Ignition Approach by Neutral Beam Injection Heating in Impurity Contaminated Tokamak Reactors

Available heating power by neutral beam injection in a tokamak reactor is evaluated semi-empirically. Using this estimated value, device and plasma parameters to ignite the plasma in impurity contaminated tokamak reactors are investigated. By lowering the plasma density and concurrently by enlarging the plasma minor radius or aspect ratio, the difficulty of NBI heating can be avoided, and the ignition is almost always possible both for trapped ion mode and Alcator scaling laws.     

The Principle of Plasma Boundary Identification on HL-2A Tokamak

Plasma boundary identification is a basic task for studies on equilibrium and confinement in a divertor tokamak. With the progress on the experiments after engineering experiments, the boundary identification becomes an important issue for HL-2A. In order to satisfy the requirements of preciseness, simplified measurements and quickness, the filament current method instead of solving the equilibrium equations is used to identify plasma boundary on HL-2A. The involved principle, mathematics and the progresses, which have been made with this method, are given.     

Influence of Long-Term Plasma Irradiation on Metallic First Mirrors in HT-7 Tokamak

The first mirror （FM） samples made of polycrystal （PC） stainless steel （SS）, molybdenum （Mo） and tungsten （W） were mounted at different locations in HT-7 tokamak to investigate the surface modifications caused by erosion and deposition. The optical transmission characteristics of first mirror samples were measured by a spectrophotometer. It was found that different irradiation environment had different influences on the first mirror surfaces, especially with wave antenna nearby. In addition, the erosion made the reflectivity of FM degrade to some extents as a whole. But the deposition on the mirror influences more than erosion does. Comparing the mirrors of SS, W and Mo, irradiated in the same environment, the W-mirror had the least changes in reflectivity with regularity, while the SS-mirror had most serious changes.     

A Molecular Dynamics Study on the Dust-Plasma/Wall Interactions in the EAST Tokamak

The interactions between the W nano-dust and deuterium plasma at different locations of the EAST tokamak are simulated using a molecular dynamics code. It is shown that nano-dust particles, with the radius, R d , ～5 nm, can exist for at least several nano-seconds under the interactions from the ions without being ablated in some specific places of the tokamak edge plasma, while those with R d ≥25 nm may be ablated if the plasma temperature T～ 50 eV and density n～10 19 m 3 . In addition, the collisions of tungsten nano-dust grains with a tungsten wall at 100 m/s or 1000 m/s impinging speeds are simulated. It is demonstrated that the dust will stick to the wall, and the collision will not cause substantial damage to the wall, but it may be able to cause partial destruction of the dust grains themselves depending on their incident speeds.     

Electron Heating of LHCD Plasma in HT-7 Tokamak

Electron heating via lower hybrid current drive （LHCD） has been investigated in HT-7 superconducting tokamak. Experiments show that the central electron temperature Te0, the volume averaged electron temperature 〈 Te 〉 and the peaking factor of the electron temperature QTe = Teo/〈 Te 〉 increase with the lower hybrid wave （LHW） power. Simultaneously the electron heating efficiency and the electron temperature as the function of the central line-averaged electron density （ne） and the plasma current （Ip） have also been investigated. The experimental results are in a good agreement with those of the classical collision theory and the LHW power deposition theory.     

Observation of Blobs and Holes in the Boundary Plasma of EAST Tokamak

Intermittent convective transport at the edge and in the scrape-off layer (SOL) of EAST was investigated by using fast reciprocating Langmuir probe. Holes, as part of plasma structures, were detected for the first time inside the shear layer. The amplitude probability distribution function of the turbulence is strongly skewed, with positive skewed events (“blobs”) prevailing in the SOL region and negative skewed events (“holes”) dominant inside the shear layer. The statistical properties coincide with previous observations from JET [Nuclear Fusion 49 (2009) 092002.]. The generation mechanism of blobs and holes is also discussed. In addition burst structure and dynamics character of them are also presented.     

Improvement of Plasma Performance with Lithium Wall Conditioning in Aditya Tokamak

Lithiumization of the vacuum vessel wall of the Aditya tokamak using a lithium rod exposed to glow discharge cleaning plasma has been done to understand its effect on plasma performance. After the Li-coating, an increment of ～100 eV in plasma electron temperature has been observed in most of the discharges compared to discharges without Li coating, and the shot reproducibility is considerably improved. Detailed studies of impurity behaviour and hydrogen recycling are made in the Li coated discharges by observing spectral lines of hydrogen, carbon, and oxygen in the visible region using optical fiber, an interference filter, and PMT based systems. A large reduction in O I signal (up to ～ 40% to 50%) and a 20% to 30% decrease of H α signal indicate significant reduction of wall recycling. Furthermore, VUV emissions from O V and Fe XV monitored by a grazing incidence monochromator also show the reduction. Lower Fe XV emission indicates the declined impurity penetration to the core plasma in the Li coated discharges. Significant increase of the particle and energy confinement times and the reduction of Z eff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.     

