) under reactor-relevant conditions and time scales. The linear device Proto-MPEX was used as a test bed to address anticipated research and development issues associated with heating scenarios and establish the physics basis for MPEX. The SOLPS-ITER code suite has been applied to understand plasma and neutral transport in Proto-MPEX and to increase confidence in predictive simulations for MPEX. Coupling between COMSOL and SOLPS is performed to implement a 2D electron heating profile of the helicon source. The simulations show reasonable agreement with the experimental data for plasma with helicon and auxiliary electron cyclotron heating (ECH). Both Bohm and constant diffusion (D: 0.5 m2 s−1 and : 1 m2 s−1) simulations show similar levels of agreement with respect to the sparse experimental data available, assuming a few per cent impurity concentration. ECH significantly increases the target electron temperature, however, the target electron density is reduced compared to helicon-only heated plasmas due to an increase in flow velocity and radial losses. The simulations show that further increasing the ECH power results in an increase in the target electron density due to increased recycling flux and ionization. The results indicate that ECH significantly enhances the target heat flux, with ECH power of 50 kW increasing the target heat flux from 0.4 to 17 MW m−2. It is found that a small amount of gas puffing (GP) near the target plate can further increase the target heat fluxes at the higher ECH power cases, but the target heat flux is reduced at higher GP conditions due to a significant reduction in electron temperature via radiation. ECH and GP scenarios can generate a higher target flux, facilitating improved PMI studies with more reactor-relevant plasma conditions.

Experimental study of electron cyclotron heating assisted start-up on J-TEXT

Second harmonic X-mode (X2 mode) electron cyclotron heating (ECH) assisted start-up has been studied experimentally on the J-TEXT device to determine the minimum ECH power requirements to assist breakdown and develop a better physical description of the process. Results indicate that the minimum toroidal electric field for a successful start-up on J-TEXT changed from to , and that injecting of X2-mode ECH power can ensure robust breakdown. The critical ECH power for successful start-up was determined to be approximately . At lower powers ECH was observed to cause ionization, but this did not necessarily result in successful start-up. The effects of varying ECH power, pulse-width and toroidal magnetic field on start-up were also investigated. Higher ECH power leads to quicker, stronger ionization and CIII emission, and is beneficial for burn-through. Higher pre-plasma density caused by high ECH power can decrease the required toroidal electric field for ohmic breakdown, while the enhanced CIII emission may not be good for start-up. The characteristics of the pre-plasma formed by ECH prior to application of the loop voltage were also studied. The toroidal magnetic field affects the initial location where pre-plasma forms and this also affected the subsequent tokamak start-up. The analysis of spatial density showed that the pre-plasma can radially develop at a velocity of . Furthermore, it was found that injecting ECH power at the appropriate time with a low power of can achieve similar pre-ionization results to the high-power case. The transition from ECH plasma to ohmic plasma suggests that the ECH assisted start-up modified the process of purely ohmic breakdown.

Particle orbit description of cyclotron-driven current-carrying energetic electrons in the EXL-50 spherical torus

In EXL-50 plasma currents over 100 kA are non-inductively generated and maintained solely by electron cyclotron heating (ECH) power with an efficiency of ∼1 A W−1. These currents are carried by energetic electrons (EEs) in the energy range from several tens of keV up to several hundreds of keV which also account for almost all pressure in plasma. This EE component can be viewed as a large number collection of various periodic orbits of energetic particles. Based on this picture we have developed a method for particle orbit description of the EE component in a typical plasma at = 121 kA as analysis target. We use a fluid description as a bridge to describe successfully the EE component as a collection of various passing and trapped orbits in the approximation of monochromatic particle energy with good matching to the flux loop signals. The description has revealed characteristics of passing and trapped particles. Passing particles carry almost all toroidal current, while they account for only 20 of total particle number of the EE component. While net current carried by trapped particles is a very small fraction, they account for a major fraction in number and carry a large positive current outside the last closed flux surface (LCFS) and a large negative current inside. As a result, trapped particles redistribute the current from inside of the LCFS to outside both radially and vertically, generating a large vertically elongated cross section in current as well as number density profiles. There is a ridge-like structure along the LCFS in the current density profile, with no such structure in the number density profile. The results suggest that forward passing particles are more advantageous in confinement than backward passing particles. This advantage increases with particle energy and contributes to the current generation observed in EXL-50 experiments.

