Study of the tungsten sputtering source suppression by wall conditionings in the EAST tokamak

The steady fusion plasma operation is constrained by tungsten(W) material sputtering issue in the EAST tokamak. In this work, the suppression of W sputtering source has been studied by advanced wall conditionings. It is also concluded that the W sputtering yield becomes more with increasing carbon(C) content in the main deuterium(D) plasma. In EAST, the integrated use of discharge cleanings and lithium(Li) coating has positive effects on the suppression of W sputtering source. In the plasma recovery experiments, it is suggested that the W intensity is reduced by approximately 60% with the help of ～35 h Ion Cyclotron Radio Frequency Discharge Cleaning(ICRF-DC) and ～40 g Li coating after vacuum failure. The first wall covered by Li film could be relieved from the bombardment of energetic particles, and the impurity in the vessel would be removed through the particle induced desorption and isotope exchange during the discharge cleanings. In general, the sputtering yield of W would decrease from the source, on the bias of the improvement of wall condition and the mitigation of plasmawall interaction process. It lays important base of the achievement of high-parameter and longpulse plasma operation in EAST. The experiences also would be constructive for us to promote the understanding of relevant physics and basis towards the ITER-like condition.     

Vacuum and wall conditioning system on EAST

EAST is a non-circular advanced steady-state experimental device and the first entirely superconducting tokamak in the world. Vacuum system is one of the most important sub-systems of EAST device. Wall conditionings, such as baking, discharge cleaning and boronization, also play a very important role for the plasma operation. Due to Ion Cyclotron Resonance Frequency (ICRF) wall conditioning technique could be carried out in the presence of a high toroidal magnetic field, it is routinely used for wall cleanings on EAST. After the 2nd campaign in 2007, the plasma facing walls was modified to full carbon walls and vacuum system was upgraded to meet the requirement of particles exhaust. This paper will introduce the new statues of vacuum system, such as pumping, fueling and wall conditionings on EAST. Then the vacuum operation and wall conditioning in the 2008 campaign is introduced.     

Surface impurities and “clean-up” techniques in ORMAK

The detrimental role played by impurities in thermonuclear devices is now well-known. Experimental studies to identify those impurities residing on the liner, or first wall, of the Oak Ridge Tokamak (ORMAK) have been carried out in the laboratory using Auger Electron (AES) and X-ray Photoelectron (XPS) spectroscopic techniques. Additionally, liner measurements have been made in situ using a small Soft X-ray Appearance Potential Spectrometer (SXAPS). Oxygen, iron and carbon were found to be the major surface impurities and, as confirmed by plasma diagnostics, also the major plasma impurities. Glow discharge cleaning of gold and stainless steel surfaces has been studied using various gases and gas mixtures. Oxygen discharges are very effective and hydrogen moderately effective in removing carbon and hydrocarbon deposits from both types of surfaces. Other parameters involved in the contamination-decontamination process, such as pressure and temperature, have been studied using techniques to controllably contaminate surfaces with hydrocarbons.     

Wall conditions in ORMAK

ORMAK is a diffuse toroidal pinch with typical plasma currents of 100 kA, electron temperatures of 800 eV, and ion temperatures of 300 eV. The walls of the plasma region are made of stainless steel coated with an intermediate layer of platinum 0.5 μm thick and an outer 1–2 μm layer of gold. Tests with an Ion Microprobe Mass Analyzer have shown that the platinum acts to decrease diffusion of impurities from the stainless steel to the surface. Gold was chosen to inhibit the surface chemical adsorption of gases. Studies with a movable limiter lead us to believe that electron energy is lost at the plasma edge mainly via line radiation and cooling on ions, while ions are lost from the plasma by charge exchange. Thus the walls are bombarded by energetic neutrals and line radiation. To measure the sputtering rate from these neutrals, a wall sample was coated with 105 Å of radioactive gold and bombarded with neutrals from ORMAK during a day's run. No measurable sputtering was found within the counting statistics of the measurement, but surface carbon contamination of the sample prevents us from any final conclusions.     

Experiment Results about the Relationship of Edge Turbulence and Atomic Processes in the HT-7 Tokamak

Using a reciprocating Langmuir probe system, the boundary plasma behaviors were investigated before and after lithium/silicon coating. Accompanying the effective reduction of impurity radiation, strong shears of radial electric field and poloidal velocity came into being and the turbulence suppression and de-correlation were observed in the edge region of coated wall plasma. All these led to the reduction of the edge transport and improvement of plasma confinement. In the central line averaged density scanning experiments, an enhanced shear of the radial electric field was observed in the edge plasma with the increase of the density, which may account for the enhancement of the transport barrier and the improvement of particle confinement.The results suggest a close link between wall conditions and boundary plasma. In addition to the relationship, (～Te)/Te ～(～n)n/ne and θ_(～T)e(～n)e ～π, had been observed in the plasma edge region, which indicates the important role of the ionization and radiation in turbulence driving.     