Magnetic Sensorless Sensing of Plasma Position in the Superconducting Tokamak HT-7

Magnetic sensorless sensing experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The horizontal position is calculated from the vertical field coil current and voltage without using signals of magnetic probes placed nearby a plasma. The calculations are focused on the ripple frequency component of the power supply.There is no drift problem with the time integration of magnetic probe signals. The error of the derived plasma position is lower than 2% of the plasma minor radius.     

Analyses of Passive Plasma Shutdown during Ex-vessel Loss of Coolant Accident in the First Wall/Shield Blanket of Fusion Reactor

An ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of a fusion reactor has been analyzed by a hybrid code consisting of plasma dynamics and heat transfer analysis of in-vessel components. We investigated possibility of passive plasma shutdown scenario during the accident in International Thermonuclear Experimental Reactor (ITER). The safety analysis code which we developed can treat impurity concentration from the first wall and the divertor with a transport probability into the main plasma and a time delay given as input. It was found that the plasma is passively shutdown by a density limit disruption due to beryllium release from heated first wall surfaces about 168 seconds after the LOCA, when the transport probability of beryllium from the first wall into the plasma and the time delay were assumed to be 10^?2 and the energy confinement time, respectively. At that time, the surface temperature of the outboard center (plasma facing component (PFC) with beryllium) and the temperature of the coolant tube in the first wall (stainless steel 316) reach about 1,120°C and about 1,080°C, respectively. Although the coolant tube does not melt, the copper heat sink between the PFC and the coolant tube melts before the passive shutdown. The heat sink of copper in the outboard baffle also melts before the passive shutdown, though the PFC surface of tungsten does not melt. Consequently, we have a possibility of passive plasma shutdown before the cooling tubes melt during the ex-LOCA of the first walllshield blanket in ITER, however, further studies are needed on the effects on plasma burn control, impurity release and emission of implanted D-T fuel.     

Nuclear Analysis of Blanket/Shield Design for D-D Tokamak Fusion Reactor

A study is made of the nuclear characteristics of blanket/shield design intended for the D-D tokamak reactor incorporating a cooperative effort by the University of Illinois, Brookhaven National Laboratory and Lawrence Livermore Laboratory. The reactor is characterized by high value of plasma beta (β=0.3) and low value of neutron wall loading (W _n =0.436 MW/m²). The 1 m-thick blanket is composed mainly of graphite, and the 1 m-thick shield is a combination of B₄C, Al and H₂O. Multi-dimensional calculations are carried out to confirm the results of one-dimensional calculation and to assess the problems inherent in the design.

Compared to blankets constituted of other materials, the graphite blanket possesses the characteristics of much smaller residual radioactivity, afterheat and biological hazard potential, which could possibly more than offset the lower nuclear heating that can be provided. The design requirements for the magnet shield are satisfied by the present combination and thicknesses of blanket/shield materials.     

Experimental Study of Plasma Materials' Interaction in Plasma Focus “Dena”

It is widely recognized that plasma material interaction in fusion devices is a critical issue that affects the overall machine performance. The process of material selection with a low degradation effect on the confined plasma and on the nuclear fusion production in tokamaks is of great importance. However, plasma materials' interaction studies in tokamaks have some limitations, so it is suggested that a new method for investigating this phenomenon be developed. The high capabilities of plasma focus devices allow one to simulate the interaction studies of tokamak plasma and first-wall materials. Our experiments were performed on a 90-kJ Filippov-type plasma focus, Dena. Experimental results clearly show the influence of materials on the neutron yield and neutron production mechanisms.     

托卡马克等离子体中快粒子触发的高频鱼骨模不稳定性研究

本文考虑托卡马克等离子体中快粒子的空间密度分布剖面,在中性束及电子回旋共振加热的条件下,建立了研究鱼骨模的色散关系并对鱼骨模作了数值研究。结果表明:快粒子对内扭曲模有致稳作用,且在近轴加热条件下可激发高频鱼骨模,其频率与快粒子的环向进动频率一致。鱼骨模的增长率与快粒子的密度梯度有关,一般随密度梯度的增加而增大。在高比压区间,勉强通行快粒子可驱动鱼骨模进入第二稳定区,在该区域鱼骨模是稳定的。     

The Plasma Channel Evolution Characteristics of Pulsed Flashlamps Working in an Array

This work is devoted to the study of plasma channel evolution characteristics in pulsed xenon flashlamps working in an array. Influencing factors on the plasma channel evolution process are studied, including pre-ionization pulse and neighbor flashlamps. It has been found that neighbor flashlamps affect the plasma channel by shaping the electric potential distribution, rather than by Lorentz force. Branching is observed in the plasma channels of the flashlamps in the middle of the array. Inconsistency also exists in the plasma channels of these flashlamps in different tests. The branching and inconsistency are both caused by the unique electric field distribution in these flashlamps. Besides, the pre-ionization pulse can help the main pulse plasma channel to develop more smoothly and faster, which will weaken the shock wave and benefit the mechanical strength of the flashlamp.     

