Simulation of the Radial Electric Field in the Small Size Divertor Tokamak for Discharge with Neutral Beam Injection (NBI)

The radial electric field in the edge plasma of small size divertor tokamak can be simulated by B2SOLPS0.5.2D fluid transport code. The simulation provides the follow results: (1) Switching on and off the part of the parallel plasma viscosity driven by parallel ion diamagnetic heat flux (Bekheit in J. Fusion Energ 27(4), 338–345, 2008; Schneider et al. in Nucl. Fusion 41:387, 2001) and Counter-NBI plasma heating change profile of radial electric field significantly. (2) Switching on and off the parallel plasma viscosity driven by parallel ion diamagnetic heat flux leads to the radial electric field is toroidal magnetic field dependence (3) For the case of counter-NBI plasma heating, the switching on and off the current driven by part parallel plasma viscosity depends on the ion diamagnetic heat flux leads to the ion poloidal velocity is toroidal magnetic field B_T dependence. (4) The profile of the radial electric field in edge plasma of small size divertor tokamak is consistent with poloidal rotation velocity.     

The effect of paramagnetic shift during thermal quench on internal components in fusion devices

A plasma current disruption is usually initiated by impurity influx that causes a rapid decrease in plasma thermal stored energy (thermal quench). Thermal quench occurs in 500–2000 μs on a large device like ITER. Depending on the β value, the plasma may be either paramagnetic or diamagnetic. Thermal quench causes a large shift in paramagnetism (or diamagnetism) and a corresponding change in toroidal flux. The flux swing can be 1–2 Weber with the rate of change of the toroidal field between 25 and 150 T/s for a device like ITER. The toroidal field shift induces poloidal current in the vessel and possibly in internal components. We have developed a method for simulating the thermal quench field shift that is compatible for use with the electromagnetic simulation codes. The method is based on a radially thin shell having the shape of the last closed flux surface with poloidal current driven to duplicate the toroidal field shift. The magnitude of the current and its time history are adjusted to duplicate the flux change during a disruption thermal quench. We will present the results of using this method to simulate the induced currents in a vacuum vessel having two shells.     

Simulation of Toroidal Rotation Effect on Heat Flux Transport in the Edge Plasma of Small Size Divertor Tokamak

The effect of toroidal rotation on heat flux transport in the edge plasma of small size divertor was simulated by B2SOLP0.5.2D transport code. The main results of simulation shows that, the following: (1) the radial heat flux is strongly influenced by toroidal rotation. (2) The amplification of conduction part of radial heat flux imposes nonresilient profile of ion temperature, under which the effect of toroidal rotation on ion temperature profile is strong. (3) The ion distribution and its gradients are lower for counter-injection neutral beam than for co-injection neutral beam. (4) Reversal of toroidal rotation during using neutral beam injection result in reverses of radial electric field and E × B drift velocity. (5) The toroidal rotation strong influence on the ion temperature scale length of the ion temperature gradient (ITG). (6) Switch on and off all drifts leads to higher change in the ion density distribution in edge plasma of small size divertor tokamak when the unbalance neutral beam injection are considered (7) the comparison between radial heat flux at different momentum input shows that, the radial ion heat flux with larger ion temperature scale length in the case of co-injection neutral beam is larger than the ion heat flux with smaller ion temperature scale length in the case of counter-injection neutral beam.     

Modelling of Small Size Divertor Tokamak Plasma Edge in the Higher Confinement Regime

A version of the B2SOLPS0.5.2D fluid transport code is the new version of B2SOLPS fluid transport code, which is suited technique to simulate the edge plasma of small size divertor tokamak in the H- regime. The results of simulation provide the following: (1) the radial electric field inside the transport barrier is consistent with the neoclassical nature of the radial electric field. (2) The absolute value of the radial electric field shear at inner side of internal transport barrier is small and consistent with the value of shear before the L–H transition, while the value of shear at barrier is significantly large. (3) As a result of strong radial electric field shear and strong barrier formation the diffusion coefficient reduced by factor ~3 with respect to L-mode while ion heat conductivity reduced by factor ~22 with respect to L-mode inside the barrier. (4) The toroidal (Parallel) flux is directed along co-current direction as L-mode but at inner side of barrier is significantly large in absolute value. (5) The radial profile of toroidal rotation in vicinity of transition layer is determined by the parameter δ (width of the transition layer) depending on the collisionality and anomalous diffusion coefficient.     