Validation of a plasma burn-through simulation with an ECH power absorption model in KSTAR

An electron cyclotron heating (ECH) power absorption model was integrated with a plasma burn-through simulator, DYON, and the new version, DYON-EC, was validated against KSTAR ECH-assist start-up experiments. The absorbed ECH power was calculated with an analytic formula which is a function of the electron density and electron temperature and ECH hardware settings, namely, the injected power, wave frequency, harmonic number, mode fraction, and beam injection angles. Wave parameter changes by wall reflection was also included to simulate multiple reflections. The absorbed ECH power was self-consistently included in the electron energy balance equation. The simulation settings of the plasma-wall interaction model and the electromagnetic scenario including the eddy current model were optimized to reproduce the plasma parameter evolution and line emission data in a pure ohmic start-up discharge. The study revealed that assumption of double-path EC beam absorption is required to reproduce a KSTAR ECH-assisted start-up discharge. Using the same optimized settings, DYON-EC modelling successfully reproduced multiple KSTAR EC-assisted discharges in a large range of operation parameters. The good statistical reproduction of measured plasma parameter evolution confirms the validity of the DYON burn-through modelling with ECH.

Electron cyclotron heating assisted start-up experiments in J-TEXT

Electron cyclotron heating (ECH) breakdown and burn-through assist has been adopted to make the ITER start-up, which uses a low toroidal electric field of about 0.3 V/m, more robust. Related studies have been carried out on many fusion machines such as DIII-D, JT-60U, Tore Supra and FTU. However, the required ECH power to ensure an effective breakdown assist at ITER is not yet clear. Experiment was also performed in J-TEXT to determine the minimum ECH power requirements to assist breakdown and develop a better physics description of the process. The breakdown loop voltage for a successful shot was reduced from 34 to 3.7 V (corresponding to 0.56 V/m) by 300 kW X2-mode ECH power. The critical ECH power to assist the successful formation of a tokamak discharge is about 200kW, and below this value, ionization caused by ECH power is observed, however, the tokamak discharge start-up does not succeed. The effect of different loop voltages and ECH pulse-width on start-up was also studied. As injecting ECH power at a proper time, low ECH power can achieve a similar pre-ionization result as the case with high power. Extremely low breakdown voltage leads to a higher toroidal field later when putting a capacity to continue discharging in J-TEXT. The earlier shutdown of ECH power caused a failed discharge.

Design of Stray Radiation Sensor for ITER ECE Diagnostic

The Electron Cyclotron Emission (ECE) diagnostic has a primary role in the measurement of electron temperature profile and electron temperature fluctuations in ITER. This diagnostic shall be exposed to significant power due to unabsorbed Electron Cyclotron Heating (ECH) power in the plasma. The expected stray power loads could be a few tens of watts, and therefore, the protection of millimetre wave components is one of the design challenges of ITER ECE diagnostic. This protection system includes sensors, a band stop notch filter, and a shutter to stop the RF stray radiation from being incident on the sensitive components. The sensors will be positioned along the ECE transmission line, and shall be used for real-time power monitoring of the stray radiation. Here, we describe a novel design of a sensor for monitoring the stray radiation power. This sensor is a Schottky Diode rectenna, known for high-power and high-speed millimetre wave detection capability. It consists of a 2x2 microstrip patch antenna array, a matching circuit, a diode, and a low pass filter. The antenna array is designed analytically and optimized in CST Microwave Studio, for wide reception angle, high gain, and low side lobe levels. Furthermore, the rectifying circuit is optimized using Agilent Advanced Design System (ADS) software to get better rectification and impedance matching of the signal, thereby improving its detection sensitivity. The ADS simulation results show that the detection sensitivity is about 1000V/W for input power of -30 dBm at 170 GHz, thereby achieving the required performance of the sensor.

Predictive modeling of operation scenarios with negative central magnetic shear and large-radius ITBs on EAST

Negative central magnetic shear or magnetic shear reversal is an attractive aspect for advanced operation scenarios in tokamaks as it suppresses turbulence and facilitates the formation of internal transport barriers (ITBs). The advanced operation scenario with negative magnetic shear, which is associated with high confinement quality, a large bootstrap fraction, ITBs, and so on, is one of the future goals of the experiments on EAST. In this work, modeling efforts have been made to find the operation regimes on EAST with negative central shear and large-radius ITBs. The modeling results indicate that negative central shear and large-radius ITBs can be achieved when the electron density is 〈ne〉/nG &amp;gt; 0.7 and the off-axis deposited electron cyclotron heating (ECH) power is 2 MW. Additionally, scenarios with lower ECH power (∼1.5 MW) have been studied since at present, the maximum ECH power that can be provided on EAST is 1.5 MW. With a lower off-axis ECH power of 1.5 MW, scenarios with negative central shear and large-radius ITBs but with a higher electron density of 〈ne〉/nG &amp;gt; 0.8 are obtained. The same as the first case, if the electron density is reduced to 〈ne〉/nG = 0.8 in this case, large-radius ITBs disappeared in the predicted scenario. This reveals that high electron density and large off-axis deposited ECH power are important for scenarios that have q-profiles with negative central shear and large-radius ITBs since lower hybrid current drive (LHCD) tends to peak near the axis at lower electron density, and hence, it helps to increase the temperature gradient near the axis, which will promote bootstrap in this region, leading to a centrally peaked total current density profile. Higher off-axis deposited ECH power helps to obtain scenarios with strong negative central shear and large-radius ITBs at a lower density. Scenarios predicted in this work will guide future experiments on EAST.

