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.     

Evaluation of The Thermal Performance of Multi—Element Doped Graphite under Steady—State High Heat Flux

Multi-element doped graphite,GBST1308 has been developed as a plasma facing material(PFM) for high heat flux components of the HT-7U device.The thermal performance of the material under steady-state(SS) high heat flux was evaluated under actively cooling conditions,the specimens were mechanically joined to copper heat sink with supercarbon sheet as a compliant layer between the interfaces.The experiments have been performed in a facility of ACT (actively cooling test stand) with a 100kW electron gun in order to test the suitability and the loading limit of such materials.The surface temperature and bulk temperature distribtuion of the specimens were investigated.The experimental results are very encouraging that when heat flux is not more than 6 MW/m^2,the surface temperature of GBST1308 is less than 1000℃,which is the lowest,compared with IG-430U and even with CX-2002U(CFC),The primary results indicate that the mechanically-joined material system by such a proper design as thin tile.Super compliant layer,GBST as PFM and copper-alloy heat sink,can be used as divertor plater for HT-7U in the first phase.     

The Erosion and Erosion Products of Tungsten and Carbon Based Materials Bombarded by High Energy Pulse Electron Beam

High Z and low Z materials are both the candidate plasma facing materials (PFM), up to now, the typical representative of high Z materials is tungsten, and the representatives of low Z materials are carbon materials (such as graphite, C/C composite) and beryllium. Most of these materials have been used as PFM limiters     

Surface Cracking Behaviors of Ultra-Fine Grained Tungsten Under Edge Plasma Loading in the HT-7 Tokamak

Tests of the candidate plasma facing materials(PFMs) used in experimental fusion devices are essential due to the direct influence of in-situ plasma loading.A type of ultrafine grained(UFG) tungsten sintered by resistance sintering under ultra-high pressure(RSUHP) method has been exposed in the edge plasma of the HT-7 tokamak to investigate its performance under plasma loading.Under cychc edge plasma loading,the UFG tungsten develops both macro and micro cracks.The macro cracks are attributed to the low temperature brittleness of the tungsten material itself,while the micro cracks are generated from local intense power flux deposition.     

Erosion and Modifications of Tungsten-Coated Carbon and Copper Under High Heat Flux

Tungsten-coated carbon and copper was prepared by vacuum plasma spraying (VPS) and inert gas plasma spraying (IPS), respectively. W/CFC (Tungsten/Carbon Fiber-Enhanced material) coating has a diffusion barrier that consists of W and Re multi-layers pre-deposited by physical vapor deposition on carbon fiber-enhanced materials, while W/Cu coating has a graded transition interface. Different grain growth processes of tungsten coatings under stable and transient heat loads were observed, their experimental results indicated that the recrystallizing temperature of VPS-W coating was about 1400℃ and a recrystallized columnar layer of about 30μm thickness was formed by cyclic heat loads of 4 ms pulse duration. Erosion and modifications of W/CFC and W/Cu coatings under high heat load, such as microstructure changes of interface,surface plastic deformations and cracks, were investigated, and the erosion mechanism (erosion products) of these two kinds of tungsten coatings under high heat flux was also studied.     

Radiative Divertor Plasma Behavior in L- and H-Mode Discharges with Argon Injection in EAST

Introducing strong radiative impurities into divertor plasmas has been considered as an important way to mitigate the peak heat load at the divertor target plate for ITER, and will be employed in EAST for high power long pulse operations. To this end, radiative divertor experiments were explored under both low (L) and high (H)-mode confinement regimes, for the first time in EAST, with the injection of argon and its mixture (25% Ar in D 2 ). The Ar injection greatly reduced particle and heat fluxes to the divertor in L-mode discharges, achieving nearly complete detached divertor plasma regimes for both single null (SN) and double null (DN) configurations, without increasing the core impurity content. In particular, the peak heat flux was reduced by a factor of ～6, significantly reducing the intrinsic in-out divertor asymmetry for DN, as seen by both the new infra-red camera and the Langmuir probes at the divertor target. Promising results have also been obtained in the H-modes with argon seeding, demonstrating a significant increase in the frequency and decrease in the amplitude of the edge localized modes (ELMs), thus reducing both particle and heat loads caused by the ELMs. This will be further explored in the next experimental campaign with increasing heating power for long pulse operations.     

Numerical and Experimental Investigation of Electron Beam Air Plasma Properties at Moderate Pressure简

Large size of air plasma at near atmospheric pressure has specific effects in aerospace applications. In this paper, a two dimensional multi-fluid model coupled with Monte Carlo （MC） model is established, and some experiments were carried out to investigate the characteristics of electron beam air plasma at pressure of 100-170 Torr. Based on the model, the properties of electron beam air plasma are acquired. The electron density is of the order of 1016 m-3 and the longitudinal size can exceed 1.2 m. The profiles of charged particles demonstrate that the oxygen molecule is very important for air plasma and its elementary processes play a key role in plasma equilibrium processes. The potential is almost negative and a very low potential belt is observed at the edge of plasma acting as a protection shell. A series of experiments were carried out in a low pressure vacuum facility and the beam plasma densities were diagnosed. The experimental results demonstrate that electron density increased with the electron beam energy, and the relatively low pressure was favorable for gaining high density plasma. Hence in order to achieve high density and large size plasma, it requires the researchers to choose proper discharge parameters.     