掺杂石墨在高能激光束和电子束作用下的热冲击行为

石墨被广泛用于当今的托卡马克装置中 ,作为真空室第一壁和偏滤器靶板的保护材料 ,也是未来聚变堆的一种候选面对等离子体材料。其抗化学溅射性能和抗热冲击性能受到广泛关注。用高能激光束和电子束轰击实验材料 ,模拟聚变堆面对等离子体材料在等离子体破裂时的工作状态 ,考察了 4种掺杂石墨材料在热冲击下的热腐蚀规律。实验结果表明 ,石墨掺杂能有效降低材料的烧蚀率。当激光单脉冲能量密度为 491 5KJ m2 时 ,冲击频率 1 0Hz,持续辐照 3 0秒后 ,几种掺杂石墨的失重率不超过2 1 3 6mg cm2 ,表现出了比纯石墨更优良的抗热冲击性能。     

Comparison of radiation damage in silicon induced by proton andneutron irradiation

The subject of radiation damage to Si detectors induced by 24-GeV/c protons and nuclear reactor neutrons has been studied. Detectors fabricated on single-crystal silicon enriched with various impurities have been tested. Significant differences in electrically active defects have been found between the various types of material. The results of the study suggest for the first time that the widely used nonionizing energy loss (NIEL) factors are insufficient for normalization of the electrically active damage in case of oxygen- and carbon-enriched silicon detectors. It has been found that a deliberate introduction of impurities into the semiconductor can affect the radiation hardness of silicon detectors     

The Influence of Resonance Helical Field on the Z eff and Impurity Radiation in IR-T1 Tokamak

The influence of resonant helical field, RHF, on effective ion charge, Z _eff, and impurity radiations on IR-T1 tokamak discharges was studied. The theoretical calculation of Z _eff with RHF indicated that the Z _eff decreased. To observe the effects of reduced Z _eff on impurity radiation, two important parts of plasma were investigated, equilibrium region of plasma and disruptive plasma. The results obtained from previous experiments on equilibrium plasma showed the increased radiation of impurities, in comparison with preceding and next regions, whereas the new results indicate that the impurities radiation decreases remarkably in disruptive part of plasma.     

Two-dimensional modeling of disruption mitigation by gas injection

In ITER, wall damage after the disruptions can be mitigated using preventive massive gas injection (MGI) of noble gases into confined plasma. In the plasma the injected gas gets ionized and the core contamination results in fast loss of energy by radiation which distributes the wall load in a propitious way. For the modeling of MGI the tokamak code TOKES has been applied. This work reports further development of its models. The simulations earlier limited by the confinement region are expanded over the whole vessel of arbitrary wall shape. A simplified plasma model earlier employed for MGI is replaced by more adequate radiative model with dynamically changing level populations of plasma species. The processes in the plasma such as longitudinal motion, cross thermal conductivity and striking the wall are simulated neglecting wall response. The upgraded code is validated against an argon injection experiment of tokamak DIII-D.     

Description of tokamak discharges in present and future devices at Princeton PPL

The evolution in time and space of plasma temperatures and densities in typical discharges of presently operating tokamak is reviewed. The characteristics of the concomitant bombardment of vacuum walls and aperture limiters by electrons ions, neutral atoms, and photons are described together with the behavior impurities of wall and adsorbed materials in the discharge. The influence of the impurities on the plasma properties is discussed, with special consideration for cumulative effects. The discussion is extended to the extrapolated behavior of the discharges in the next generation of tokamaks at PPL.     

Radiative Cooling Rates for Low-Z Impurities in Non-coronal Equilibrium State

The local or transient radiation losses in tokamak plasmas can greatly exceed those in the coronal equilibrium. This excess is especially pronounced at the plasma edge. The reason for the increase of radiation in a peripheral plasma is as follows. The impurities are lost fast from the plasma edge and the new impurity source is supplied to this region. The charged states of impurities, therefore, do not reach their coronal equilibrium ones. These impurity ions have more electrons than those in the coronal equilibrium, and as a result emit the higher radiation power. In the simplest case, the non-coronal radiative rate can be determined only by two parameters: the electron temperature \(T_{\text {e}}\) and the so-called “residence parameter” \(n_{\text {e}}\tau _{\text {i}}\), where \(\tau _{\text {i}}\) is the impurity residence time in the plasma. Despite the strong simplification, such an approach allows to do simple estimates of non-coronal radiation. In this paper, two dimensional polynomial fits describing radiative cooling rates and mean charge are obtained for eight impurity species: helium, lithium, beryllium, carbon, nitrogen, oxygen, neon, and argon. The results are presented in figures and tables. The figures show curves calculated from the original atomic database and least-squares polynomial fits to these curves. The tables contains coefficients for this fits. The obtained fits can be useful for qualitative estimates and simple numerical calculations.     