Upgrading LHDGauss Code by Including Obliquely Propagating Wave Absorption Effect for ECH

LHDGauss is a multi-ray tracing code to calculate microwave beam propagation and power deposition of electron cyclotron heating (ECH) using the electron density and electron temperature profiles in the Large Helical Device (LHD) plasma. LHDGauss also takes into account the injected wave polarization purity. LHDGauss uses the cold plasma dispersion to calculate ray propagation and derives the power absorption profile of ECH by using the absorption coefficient based on weakly relativistic plasmas and the wave propagating perpendicularly to the magnetic field. However, the ECH beam propagation is mainly oblique to the magnetic field in the LHD and it is necessary to take into account the angular dependence of the absorption coefficient in order to derive more accurate ECH power deposition profiles. We upgraded LHDGauss and were able to calculate the absorption coefficient using the weakly relativistic dielectric tensor which considers the influence of the angle between the magnetic field and the wave vector. We compared the power deposition profiles calculated by previous and upgraded LHDGauss with the changes of the electron temperature profile of the LHD plasma in the oblique and the perpendicular O1-mode ECH injection cases. We have obtained more reasonable power deposition profiles by using the upgraded LHDGauss in both the perpendicular and the oblique injection cases.

Stability of beta-induced Alfvén eigenmodes (BAE) in DIII-D

Although the stability of ellipticity, toroidal and reversed-shear Alfvén eigenmodes (EAE, TAE, RSAE) are relatively well understood, less is known about the stability of lower-frequency modes such as the beta-induced Alfvén eigenmode (BAE) but, because they are often unstable in present devices and are implicated in fast-ion transport, understanding their stability is vital. BAE stability is studied in primarily weak or reversed shear DIII-D plasmas with sub-Alfvénic deuterium beams. Modes are classified based on electron cyclotron emission, beam emission spectroscopy, magnetics, and interferometer data. The study is limited to the initial two seconds of the discharge, where the evolving q profile provides an effective scan of the dependence of stability upon q. In a dedicated experiment, BAEs are unstable at times in the discharge when the minimum of the safety factor q min is close to a rational number. The observed mode frequencies are usually close to analytic estimates of the BAE accumulation point and the eigenfunction peaks in the vicinity of q min. Unstable BAEs usually occur in bursts that chirp rapidly in frequency. To isolate the importance of thermal and beam gradients in driving the modes, the beam and electron cyclotron heating power is altered for 50–100 ms durations in reproducible discharges. As expected from the resonance condition, BAEs depend sensitively on the beam power and injection geometry. Modes only persist for ∼25 ms because the anisotropic beam population only interacts strongly with the modes over a relatively narrow range of q. A database of over 1000 beam-heated discharges shows that BAEs are more likely to be unstable when the poloidal beta exceeds 0.5.

Expansion of the Operation Region of Tokamak Plasmas Using the Stationary Direct Current of a Central Solenoid

A stationary direct current of a central solenoid (DCCS) can expand the operation region of tokamak plasmas. For tokamak plasma formation experiments using electron cyclotron heating (ECH) alone, the minimum ECH power and plasma current necessary for tokamak plasma formation were reduced using the DCCS, which was applied in the counter rotational direction to the plasma current. In TST-2, the minimum ECH power necessary for the formation of a tokamak plasma was reduced from 3.3 to 1.6 kW when the DCCS was changed from 0 to 184 A. Simultaneously, the plasma current that was needed to sustain the tokamak plasma configuration was reduced to 0.6 kA.