The Gridless Plasma Ion Source （GIS） for Plasma Ion Assisted Optical Coating

High-quality optical coating is a key technology for modern optics. Ion-assisted deposition technology was used to improve the vaporized coating in 1980's. The GIS (gridless ion source), which is an advanced plasma source for producing a high-quality optical coating in large area, can produce a large area uniformity＞1000 mm(diameter), a high ion current density ～ 0.5mA/cm2, 20 eV ～ 200 eV energetic plasma ions and can activate reactive gas and film atoms. Now we have developed a GIS system. The GIS and the plasma ion-assisted deposition technology are investigated to achieve a high-quality optical coating. The GIS is a high power and high current source with a power of I kW ～ 7.5 kW, a current of 10 A ～ 70 A and an ion density of 200μA/cm2 ～ 500μA/cm2. Because of the special magnetic structure, the plasma-ion extraction efficiency has been improved to obtain a maximum ion density of 500μA/cm2 in the medium power (～ 4 kW) level. The GIS applied is of a special cathode structure, so that the GIS operation can be maintained under a rather low power and the lifetime of cathode will be extended. The GIS has been installed in the LPSX-1200 type box coating system. The coated TiO2, SiO2 films such as antireflective films with the system have the same performance reported by Leybold Co, 1992, along with a controllable refractive index and film structure.     

Divertor Experiments with MBI and Strong Gas Puffing on HL-2A

In the HL-2A 2004 experiment campaign, pulsed molecular beam injection （MBI） and strong hydrogen gas puffing under the divertor configuration were used for gas fueling. The experimental results show that the MBI of hydrogen can reduce the heat flux to the divertor target plate. The electron temperature measured by the Langmuir probe array decreases significantly during the injection of the molecular beam whereas the electron density increases. This indicates that the plasma pressure near the target plates tends to be constant at a new equilibrium level. In the divertor plasmas with strong hydrogen gas puffing a high plasma density up to 4.4 × 10^19 m^-3 was achieved. In addition, a phenomenon similar to the partially detached divertor regime was observed, which is being studied in open divertor tokamaks such as DIII-D to reduce the peak heat flux on the target plates near the separatrix. After a strong gas puffing the electron temperature measured on the outer divertor target plate near the separatrix decreases till below 5 eV or even lower, but that of the farther outer divertor target plate does not change obviously; and the CIII and the Ha emissions at the plasma edge decrease as expected, but the Ha emission near the Xpoint increases. These results reflects some interesting characteristics, which needs to be studied by further modeling and experiments.     

Radio Frequency Induction Plasma Spraying of Molybdenum

Radio frequency (RF) induction plasma was used to make free-standing deposition of molybdenum (Mo). The phenomena of particle melting, flattening, and stacking were investigated. The effect of process parameters such as plasma power, chamber pressure, and spray distance on the phenomena mentioned above was studied. Scanning electron microscopy (SEM) was used to analyze the plasma-processed powder, splats formed, and deposits obtained. Experimental results show that less Mo particles are spheroidized when compared to the number of spheroidized tungsten (W) particles at the same powder feed rate under the same plasma spray condition. Molten Mo particles can be suf[iciently flattened on substrate. The influence of the process parameters on the flattening behavior is not significant. Mo deposit is not as dense as W deposit, due to the splash and low impact of molten Mo particles. Oxidation of the Mo powder with a large particle size is not evident under the low pressure plasma spray.     

Efforts Made to Handling Steady-State High Heat Flux in EAST

There have been three generations divertor designed for EAST to handle steady-state high heat flux form plasma. The first generation divertor was used on the initial phase of the plasma burning. The first generation divertor was just stainless plate 5 mm in thickness bolted on supports which had been applied since 2006–2007. From 2008 to 2013 the second generation divertor has been used. The second generation divertor was graphite divertor that consisted of graphite tiles, heat sink (CuCrZr) and supports (316L). The third generation divertor was tungsten divertor with ITER like design that had been used science 2014. Now days the upper divertor is tungsten divertor (80 modules) and the lower divertor is graphite divertor (16 modules) in EAST. Tungsten divertor is able to withstand 10 MW/m² heat flux on its strike point and graphite divertor can bear 2 MW/m² under same conditions. It is very important to make every efforts to improve thermal extraction technology of divertor by comparing and practice different designs. Such efforts made in EAST can bring experiences and answers for ITER or any next divertor fusion device on nuclear phase.     