Impurity Measurements in Plasma Boundary of JIPP T-II Tokamak/Stellarator

Impurity transport in the scrape-off layer of the JIPP T-H Tokamak/Stellarator hybrid torus machine in Institute of Plasma Physics, Nagoya University has been investigated by a probe measurement. Silicon and Cu probes were inserted into the boundary plasma during discharges. The surface of the probes was analyzed by several analytical techniques including AES, XPS, RBS and PIXE. Main metallic impurities deposited on the probes were Mo and Fe which originated from the limiters and the inner wall of the vacuum chamber. Depth concentration profile of deposited impurities was measured by AES along with sputter-etching of the surface with Ar ions. Deposited Fe impurity had the maximum concentration at the top surface of the probe, while the depth profile of Mo showed the maximum concentration at around 8 A from the top surface. Transportation energy of these impurities are discussed on the basis of the projected range data.     

Observation of coherent mode induced by a molybdenum dust on EAST

The influence of a molybdenum dust buildup on plasma edge turbulence has been studied in the EAST tokamak.The motion of the dust from the upper divertor region is detected by a fast visible CCD camera,the XUV spectrometer arrays,and the EUV spectrometer.The MoXV emission intensity sharply increases compared with the spectral lines of various ionization states of other elements,which implies that the dust particles are the molybdenum impurities.The radial distribution of Mo14+ion simulated by a simplified 1D transport model indicates that the molybdenum dust mainly deposits in the pedestal bottom region.Moreover,it is observed that the coherent mode(CM)appears at ρ=0.94 after the molybdenum impurities enter the main plasma region.The influx of molybdenum impurities results in increasing pedestal electron density and decreasing pedestal electron temperature in contrast to that before the event of impurities dropping.It is also found that the electron density gradient in the pedestal increases when the ablation of the molybdenum impurities is observed in the pedestal region.The qualitative experimental results indicate that the onset of CM is likely related to the increase of the density gradient and edge collisionality in the pedestal.In comparison to the density gradient,the enhancement of CM amplitude largely depends on the increase of the edge collisionality.     

Impurity and surface studies in ORMAK

The ORMAK tokamak has been in operation since 1971, and surface impurities problems have been pursued from the beginning. Surface studies of materials removed from ORMAK have revealed the presence of C, O, and Fe. These are also the principal impurities observed spectroscopically in plasma discharges, although numerous other elements are present in lesser amounts. Spectroscopy, X-ray measurements, plasma resistance, and fast ion scattering have been used in an effort to determine Z_eff, the effective nuclear charge ot plasma ions. All four measurements have practical difficulties leading to relatively large experimental error limits. Oxygen discharge prc-cleaning has allowed ORMAK discharges to reach higher currents and correspondingly higher ion and electron temperatures; spectroscopic studies reveal a lower level of contaminants, particularly C and N. Power measurements indicate that most of the input power strikes the walls, mostly as radiation. By varying operating parameters it is found that $\text{Z}{}_{\mathrm{eff}}\text{～}\text{I}{}_{\mathrm{p}}\text{n}{}_{\mathrm{e}}$.     

First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.     

First-principles study of hydrogen in perfect tungsten crystal

Tungsten-based materials are used as the first wall materials in ITER. Hydrogen impurities were introduced via bombarding with the reaction plasma, which are important for the behavior and stability of the tungsten wall. Using the first-principles density functional theory and planewave pseudopotential technique, we have simulated the behaviors of hydrogen atoms inside the perfect tungsten bcc lattice. The binding energies for different interstitial sites were compared to determine the optimal trapping site for the hydrogen atom inside the tungsten lattice. The diffusion barriers for hydrogen atom between nearby trapping sites and the interaction between two interstitial hydrogen atoms were also calculated. The implication of our theoretical results on the hydrogen diffusion and accumulation behavior was discussed.     