Control of vacuum arc source cathode spots contraction motion by changing electromagnetic field

This paper investigates the magnetic field component impact on cathode spots motion trajectory and the mechanism of periodic contraction.Electromagnetic coils and permanent magnets were installed at the different sides of cathode surface,the photographs of cathode spots motion trajectory were captured by a camera.Increasing the number of magnets and decreasing the distance between magnets and cathode both lead to enhancing cathode spots motion velocity.Radii of cathode spots trajectory decrease gradually with the increasing of electromagnetic coil's current,from 40 mm at 0 A to 10 mm at 2.7 A.Parallel magnetic field component intensity influence the speed of cathode spots rotate motion,and perpendicular magnetic field component drives spots drift in the radial direction.Cathode spot's radial drift is controlled by changing the location of the ‘zero line' where perpendicular magnetic component shifts direction and the radius of cathode spots trajectory almost equal to ‘zero line'.     

Measurement of Plasma Energy Confinement Time in Presence of Resonant Helical Field in IR-T1 Tokamak

Plasma energy confinement time is one of the main parameters of tokamak plasma and Lawson criterion. In this paper we present an experimental method especially based on diamagnetic loop (toroidal flux loop) for measurement of this parameter in presence of resonance helical field (RHF) in IR-T1 tokamak. For this purpose a diamagnetic loop with its compensation coil constructed and installed on outer surface of the IR-T1. Also in this work we measured the plasma current and plasma voltage from the Rogowski coil and poloidal flux loop measurements. Measurement results of plasma energy confinement time with and without RHF (L = 2, L = 3, L = 2 & 3) show that the addition of a relatively small amount of RHF could be effective for improving the quality of tokamak plasma discharge by flatting the plasma current and increasing the energy confinement time.     

Simulation of Small Size Divertor Tokamak Plasma Edge Including Self-Consistent Electric Fields

The B2.SOLPES.0.5.2D code (Braams, Contrib Plasma Phys 36:276, 1996; Rozhansky and Tendler, Rev Plasma Phys 19:147, 1996) is applied for modeling SOL (Scrape off Layer) plasma in the small size divertor tokamak. Detailed distributions of the plasma heat flux and other plasma parameters in SOL, especially at the target plate of the divertor are found by modeling. The modeling results show that most of the electron heat flux and small part of ion heat flux arrive at target plate of the divertor, while, a large part of the ion heat flux and part of electron heat flux arrive at the outer wall. Also analysis of the role of poloidal E × B drifts in the redistribution of edge plasma is fulfilled.     

Simulation of Small Size Divertor Tokamak Plasma Edge in the L-regime with ion Transport Barrier

Simulations of L-regimes of small size divertor tokamak plasma edge have been performed with the B2SOLPS5.0 2D fluid transport code for wide range parameters. A conclusion has been made that, radial electric field in the vicinity and inside separatrix is near to neoclassical electric field value. The poloidal E × B drifts and compensating parallel fluxes in the scrape off layer are large in the L-regime with ITB due to steeper gradients while the qualitative pattern of the flows is similar to that of the L-mode.     

Design,Construction, and Installation of Movable Electrode Biasing System on Tokamak

The first results of the movable electrode biasing experiments performed on the IR-T1 tokamak are presented. For this purpose, a movable electrode biasing system was designed, constructed, and installed on the IR-T1 tokamak, and then the positive voltage applied to an electrode inserted inside the tokamak limiter and the plasma current, poloidal and radial components of the magnetic fields, loop voltage, and diamagnetic flux in the absence and presence of the biased electrode were measured. Results compared and discussed.     

Multi-scale interaction between tearing modes and micro-turbulence in the HL-2A plasmas

The influence of m/n=2/1(m and n are poloidal and toroidal mode numbers) tearing modes on plasma perpendicular flows and micro-fluctuations has been investigated in HL-2 A neutral beam injection heated L-mode plasmas. It is found that the local perpendicular rotation velocity and turbulence energy are modulated by the alternation between the island X-point and O-point of the naturally rotating tearing modes. Cross-correlation analysis indicates that the modulation of density fluctuations by the tearing mode is not only limited to the island region, but also occurs in the edge region near the last closed flux surface. The turbulence exhibits distinct spectral characteristics inside and outside the island region. In addition, it is observed that the particle flux near the strike point is also significantly impacted by the tearing modes. The experimental evidence reveals that there are strong core-edge interactions between the core tearing modes and the edge transport.     

Simulation of Low–High Transition by Neutral Beam Injection in Edge Plasma of Small Size Divertor Tokamak

This paper reports simulation of L–H transition by fluid transport code B2SOLPS0.5.2D at low ion plasma density on neutral beam injection (NBI) in the edge plasma of small size divertor tokamak. The simulation provides the following results: (1) the transition is possible at plasma density 2 × 10¹⁹ m^?3 with NBI at temperature heating T^heating 3.62 keV. (2) The simulation predicts the generation of large negative radial electric field E _r, which is thought to help L–H transition during NBI, is suggested in the edge plasma of small size divertor tokamak. (3) The toroidal current density in the edge plasma of small size divertor tokamak is plasma density and direction of NBI dependence. (4) Parallel flux transport by anomalous viscosity (turbulent) through separatrix leads to the variation of toroidal current density.     