Effects of Particle Confinement and Recycling on Thermally Stable Regions in D-T Tokamak Plasma

The water accumulation phenomena in the upper plenum of a PWR was investigated by using the data of the Cylindrical Core Test Facility (CCTF) test.

When the liquid inventory in the upper plenum is small, the liquid carryover rate from the upper plenum was governed by the liquid inlet flow rate. After some amount of the water was accumulated in the upper plenum, the liquid carryover rate was governed by the steam up-flow rate and the liquid inventory in the upper plenum. The liquid carryover rate was higher with the higher steam up-flow rate and the larger liquid inventory. This trend on the liquid carryover rate was also observed in small scale model test.

Based on the CCTF data, the empirical correlation on the liquid carryover rate was obtained.     

Study of Nanodispersed Iron Oxides Produced in Steel Drilling by Contracted Electric-Arc Air Plasma Torch

The optimal conditions on the plasma-forming gas flowrate, discharge current and voltage, distance between the plasma-torch nozzle and the metal plate surface for the process of penetration in and vaporization of steel plates by the contracted electric-arc air plasma torch accompanied by water quenching, were determined. The X-ray structural and phase studies as well as Mossbauer and electron microscope studies on the samples treated were performed. It was demonstrated that the vaporized elemental iron was oxidized by the oxygen present in the air plasma jet to form iron oxides （wiistite, magnetite, hematite）, which, depending on their mass ra- tios, determined the color of the iron oxide pigments, namely, beginning from light yellow, through deep yellow, light brown, deep brown, violet, red-violet, to black. A high degree of dispersity of the iron oxides is thus produced, with an averaged diameter of the particles below 500 nm, and their defective crystal structure form the basis of their potential application as components of iron-containing catalysts and pigments.     

Design and Fabrication of a Movable Langmuir Probe for Plasma Edge Parameters Measurement in the IR-T1 Tokamak

For the edge plasma parameters measurement, a movable Langmuir probe is fabricated and installed on the IR-T1 tokamak. The set-up consists of two sets of single Langmuir probes with tungsten tip movable in the radial direction. Edge plasma parameters including electron temperature, Ion density, floating potential and the corresponding radial changes are measured. Using two-point correlation technique clearly reveal that in the SOL region the poloidal propagation of floating potential fluctuation is in the direction of ion diamagnetic drift and in the edge it propagates in the electron diamagnetic drift direction.     

核乳胶在激光直接驱动内爆DT燃料离子温度诊断中应用

在神光Ⅱ基频光直接驱动内爆实验中，采用国产核乳胶测量了DD反应产生的质子产额，根据DD反应的质子产额与DT反应的中子产额的比值，诊断了DT燃料离子温度；由于核乳胶对质子的探测效率比中子探测效率高，产额比值法在低产额特别是间接驱动实验条件下诊断燃料离子温度是一种较好的方法。     

Thermohydraulic and mechanical design assessment of tube-panel first wall for Tokamak Fusion Power Reactor

A first wall structural concept cooled by high temperature and pressurized water has been proposed for the Tokamak Fusion Power Reactor (SPTR-P). Among a number of candidate design concepts, a tube-panel structure was selected for the first wall design. Stainless steel serves as the first wall structural material. The first wall is separated from the blanket wall and has a circular cross-section coolant channel since this shape is the most desirable for resisting the mechanical load due to the pressurized cooling water. Feasibility of the thermohydraulic and mechanical design has been established by analyses under steady-state operating conditions. The effect of the heat load during plasma disruptions on the thermomechanical characteristics of the first wall has been clarified. The mechanical strength of the first wall of power reactor is inadequate to withstand the thermal load expected during plasma disruption in an experimental reactor.On leave from Kawasaki Heavy Industries Ltd.     

Computing Open-Loop Optimal Control of the q-Profile in Ramp-Up Tokamak Plasmas Using the Minimal-Surface Theory

The q-profile control problem in the ramp-up phase of plasma discharges is consid- ered in this work. The magnetic diffusion partial differential equation (PDE) models the dynamics of the poloidal magnetic flux profile, which is used in this work to formulate a PDE-constrained op- timization problem under a quasi-static assumption. The minimum surface theory and constrained numeric optimization are then applied to achieve suboptimal solutions. Since the transient dy- namics is pre-given by the minimum surface theory, then this method can dramatically accelerate the solution process. In order to be robust under external uncertainties in real implementations, PID (proportional-integral-derivative) controllers are used to force the actuators to follow the computational input trajectories. It has the potential to implement in real-time for long time discharges by combining this method with the magnetic equilibrium update.