Trapping of Free Electrons in Multipole System

In this paper, the effective Parameters in the confinement and trapping of fast electrons in plasma source Such as; plasma pressure, wall material of plasma chamber and magnetic mirror rate have been investigated with using Comsol & Geant4 code. The calculations are shown that the Multicusp magnetic field was effective the pressure less than 5?mTor, and the confinement effect becomes stronger with decreasing pressure. It is equivalent to a higher yield of output ions of plasma source. The number of fast electrons trapped in the magnetic field increases with increasing magnetic field intensity and using aluminum for wall material. Optimum conditions of confinement plasma, leading to increased the hot electron density, and ionization efficiency is increased. The results of investigations have demonstrated good correspondence with theoretical calculations, therefore there is the adequacy of the developed approach and the possibility to build more effective source ion on this basis.     

Main Directions and Recent Test Modeling Results of Lithium Capillary-Pore Systems as Plasma Facing Components

At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing materials. A solid CPS filled with liquid lithium will have a high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stressinduced cracking in steady state and during plasma transitions to provide the normal operation of divertor target plates and first-wall protecting elements. These materials will not be the sources of impurities inducing an increase of Zeef and they will not be collected as dust in the divertor area and in ducts. Experiments with lithium CPS under simulating conditions of plasma disruption on a hydrogen plasma accelerator MK-200 [-(10 - 15) MJ/m^2, - 50 μs] have been performed. The formation of a shielding layer of lithium plasma and the high stability of these systems have been shown. The new lithium limiter tests on an up-graded T-11M tokamak (plasma current up to 100 kA, pulse length -0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, deposited power of the limiter are investigated in these experiments. The first results of experiments are presented.     

Radiation Heat Transfer and Vapor Shielding in a Two-Dimensional Model of an Electrothermal Plasma Source

Electrothermal (ET) plasma discharges are emerging as valuable mechanisms for pellet injection in magnetic confinement fusion reactors. They have been shown to be capable of achieving the required pellet velocities and pellet launch frequencies required for edge localized mode control. Another advantage of ET plasma discharges is their ability to simulate fusion disruption events by depositing large heat fluxes on exposed materials. A deeper understanding of the heat transfer processes occurring in ET plasma discharges will aid in this particular application. ET plasma discharges involve the passage of high currents (order of tens of kA) along the axis of a narrow, cylindrical channel. As the current passes through the channel, radiant heat is transferred from the plasma core to the capillary wall. Ablated particles eventually fill the plasma channel and the partially ionized plasma is ejected. It is well known that the ablated material separating the plasma core from the ablating surface can act as a vapor shield and limit the radiation heat flux reaching the ablating surface. In this work, the results from a two-dimensional simulation model for ET plasma discharges are presented. The simulation of the plasma in a two-dimensional domain combined with the diffusion approximation for radiation heat transfer is shown to successfully simulate the effects of the vapor shield layer that develops inside these devices.     

The use and performance of graphite and metal ISSECs in tokamak fusion reactors

Carbon and the metallic elements molybdenum, niobium, vanadium and tungsten have been considered for use as an ISSEC (Internal Spectral Shifter and Energy Converter) in tokamak fusion reactors. All five materials have been shown to reduce the radiation damage in the 316 SS structural first wall and thus increase the first wall lifetime. On a per unit thickness basis, a tungsten ISSEC is most effective in this regard followed by Mo, Nb, V and carbon ISSECs in decreasing order. If the ISSEC is restricted to transfer its heat to the first structural wall by thermal radiation only, the maximum allowable thickness a carbon ISSEC can have is limited to 9.5 cm, Mo ISSEC to 7.5 cm, Nb to 8.5 cm, V to 4.5 cm and a W ISSEC to 6 cm for a 1 MW/m² neutronic and 4 W/cm² surface heat loading. If the ISSEC is cooled by radiation plus conduction, the maximum allowable thickness goes up to 13 cm for C, 10 cm for Mo, 8 cm for V and stays the same for Nb and W. The only ISSEC material to result in an overall reduction in total blanket radioactivity at shutdown is carbon while all of the metallic ISSECs increase the total activity. On the other hand, the long term activity (at 1000 years after shutdown) is increased for Mo, Nb and V ISSECs while it is reduced by C and W. The carbon and V ISSECs reduce the energy production per fusion while Nb and W increase it slightly and Mo results in a 15–17% energy production increase. On a relative cost basis, metallic ISSECs cost 30–55 times more than a carbon ISSEC when used at the maximum thicknesses given above. Among the four metals studied, Mo is considered to be the best material for use as an ISSEC. A definitive choice between a graphite and Mo ISSEC is difficult at this time as both materials have strong positive features; carbon being superior from radioactivity, afterheat, cost and fabricability standpoint, but molybdenum being more effective in reducing the radiation damage in the first structural wall and increasing the energy multiplication in the blanket.