Reversed-field pinch studies in the Madison Symmetric Torus

Studies of large-size (R=1.5 m,a=0.5 m), moderate current (I <750 kA) reversed-field pinch (RFP) plasmas are carried out in the Madison Symmetric Torus in order to evaluate and improve RFP confinement, study general toroidal plasma MHD issues, determine the mechanism of the RFP dynamo, and measure fluctuation-induced transport and anomalous ion heating. MST confinement scaling falls short of the RFP scaling trends observed in smaller RFPs, although the plasma resistance is classical. MHD tearing modes with poloidal mode numberm=1 and toroidal mode numbersn=5–7 are prevalent and nonlinearly couple to produce sudden relaxations akin to tokamak sawteeth. Edge fluctuation-induced transport has been measured with a variety of insertable probes. Ions exhibit anomalous heating, with increases of ion temperature occurring during strong MHD relaxation. The anomalous heating fraction decreases with increasing density, such that ion temperatures approach the lower limit given by electron-ion friction. The RFP dynamo has been studied with attention to various possible mechanisms, including motion-EMF drive, the Hall effect, and superthermal electrons. The toroidal field capacity of MST will be upgraded during Summer 1993 to allow low-current tokamak operation as well as improved RFP operation.     

Peaked density profiles in neon and lithium doped discharges on FTU

Peaked density profiles are observed in FTU discharges when the recycling condition of the chamber is influenced by the action of the liquid lithium limiter (LLL) [1]. Turbulence analysis of lithium doped FTU plasmas [2], [3] has shown that the presence of the light impurity modifies the phase between fluctuating fields responsible for transport and consequently leads to an inward deuterium pinch and outward impurity flux.Analogous peaked discharges were produced by Ne-gas puffing in different L mode plasma scenarios that have been recently obtained on FTU with following plasma parameters: I = 360 kA, B = 5–6.5 T, n_e0 = 0.2–1 × 10²⁰ m⁻³, T_e0 = 1–4 keV, as well as in similar experiments on other machines [4]. In fact the Ne seeded plasmas show an increase of the peaking factor around 30% [5]. UV spectroscopy measurements confirm that the electron-density peaking arises from a convective flow and cannot be attributed to the contribution of the injected Ne alone.The Ne doped discharges analysis together with lithium conditioned ones is useful to extend the interpretative framework of the particle transport. In this work a comparison of the diffusion coefficient and of the pinch velocity of the two cases is conducted. By using a two-colors scanning interferometer providing very high spatial and time resolution, it is indeed possible to estimate the D and U coefficients of a simple model for the particle flux [6].     

Observations and Simulations of Magnetic Fluctuations in MCX

Recently magnetic fluctuations in the Maryland centrifugal experiment (MCX) have been measured by an azimuthal array of 16 probes in the edge region of the plasma. A detailed analysis indicates that there is primarily a convection of m = 2 fluctuations by the azimuthally rotating plasma. However, the frequency spectrum of this mode is broad and is almost of the same order as the frequency. Furthermore, bicoherence analysis indicates dominant nonlinear interaction between m = 2 and a low frequency m = 0 mode. We utilize a 2D MHD code to investigate the dynamics of the primary interchange instability. For very low sheared rotation there is a broad spectrum (in m) of unstable modes. However, as the sheared rotation is increased the high mode numbers become stabilized. We will present detailed comparisons of spatio-temporal dynamics of our simulations with the data from the magnetic probes.     

Recent Results in JT-60 Advanced Experiment JT-60 Team,JAERI (Presented by Naoyuki MIYA)

Major efforts in the recent JT-60 experiments have been concentrated on the improved confinement of plasmas with profile control and on the steady state operation study. Peaked density profiles were produced with the successive pellet injection. The energy confinement time was improved by 40% as large as that with the gas fuelled discharges. The fusion products n _e(0)τET _i(0)reached 1.2 × 10²⁰m^?3·s·keV, which was twice that of gas fuelled discharges. High-β_p, plasmas were obtained in low-I _p discharges with improved confinement and a high ion temperature T _i, (0) of 12 keV. The bootstrap current reached 80% of the total plasma current at β_p=3.2. The new concept of a steady-state tokamak power reactor has been proposed on the basis of this result. The maximum current drive efficiency ηCD of 3.4 × 10₁₉m^?2·MA/MW was obtained in the LH current drive experiments. Helium ash exhaust experiments using He-beam injection into H⁺ plasmas showed promising results for α-particle exhaust in a fusion power reactor.     