Comparative Measurements of the Asymmetry Factor in Tokamaks Using the Magnetic Probes,Poloidal and Toroidal Flux Loops

Measurement of the Asymmetry factor (Shafranov parameter) is essential in tokamak plasma experiments. The purpose of this paper is comparing of the magnetic probes, poloidal flux loops, and diamagnetic loops techniques in determination of the Asymmetry factor in tokamaks. For this reason, array of magnetic probes, flux loops, and diamagnetic loop with its compensation coil, were designed, constructed, and installed on outer surface of the IR-T1 tokamak chamber, and then the Asymmetry factor and poloidal beta measured. Moreover, a few approximate values of the internal inductance for the different plasma current density profiles are also calculated. Experimental results compared.     

Current Sheath Dynamics and its Evolution Studies in Sahand Filippov Type Plasma Focus

One of the most important factors for optimizing the plasma focus device operation is the dynamics of the plasma. In this paper, we investigated the profile and dynamics of the current sheath by measuring the velocity and distribution of current sheath in Sahand as a Filippov type plasma focus device. For this purpose, the discharge is produced in pure neon gas with capacitor bank stored energies in the range of 14–50 kJ. The current sheath is monitored using two sets of magnetic probes, one with four and other with three equi-distant probe coils. These probes, installed in both radial and axial directions near the edge of the interior electrode (anode), are used for monitoring the distributions and dynamics of the current sheath. The maximum current sheath velocities at radial and axial phase are 4 ± 0.13 and 3.51 ± 0.22 (cm/μs) respectively for 0.25 Torr. The decreasing of CS velocities in going move away from anode surface is one of the our results in this paper. In this paper we conclude that the current sheath velocity at radial phase in Sahand is greater than axial phase. The effect of the neon working gas pressure and working voltage on the current sheath dynamics and its spatial evolution is investigated and presented.     

Differences Between the Toroidal and Poloidal Flux Loops in the Measurement of Plasma Position in Tokamaks

Measurement of the Asymmetry factor (Shafranov parameter) is essential in tokamak plasma experiments. The purpose of this paper is comparing of the magnetic probes, poloidal flux loops, and diamagnetic loops techniques in determination of the Asymmetry factor in tokamaks. For this reason, array of magnetic probes, flux loops, and diamagnetic loop with its compensation coil, were designed, constructed, and installed on outer surface of the IR-T1 tokamak chamber, and then the Asymmetry factor and poloidal beta measured. Moreover, a few approximate values of the internal inductance for the different plasma current density profiles are also calculated. Experimental results compared.     

The Search for Reconnection and Helicity During Formation of a Bounded Spheromak

Recent results from investigations using insertable magnetic probes at the Sustained Spheromak Physics Experiment (SSPX) [E. B. Hooper et al., Nucl. Fusion 39, 863 (1999)] are presented. Experiments were carried out during pre-programmed, constant amplitude coaxial gun current pulses, where magnetic field increases stepwise with every pulse, but eventually saturates. Magnetic traces from the probe, which is electrically isolated from the plasma and spans the flux conserver radius, indicate there is a time lag at every pulse between the response to the current rise in the open flux surfaces (intercepting the electrodes) and the closed flux surfaces (linked around the open ones). This is interpreted as the time to buildup enough helicity in the open flux surfaces before reconnecting and merging with the closed ones. Future experimental and diagnostic plans to directly estimate the helicity in the open flux surfaces and measure reconnection are briefly discussed. This work performed under the auspices of the USDOE by LLNL under contract 7405-Eng-48. The authors are grateful to Paul M. Bellan and Caltech for their continued support to SSPX with the high-speed imaging hardware, and to the Center for Magnetic Self-Organization in Laboratory and Astrophysical Plasmas (CMSO) for their continued financial support.     

Modified solution of radial nozzle with thick reinforcement in spherical vessel head subjected to radial load

When nozzles in the spherical vessel head are designed by area replacement method defined in the majority of pressure vessel code, the thickness of the reinforcements is so thick that the application of the thin shell theory may not be appropriate in the shell assembly problems. To obtain the solution of the thick reinforcement in the radial nozzle of the spherical vessel head subjected to radial load, the nature of the thick shell is introduced to the current solution for the thin shell: (1) use of the meridional moment at the junction in moment equilibrium equation instead of the meridional moment at middle plane of reinforcement, (2) omission of derivative of rotation in the meridional moment equation, (3) omission of double derivative of radial displacement in meridional moment equation. The current analysis based on the thin shell theory is found to be less conservative as the thickness of reinforcement increases when the middle plane of the reinforcement does not coincide with the middle plane of the main shell of pressure vessel head. The rotation modified method addressed in (2) above is applicable to the design of radial nozzle reinforced by thick shell with t/R≥0.1 in spherical pressure vessel head subjected to radial load.     

