Evaluation of a method to reduce the redeposition of hydrogenated coatings containing carbon and tungsten in fusion reactors

The International Thermonuclear Experimental Reactor (ITER) project will be the first experimental fusion reactor with the objective to demonstrate the scientific and technological feasibility of fusion energy for commercial energy production. Erosion of materials by physical sputtering is the most fundamental of plasma-surface interactions in fusion reactor devices. Carbon and tungsten materials planned to be used in ITER divertor will be subjected to erosion, which produces local redeposition of mixed layers. Tritium retention in mixed materials is the major concern due to the limits imposed for safety reasons by nuclear licensing. The scavenging effect to reduce the redeposition phenomenon has been reported in low-pressure technical plasmas. In the present work, the minimization of co-deposits of hydrogenated C/W coatings by the injection of scavenger particles is analyzed.     

The observation of unipolar arc damage on stainless steel and TiC coatings on stainless steel

Unipolar arcing represents a plasma-surface interaction process leading to surface damage and metal impurity influx in tokamaks. It develops if the sheath potential formed in the plasma-surface contact is high enough to ignite and sustain a micro-arc. A laser-produced plasma was used for a comparison of the arc damage produced on stainless steel surfaces with that on surfaces protected with a coating of TiC a few microns thick. Smooth coatings of superior-quality TiC were produced by the activated reactive evaporation process, which is a plasma-assisted physical vapor deposition process. For a sufficiently high electron temperature in the laser-produced plasma a large number (about 300 000 cm^?2) of unipolar arc craters were observed on the stainless steel surface which had been exposed to the expanding laser-produced plasma cloud for a few hundred nanoseconds. In comparison, no similar arc craters were detected on the surfaces protected by the TiC coating which showed minimum damage limited to the laser beam impact area.     

Stand der Kernfusionsforschung

The aim of nuclear fusion research is to reproduce on earth the energy generating process of the sun and derive energy from the fusion of atomic nuclei. This is to be done by confining the fuel, a low-density hydrogen plasma, in magnetic fields and heating it to high temperatures of over 100 million degrees. Research has made major progress on the way to a burning plasma. JET, the joint European experiment, is just a factor of 6 short of achieving plasma ignition. ITER, the test reactor now being planned in international cooperation, is intended to produce the first burning plasma providing energy over a long period.     

Competition between recovery and recrystallization in two tungsten supplies according to ITER specifications

In thermonuclear fusion devices, tungsten, implemented as armour material of plasma facing components, is in direct contact with the plasma. Due to high heat flux (20 MW/m\(^{2}\) ), a premature cracking can be observed in relation with the loss of tungsten mechanical properties. It is usually attributed to two competing restoration processes: recovery and recrystallization. A recent investigation on two tungsten supplies according to ITER specifications has highlighted that hardness abatement at high temperature leads to overestimate the recrystallization fraction, which may be a consequence of the significant contribution of recovery during annealing. The present article aims at investigating this phenomenon through the use of a dedicated mean field recrystallization model that, unlike JMAK models, accounts for physical parameters at the microstructure scale such as recovery parameter or grain boundary mobility. The methodology is applied on the two tungsten supplies for ITER. It allows discriminating, for the first time, the respective contributions of recovery and recrystallization to the macroscopic softening in the high temperature range (from 1450 to \(1800\,^{\circ }{\text {C}}\)) and annealing times (0–3500 s). The approach has led to the conclusions that the two supplies merely differ from their initial (delivery) state through the stored energy, the initial recrystallized fraction and the grain size but not from intrinsic physical parameters such as recovery parameter or grain boundary mobility.