Radiation Emission Correlated with the Evolution of Current Sheath from a Deuterium Plasma Focus

The time resolved emission of neutrons and X-rays (both soft and hard) is correlated with the current sheath evolution during the radial phase of a 3.2 kJ Mather-type plasma focus device operated in deuterium at an optimised pressure of 4 mbar. A three-frame computer-controlled laser shadowgraphy system was incorporated in the experiment to investigate the time evolution of the radial phase of the plasma focus. The dynamics of the sheath was then correlated with the time resolved X-rays and neutron emission. The time-resolved neuron and hard X-ray emission was detected by a Scintillator-photomultiplier system while the time resolved soft X-rays were detected employing filtered PIN photo diodes. The observations were recorded with a temporal accuracy of a few ns. For the reference, the total neutron yield was also monitored by an Indium Foil activation detector. The correlation with the High Voltage Probe signal of the discharge, together with the X-ray and neutron emission regimes enabled to identify the important periods of the sheath evolution i.e. the radial compression (pre focus), minimum pinch radius (focus) and the post focus phenomena. During the initial stage of the radial phase, velocities of 10–23 cm/μs, while at the later stage of the radial phase (up till the compression), velocities up to 32–42 cm/μs were measured in our experiment. For the discharges with the lower neutron yield (lower than the average value ~1 × 10⁸ n/discharge), the current sheath appears to be disturbed and neutron and hard X-ray signal profiles do not carry much information whereas the soft X-ray emission is significant. For the discharges with high neutron yield (higher than the average value), the current sheath has a smooth structure until the maximum compression occurs. Hard X-ray emission is maximum for the discharges with high neutron yield, especially whenever there is development of m = 0 instability compressing the column to very high densities. The neutron are emitted long after the maximum compression supporting the beam target fusion. For the discharges with High neutron yield, the soft X-ray production is less as compared to the discharges with low neutron yield.     

Non-inductive Production of ST Plasmas with Washer Gun Sources on the Pegasus Toroidal Experiment

Formation of tokamak-like plasmas via electrostatic helicity injection in the ultra-low aspect ratio Pegasus Toroidal Experiment is reported. Two low-impurity, high-current (1 kA) washer gun current sources have been installed in the lower divertor region. These initially drive current along helical field lines produced by the applied toroidal and vertical fields. At sufficiently low values of externally applied vertical field, the poloidal field generated by the plasma is large enough to cause a poloidal flux reversal. In these cases the plasma relaxes into a tokamak-like configuration. Discharges with I _ϕ≈ 30 kA are produced with less than 2 kA of injected current. These discharges exhibit features indicative of tokamak plasmas, including reversal of poloidal flux at the center column, strong vacuum field deformation, increased current decay times, increased core heating, and characteristic MHD modes common to other helicity-injection-driven toroidal devices.     

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n_e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n_e/n_GW  1), to enable n_e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n_e/n_GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m²) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.     

2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

KSTAR has reached a plasma current up to 630 kA, plasma duration up to 12 s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3 s pulse durations and I_p of 611 kA) is 0.01 MW/m² while that in 2011 campaign (with 12 s pulse duration and I_p of 630 kA) is about 0.02 MW/m². The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.     

Computation of Mass Sweeping Efficiency Factor and Current Efficiency Factor in Z-pinch Devices

In this paper, mass sweeping efficiency factor (f _m ) and current efficiency factor (f _i ) have been computed for Z-pinch devices. We used slug model for analysis of Z-pinch dynamics. Magnetic piston reaps electrons and ions in duration of motion. But only a fraction of plasma mass sweeps with magnetic piston, therefore we should add mass sweeping efficiency factor (f _m ) in equations. Such like alone the fraction of electrical current flows of magnetic piston and remainder of it flows of internal and external radial of magnetic piston, so we should add f _i in equations. In this paper, equations are solved with characteristics of CERN Z-pinch device (its length and radius, resistivity, circuit inductance and capacitanc and plasma inductance) and with values of Boggasch experiments (discharge voltage: 15 kV, initial pressure: 400 pa). Recorded code runs with different values of f _m and f _i and in each section, pinch time and pinch current are compared with Boggasch experimental values. Optimum values for f _m and f _i obtain with Comparing between numerical values and experimental values. These values are f _i  = 0.8 and f _m  = 0.08.     

Preliminary Simulations of FLR Effects on RFP Tearing Modes

Preliminary simulations of finite Larmor radius (FLR) effects on m = 1 tearing instabilities in a reverse field pinch (RFP) plasma are presented. δ f particle-in-cell (PIC) simulations were performed by extending a hybrid drift kinetic-MHD model to include the full Lorentz equations of motion to take into account finite Larmor radius (FLR) effects. The simulations show that for an idealized phase space distribution, sufficiently energetic ions stabilize the tearing mode. These simulations show good agreement with analytic theory and demonstrate the FLR physics capability of the hybrid kinetic-MHD model.     