Relations Between the Plasma Diamagnetic Effect and Plasma Basic Parameters in IR-T1 Tokamak

Precise determination of the poloidal Beta, internal inductance, plasma energy, plasma pressure, plasma temperature, plasma resistance, plasma effective atomic number, and plasma energy confinement time are essential for tokamak experiments. In this paper an experimental method especially based on the plasma diamagnetic effect for measurements of these parameters in IR-T1 tokamak are presented. For these purposes a diamagnetic loop with its compensation coil, and also an array of magnetic probes designed, constructed, and installed on outer surface of the IR-T1. Also in this work we measured the Shafranov parameter, plasma current, plasma voltage, and the plasma density by the magnetic probes, Rogowski coil, poloidal flux loop, and the Langmuir probe measurements, respectively.     

Study and optimization of boronization in Alcator C-Mod using the Surface Science Station (S3)

A Surface Science Station (S³) on the Alcator C-Mod tokamak is used to study and optimize the location and rate of boron film deposition in situ during electron cyclotron (EC) discharge plasmas using 2.45 GHz radio-frequency (RF) heating and a mixture of helium and diborane (B₂D₆) gasses. The radial profile of boron deposition is measured with a pair of quartz microbalances (QMB) on S³, the faces of which can be rotated 360° including orientations parallel and perpendicular to the toroidal magnetic field B_T ～0.1 T. The plasma electron density is measured with a Langmuir probe, also on S³ in the vicinity of the QMBs, and typical values are ～1 × 10¹⁶ m^?3. A maximum boron deposition rate of 0.82 μg/cm²/min is obtained, which corresponds to 3.5 nm/min if the film density is that of solid boron. These deposition rates are sufficient for boron film applications between tokamak discharges. However the deposition does not peak at the EC resonance as previously assumed. Rather, deposition peaks near the upper hybrid (UH) resonance, ～5 cm outboard of the EC resonance. This has implications for RF absorption, with the RF waves being no longer damped on the electrons at the EC resonance. The previously inferred radial locations of critical erosion zones in Alcator C-Mod also need to be re-evaluated. The boron deposition profile versus major radius follows the ion flux/density profile, implying that the boron deposition is primarily ionic. The application of a vertical magnetic field (B_V ～0.01 T) was found to narrow the plasma density and boron deposition profiles near the UH resonance, thus better localizing the deposition. A Monte Carlo simulation is developed to model the boron deposition on the different QMB/tokamak surfaces. The model requires a relatively high boron ion gyroradius of ～5 mm, indicating a B⁺¹ ion temperature of ～2 eV, to match the deposition on QMB surfaces with different orientation to B_T. Additionally, the boron ion trajectories become de-magnetized at high neutral gas throughput (～0.5 Pa m³ s^?1) and pressure (～2 Pa) when the largest absolute deposition rates are measured, resulting in deposition patterns, which are independent of surface orientation to B_T in optimized conditions.     

Effects of Flow Circulation on Hot-Wire Temperature at Top and Bottom of Thermal Diffusion Column for Isotope Separation

Abstract

By using the two-dimensional rigorous numerical solution of equations of change, the change in hot-wire temperature caused by flow circulation near the top and bottom plates of thermal diffusion column with 14.7 mm-radius and 80 K cold-wall was analyzed with constant heat flux condition on hot-wire temperature boundary by parametrically changing the pressure. Flow field analysis for H²-HD (tracer level) gas mixture in total-reflux operation has revealed : (i) The magnitude of the change in hot-wire temperature is almost proportional to the square of the pressure, while the magnitude of the free convection is almost proportional to the pressure. (ii) The magnitude of the change in hot-wire temperature at the bottom is larger than that near the top plate, while the magnitude of the radial flow component is almost the same at the top and bottom, and (iii) The range of temperature change is wider than the range at the top and bottom part of the column where convective flow turns around (i.e. radial velocity components significantly exist).     

Electrical connectors for blanket modules in ITER

Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak.The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was developed in 2012 on the basis of more thoughtful analysis of bi-directional cyclic loading conditions influencing a fatigue lifetime. Detail comparative simulations of current and field patterns and subsequent analysis of the fatigue strength and technological assessment allowed make a final choice for the E-strap design in ITER.