Radial electric field and density fluctuations measured by Doppler reflectometry during the post-pellet enhanced confinement phase in W7-X

Radial profiles of density fluctuations and the radial electric field, E r, have been measured using Doppler reflectometry during the post-pellet enhanced confinement phase achieved, under different heating power levels and magnetic configurations, during the 2018 W7-X experimental campaign. A pronounced E r-well is measured with local values as high as −40 kV m−1 in the radial range ρ ∼ 0.7–0.8 during the post-pellet enhanced confinement phase. The maximum E r intensity scales with both the plasma density and electron cyclotron heating power level, following a similar trend to the plasma energy content. A good agreement is found when the experimental E r profiles are compared to simulations carried out using the neoclassical codes, the drift kinetic equation solver (DKES) and kinetic orbit-averaging solver for stellarators (KNOSOS). The density fluctuation level decreases from the plasma edge toward the plasma core and the drop is more pronounced in the post-pellet enhanced confinement phase than in reference gas-fuelled plasmas. Besides, in the post-pellet phase, the density fluctuation level is lower in the high iota magnetic configuration than in the standard one. To determine whether this difference is related to the differences in the plasma profiles or to the stability properties of the two configurations, gyrokinetic simulations have been carried out using the codes stella and EUTERPE. The simulation results point to the plasma profile evolution after the pellet injection and the stabilization effect of the radial electric field profile as the dominant players in the stabilization of the plasma turbulence.

‘BAAE’ instabilities observed without fast ion drive

The instability that was previously identified (Gorelenkov 2009 Phys. Plasmas 16 056107) as a fast-ion driven beta-induced Alfvén-acoustic eigenmode (BAAE) in DIII-D was misidentified. In a dedicated experiment, low frequency modes (LFMs) with characteristic ‘Christmas light’ patterns of brief instability linked to the safety factor evolution occur in plasmas with electron temperature T e ≳ 2.1 keV but modest beta. To isolate the importance of different driving gradients on these modes, the electron cyclotron heating (ECH) power and 80 keV, sub-Alfvénic neutral beams are altered for 50–100 ms durations in reproducible discharges. Although beta-induced Alfvén eigenmodes and reversed-shear Alfvén eigenmodes stabilize when beam injection ceases (as expected for a fast-ion driven instability), the LFMs that were called BAAEs persist. Data mining reveals that characteristic LFM instabilities can occur in discharges with no beam heating but strong ECH. A large database of over 1000 discharges shows that LFMs are only unstable in plasmas with hot electrons but modest overall beta. The experimental LFMs have low frequencies (comparable to diamagnetic drift frequencies) in the plasma frame, occur near the minimum of the safety factor q min, and appear when q min is close to rational values. Theoretical analysis suggests that the LFMs are a low frequency reactive instability of predominately Alfvénic polarization.

Effect of ECH/ECCD on energetic-particle-driven MHD modes in helical plasmas

The effect of electron cyclotron heating (ECH) and electron cyclotron current drive (ECCD) on energetic-particle (EP)-driven magnetohydrodynamic (MHD) modes is studied in the helical devices LHD, TJ-II and Heliotron J. We demonstrate that EP-driven MHD modes, including Alfvén eigenmodes (AEs) and energetic particle modes (EPMs), can be controlled by ECH/ECCD. In the LHD device, which has a moderate rotational transform and a high magnetic shear, co-ECCD enhances toroidal AEs (TAEs) and global AEs (GAEs), while counter-ECCD stabilizes them, which improves the neutron rate compared with the co-ECCD case. Counter-ECCD decreases the core rotational transform and increases the magnetic shear, strengthening the continuum damping on the shear Alfvén continua (SAC). In the TJ-II device, which has a high rotational transform, moderate magnetic shear and low toroidal field period, helical AEs (HAEs) appear when the HAE frequency gap of the SAC is changed by counter-ECCD combined with a bootstrap current and neutral-beam-driven current. On the other hand, both co- and counter-ECCD are effective in stabilizing GAEs and EPMs in the Heliotron J device, which has a low rotational transform and low magnetic shear. The experimental results indicate that the magnetic shear has a stabilizing effect regardless of its sign. Modeling analysis using the FAR3d code shows that the growth rates are reduced by both co- and counter-ECCD in Heliotron J, reproducing the experimental results. ECH only affects EP-driven MHD modes, and the experimental results show that the effect depends on the magnetic configuration. In Heliotron J, some modes are stabilized with an increase in ECH power in the low-bumpiness magnetic configuration, while some modes are destabilized in the high- and medium-bumpiness magnetic configurations.