Linear and Nonlinear Development of the <Emphasis Type="Italic">m</Emphasis> = 0 Instability in Z-Pinch Equilibria with Axial Sheared Flows

The effect of sheared flows on the m = 0 instability development in a z-pinch is numerically investigated using a 2D magnetohydrodynamic (MHD) code. The behavior of both internal and free-boundary modes is studied by using two types of initial configurations: a diffuse Bennett equilibrium and a free-boundary parabolic equilibrium. It was found that sheared flows change the m = 0 development by reducing the linear growth rates, decreasing the saturation amplitude, and modifying the instability spectrum. Full stability can be obtained for supersonic plasma flows, but a larger shear is needed to obtain stabilization of free-boundary modes.     

Pinch Current and Soft X-Ray Yield Limitations by Numerical Experiments on Nitrogen Plasma Focus

The modified version of the Lee model code RADPF5-15a is used to run numerical experiments with nitrogen gas, for optimizing the nitrogen soft X-ray yield on PF-SY1. The static inductance L ₀ of the capacitor bank is progressively reduced to assess the effect on pinch current I _pinch. The experiments confirm the I _pinch, limitation effect in plasma focus, where there is an optimum L ₀ below which although the peak total current, I _peak, continues to increase progressively with progressively reduced inductance L ₀, the I _pinch and consequently the soft X-ray yield, Ysxr, of that plasma focus would not increase, but instead decreases. For the PF-SY1 with capacitance of 25 μF, the optimum L ₀ = 5 nH, at which I _pinch = 254 kA, Ysxr = 5 J; reducing L ₀ further increases neither I _pinch nor nitrogen Ysxr. The obtained results indicate that reducing the present L ₀ of the PF-SY1 device will increase the nitrogen soft X-ray yield.     

Results of the Inner Electrode Modification on the ZaP Flow Z-Pinch

The ZaP Flow Z-Pinch is a basic plasma physics experiment that uses sheared flows to stabilize an otherwise unstable configuration. The inner electrode is replaced with a larger version (15 cm diameter presently versus 10 cm previously). The goal of this modification is to increase temperature through increased adiabatic compression and to allow greater flexibility of neutral-gas injection through a greater number of gas-puff valves. Results are presented regarding the effect of neutral-gas injection characteristics and charge voltage on pinch stability. Increasing capacitor bank voltage and mass of gas injected increases stability and proximity to the machine axis. A four-chord HeNe interferometer is used to determine density at z = 0 cm and total temperature using magnetic field information from the z = 0 azimuthal array of magnetic probes. Total temperatures of 100–150 eV and densities of 2–3 × 10²² m⁻³ are calculated; temperatures are consistent with measured electron and ion temperatures.     

Controlled magnetohydrodynamic mode sustainment in the reversed-field pinch: Theory, design and experiments

A novel control system design for magnetohydrodynamic (MHD) resistive-wall mode (RWM) stabilization is developed from the viewpoint of process control. The engineering approach assumed consists of system identification, selection of feedback interconnections, and subsequently, associated feedback gain tuning. A design for general output tracking is devised, implemented and experimentally verified to be capable of sustaining MHD modes in the reversed-field pinch (RFP) machine EXTRAP-T2R. In principle, by active feedback, the plasma column boundary is forced to ‘user-specified’ helicities of prescribed amplitudes and phases. Experimental success is mainly attributed to careful identification of local magnetic field diffusion time-constants, and individual actuator channel peripheral dynamics. Addition of functionality and key features of this new MHD feedback system software might provide a versatile tool for experimental plasma dynamics and innovative MHD stability research.     

Heat flux on limiters in low-q discharges

Heat flux on the Doublet III limiters was measured with an infrared camera and thermocouples during low-q discharges. The total heat load to the limiters increases with in both Dee and circular plasmas. The peak heat flux on the limiters in low-q discharges of q_a^* ≈ 2 is ≈ 2 times higher than that in high-q discharges of q_a^* ≈ 4. Since a low-q discharge is essential in order to have a high-β tokamak reactor in the future, higher heat flux on the limiters may be an inevitable problem. It is proposed that the increase in peak heat flux during low-q discharges can be reduced by modification of the limiter to an asymmetric shape.     