Analysis of Interaction of Disruption Induced Runaway Electrons with Plasma Facing Components in Next Generation Tokamaks

Runaway electrons which are accelerated during plasma disruptions may cause damage to the plasma facing components when their energy is deposited locally. In order to assess the possible damage of plasma facing components and the associated damage thresholds in a next generation tokamak, analyses have been carried out. The energy deposition by 100 and 300 MeV electrons in component materials has been calculated using a Monte Carlo code. The effect of parametric changes of carbon armor thickness, electron energy and angle of incidence has been evaluated. Subsequently the thermal response of divertor structures with carbon armor and with bare tungsten, and of a first wall structure has been analyzed and thresholds for thermally induced component damage were derived. The damage threshold under 100 MeV electron impact on a divertor structure with 10 mm carbon coverage and dispersion strengthened copper cooling tubes is about 60 MJ/m² of incident energy, that for a divertor structure with molybdenum coolant tubes is about 115 MJ/m², whereas the damage threshold for melting of the bare tungsten divertor is only about 30 MJ/m². Damage of the first wall structure would occur above 180 MJ/m². For 300 MeV electron incidence the damage thresholds are 13 to 47% lower than the values for 100 MeV.     

Module of lithium divertor for KTM tokamak

Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium–metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z_eff of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of steady-state operating lithium divertor module project for Kazakhstan tokamak KTM. At present the lithium divertor module for KTM tokamak is under development in the framework of ISTC project # K-1561. Initial heating up to 200 °C and lithium surface temperature stabilization during plasma interaction in the range of 350–550 °C will be provided by external system for thermal stabilization due to circulation of the Na–K heat transfer media. Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Development, creation and experimental research of lithium divertor model for KTM will allow to solve existing problems and to fulfill the basic approaches to designing of lithium divertor and in-vessel elements of new fusion reactor generation, to investigate plasma physics aspects of lithium influence, to develop technology of work with lithium in tokamak conditions. Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation.     

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.     

低能氦等离子体低温条件下钨材料的表面形貌

为研究氦等离子体在钨表面造成的表面纳米结构,利用荷兰基础能源研究所Pilot-PSI直线等离子体发生装置在673 K温度下,对钨材料进行了低能（40 eV）高束流强度（4×10²³ m^-2•s^-1）氦等离子体辐照。实验结果表明,辐照后钨材料表面出现了多种不同形态的纳米结构,表面纳米结构和晶粒的表面法向之间存在明显关联。在表面法向为［111］的晶粒表面出现三角形的纳米结构,在［110］取向的晶粒表面出现条带状的纳米结构,而在［001］取向的晶粒表面没有明显的结构出现。晶粒表面的纳米结构尺寸在50 nm左右,高度起伏在5 nm以下。另外,氦等离子体辐照会造成晶界处的高度差,在25 nm左右。分析推测氦等离子体辐照造成的晶粒表面和晶界的形貌可能是由近表面的气泡所导致的。     

Geometry and expected performance of the solid tungsten outer divertor row in JET

At JET new plasma-facing components for the main chamber wall and the divertor are being designed and built to mimic the expected ITER plasma wall conditions in the deuterium-tritium operation phase. The main wall elements at JET will be made of beryllium and the divertor plasma-facing surface will be made of tungsten. Most of the divertor tiles will consist of tungsten-coated Carbon Fibre Composite (CFC) material. However one toroidal row in the outer divertor will be made of solid, inertially cooled tungsten. The geometry of these solid tungsten divertor components is optimized within the boundary conditions of the interfaces and the constraints given by the electrodynamical forces. Shadowing calculations as well as rough field line penetration analysis is used to define the geometry of the tungsten lamella stacks. These calculations are based on a set of magnetic equilibria reflecting the operation domain of current JET plasma scenarios. All edges in poloidal and toroidal direction are shadowed to exclude near perpendicular field line impact. In addition, the geometry of the divertor structure is being optimized so that the fraction of the plasma wetted surface is maximised. On the basis of the optimized divertor geometry, performance calculations are done with the help of ANSYS to assess the maximum power exhaust possible with this inertially cooled divertor row.     

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m² plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.     

The WEST project: Testing ITER divertor high heat flux component technology in a steady state tokamak environment

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m² range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program.WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.     

The WEST programme: Minimizing technology and operational risks of a full actively cooled tungsten divertor on ITER

The WEST programme consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. This is obtained by inserting in vessel coils to create the X point while adapting the in-vessel elements to this new geometry. This will allow the full tungsten divertor technology to be used on ITER to be tested in anticipation of its use on ITER under relevant heat loading conditions and pulse duration. The early manufacturing of a significant industrial series of ITER-similar W plasma-facing units will contribute to the ITER divertor manufacturing risk mitigation and to that associated with early W divertor plasma operation on ITER.     

Design,manufacture and initial operation of the beryllium components of the JET ITER-like wall

The aim of the JET ITER-like wall project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions.The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc.) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred.Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.