Preliminary conceptual design of the DTT EC heating system

The Divertor Tokamak Test (DTT) facility has been proposed in the European roadmap to study solutions to mitigate the issue of power exhaust in conditions relevant for DEMO. The Italian DTT tokamak [1] is being designed to reproduce the optimal divertor magnetic configuration and test a prototype under reactor relevant power flow in the scrape off layer (PSEP/R>15 MW/m). A mix of three heating systems will equip the machine to reach the target value of 45 MW delivered at plasma: Electron Cyclotron Heating (ECH), Ion Cyclotron Heating (ICH) and Negative Neutron Beam Injector (NNBI). The present reference design considers a capability of 20–30 MW of ECH power at plasma to support and assist different tasks.The gyrotron sources (1 MW/170 GHz/100 s) will be based on the depressed collector technology with >40% efficiency and will exploit the experience gained from the solutions developed for ITER. A Transmission Line (TL) with 90% efficiency and >1 MW/beam power handling is being considered, using Quasi-Optical (QO) multi-beam mirrors. The conceptual design of equatorial and upper launchers based on the front steering concept is being developed to reach the required deposition locations for the given application. The EC wave absorption efficiency has been investigated and is presented here with dedicated beam tracing calculations using the GRAY code [2].

Active real-time control of Alfvén eigenmodes by neutral beam and electron cyclotron heating in the DIII-D tokamak

A real-time Alfvén eigenmode (AE) controller is designed and implemented for the first time in DIII-D. The experimental test consists of two parts: (i) open-loop and feedback control using neutral beam power as an actuator and (ii) open-loop control using electron cyclotron heating (ECH) as an actuator. Part (i) demonstrates the feasibility of varying neutral beam power injected into plasma to control overall AE amplitude as well as the successful application of real-time electron cyclotron emission (ECE) as the process controller’s AE amplitude sensor. The temporal evolution of AE amplitude at a specific radial location has also been controlled by selecting a subset of the radial ECE array. Part (ii) finds a complicated dependence of AE evolution on ECH power and more dedicated investigation is needed to explore the dynamical ECH-control model. Additionally, the ratio between TRANSP calculated neutron rate and measured neutron rate is found to correspond closely with AE amplitudes in discharges with feedback control, verifying that this ratio can be another sensor for the control system.

Estimation of ECH power deposition based on neural networks and fuzzy logic in plasma fusion Tokamaks

In order to stabilize magnetic hydro dynamics (MHD) activity in a Tokamaks, the measurement data acquired by different sensors along with prior information obtained from predictive plasma models are used. Suppression of plasma instabilities is a key issue to improve the confinement time of controlled thermonuclear fusion with Tokamaks. This paper proposes a method based on Self Organizing Maps (SOM) type Neural Network to estimate the Electron Cyclotron Heating (ECH) power deposition radius (rDEP) during plasma confinement. The proposed approach that is a part of the control system to stabilize MHD instability, has been compared to the Bayesian filter approach which has been proposed previously. The Bayesian approach uses on-line information acquired from Electron Cyclotron Emission (ECE) sensors and prior information got from ray-tracing code to compute the mean and standard deviation of the estimated deposition channel. The SOM approach mostly relies on ECE sensors data instead of prior information and tries to estimate the power deposition channel in real-time with less computations. A fuzzy system is also designed to reduce the uncertainty of the SOM algorithm. These algorithms have been fully compared in different aspects too. The algorithms have been tested on off-line ECE channels data, obtained from an experimental shot at Frascati Tokamak Upgrade (FTU), Frascati, Italy.

Simulation of the Plasma Density Evolution during Electron Cyclotron Resonance Heating at the T-10 Tokamak

In ohmically heated (OH) plasma with low recycling, an improved particle confinement (IPC) mode is established during gas puffing. However, after gas puffing is switched off, this mode is retained only for about 100 ms, after which an abrupt phase transition into the low particle confinement (LPC) mode occurs in the entire plasma cross section. During such a transition, energy transport due to heat conduction does not change. The phase transition in OH plasma is similar to the effect of density pump-out from the plasma core, which occurs after electron cyclotron heating (ECH) is switched on. Analysis of the measured plasma pressure profiles in the T-10 tokamak shows that, after gas puffing in the OH mode is switched off, the plasma pressure profile in the IPC stage becomes more peaked and, after the peakedness exceeds a certain critical value, the IPC-LPC transition occurs. Similar processes are also observed during ECH. If the pressure profile is insufficiently peaked during ECH, then the density pump-out effect comes into play only after the critical peakedness of the pressure profile is reached. In the plasma core, the density and pressure profiles are close to the corresponding canonical profiles. This allows one to derive an expression for the particle flux within the canonical profile model and formulate a criterion for the IPC-LPC transition. The time evolution of the plasma density profile during phase transitions was simulated for a number of T-10 shots with ECH and high recycling. The particle transport coefficients in the IPC and LPC phases, as well as the dependences of these coefficients on the ECH power, are determined.