The spherical pinch

This is essentially a review article covering several years of work on the spherical pinch (SP) concept of plasma formation and containment. Central to this concept is the creation of a hot plasma in the center of a sphere, plasma which is then compressed by strong imploding shock waves launched from the periphery of the vessel. The experimental program, which started with the classical cylindrical theta-pinch and continued with the inductive spherical pinch, has taken a turn, in recent times, with the discovery of the scaling laws governing spherical pinch experiments, which prescribe that high gas pressures are required for achieving fusion breakeven conditions. As a consequence, energy deposition in present spherical pinch devices is done through resistive, rather than inductive, discharges. In a pilot experimental program of modest initial condenser bank energy ( 1 KJ), we find that the instantaneous energy deposition in the central plasma can lead to temperatures of the order of 2 KeV, in agreement with the prediction of the Braginskii resistivity for such a plasma, and with the relation to the velocity of the diverging shock wave generated by the sudden deposition of energy into this plasma. Moreover, when the imploding shock waves contain the central plasma, we find the containment time to be as long as 5.4 sec and the plasma to be stable. In discharges in deuterium, neutrons are emitted close to 10⁷ per shot. From the experimental parameters of the plasma, one can derive a particle density for the shocked gas equal to 3.21×10¹⁹ cm^–3, a plasma temperature equal to 730 eV and a productn=1.73 × 10¹⁴ cm^–3· sec.Brief parts of this work are abstracted from previous works of the same author: C.A.S.I. (Canadian Aeronautics and Space Institute)Transaction,2, 21 (1969);Can. J. Phys.,58, 983 (1980);J. Fusion Energy,3, 199 (1983).     

The three RF additional heating systems for FAST

FAST (Fusion Advanced Studies Torus) is a new tokamak machine proposed by the Italian Fusion Association as a Satellite Tokamak for the ITER programme. FAST will operate with deuterium plasmas to avoid the complexity deriving from the use of tritium. Therefore burning plasma conditions, where energy density of fast ions and of charged fusion products is a significant fraction of the total plasma energy density, will be achieved by accelerating plasma ions above the half-MeV range through an Ion Cyclotron Resonance Heating (ICRH) system (P = 30 MW, f = 60–90 MHz). For long pulse Advanced Tokamak (AT) scenarios, a Lower Hybrid Current Drive (LHCD) system (P = 6 MW, f = 3.7 GHz) has been envisaged to actively control the current profile, whereas an Electron Cyclotron Resonant Heating (ECRH) system (P = 4 MW, f = 170 GHz) will provide enough RF power for MHD control.     

Interaction of a straight plasma pinch with a varying magnetic field of quadrupole configuration

The interaction of a straight plasma pinch (current up to 4 kA) with a high-frequency (~1.3 Me/see) quadrupolar magnetic field (~100 Oe) is studied by very simple methods.Translated from Atomnaya Énergiya, Vol. 18, No. 4, pp. 323–329, April, 1965     

Stabilization in the ZaP Flow Z-Pinch

The ZaP flow Z-pinch experiment at the University of Washington investigates the innovative plasma confinement concept of using sheared flows to stabilize an otherwise unstable configuration. The ZaP experiment generates an axially flowing Z-pinch that is 1 m long with a 1 cm radius with a coaxial accelerator coupled to a pinch assembly chamber. Magnetic probes measure the fluctuation levels of the azimuthal modes m = 1, 2, and 3. After assembly, the plasma is magnetically confined for an extended quiescent period where the mode activity is significantly reduced. Experimental measurements show a sheared flow profile that is coincident with the low magnetic fluctuations during the quiescent period. Recent experimental modifications produce more energetic Z-pinch plasmas that exhibit the same general behavior. The plasma equilibrium is characterized with a suite of diagnostics that measure the plasma density, magnetic field, ion and electron temperatures, in addition to plasma flow. The equilibrium is shown to satisfy radial force balance.     

Numerical analysis of MD events and preliminary thermal calculation for KTX vacuum vessel

KTX is a new reversed field pinch (RFP) magnetic confinement device which is under design in ASIPP and USTC. Major disruption (MD) events may occur in future operating process, which is simulated with the finite element (FE) method. The results present that the peaks of eddy currents on vessel and conductor shell are respectively 11.791 kA and 68.637 kA with maximum stress 67.1 MPa due to high transient electromagnetic (EM) force. It is confirmed that the structure is still strong enough to bear the electromagnetic loads even if the worst case. Besides, as KTX vacuum vessel will take the method of natural cooling for heat dissipation during plasma discharge (0.5–1.0 MA), a preliminary thermal calculation was implemented in normal condition to decide suitable time parameters such as duration and interval. It is suggested that the discharge interval should be no less than 5 min for the complete 1 MA plasma with 100 ms duration, which can guarantee the temperature of vacuum vessel below 200 °C.