     

聚变堆非电应用的研究进展(英文)

近几十年来,ITER(国际热核实验堆)计划在为将来提供商业用电研究方面取得了较大的进展。ITER首先打算搭建一个实验聚变堆,然后再建设一个商业示范聚变堆为今后的正式商业发电做出一定的铺垫。但根据如今世界范围内的研究进展,实现聚变堆的商业发电至少要在2050年。在此之前有必要扩大聚变堆的研究范围,因此聚变堆的非电应用研究必须受到足够的重视。聚变堆的非电应用研究包括短期应用研究和长期应用研究,短期应用研究主要体现在医学、危险物的检测等领域的研究,长期应用研究主要体现在嬗变、氢的制备、空间助推器等领域的研究。综述了聚变堆在各个方面的非电应用研究,并且讨论和举证了聚变堆非电应用的可行性。     

国际热核实验反应堆（ITER）真空室的设计介绍

国际热核实验反应堆(ITER)是建造中的世界上最大的聚变反应堆,目前选址已确定在法国的卡达拉奇.这是一项国际合作计划,参加合作的六方为:欧盟、俄罗斯、日本、中国、韩国和美国.ITER设计的宗旨是"演示聚变能和平应用的科学和工艺可行性".主要介绍ITER-FEAT真空室的设计.     

不同辐照粒子下钨及钨合金辐照损伤行为的研究进展

研究核聚变、准稳态等离子体下面向等离子体材料的辐照行为,发展适合于先进实验超导托卡马克(EAST)、国际热核聚变实验堆(ITER)和中国聚变工程实验堆(CFETR)长脉冲高参数运行乃至未来聚变反应堆稳态运行的高性能面向等离子体材料是当前核聚变研究一项艰巨而又紧迫的任务。钨因具有高熔点、高导热率、低溅射腐蚀速率、高自溅射阀值以及低蒸气压和低氚滞留等优异性能,被认为是聚变装置最具有前景的面向等离子体材料。综合评述了钨及钨合金在不同辐照粒子下损伤行为的最新研究进展。粒子辐照造成的微观缺陷在钨及钨合金内部累积,辐照造成缺陷的形成和数量与钨基材料颗粒微观结构、第二相成分等密切相关,辐照缺陷情况各异。同时,辐照粒子种类、能量、剂量和温度等辐照条件都会对钨材料辐照后的形貌特征和缺陷产生重要影响。     

ITER计划和核聚变研究的未来

ITER(国际热核聚变实验实验堆)是规划建设中的一个为验证全尺寸可控核聚变技术的可行性而设计的国际托卡马克实验堆。是由美国、日本、欧盟、中国、俄罗斯、韩国和印度在内的7方共同参与合作的目前世界上最大型的国际大科学合作计划。ITER计划的参与各国将通过这项大型国际科学计划,学习和积累大型核聚变堆的研制技术,并培养人才。日本、中国等国家都计划通过参与ITER计划的建设和实验运行,缩短自身与国际核聚变研究最前沿的距离,在未来建造自己的聚变示范堆。最终实现核聚变能源的商业应用。     

Shielding design of the ITER NBI duct for nuclear and bremsstrahlung radiation

Three-dimensional Monte Carlo shielding analyses are conducted on the ITER Neutral Beam Injection (NBI) duct for the nuclear and bremsstrahlung radiation. The detailed distribution is evaluated about the nuclear heating rate and surface heat load of the NBI duct wall by the neutron and photon transport calculation. The analytical representations of these nuclear responses are established as a function of the distance from the blanket surface. It is clarified that these representations are different between the duct wall facing the plasma and that hidden from the plasma, and also between the duct wall in the blanket region and that in the vacuum vessel region. These results are very useful for the shielding design of the NBI duct wall in the nuclear fusion reactor.     

A new time-of-flight instrument capable of in situ and real-time studies of plasma-treated surfaces

We introduce a new time-of-flight (TOF) instrument that has been constructed to study the dynamics of plasma-surface interactions. The instrument uses a well-defined ion beam at a grazing incidence as a surface probe. Real-space and real-time profiles of scattered particles are created from the output of a position-sensitive detector. The set-up permits the recording of energy and angular distributions of scattered ions and neutrals. Changes in energy and angular distribution as a function of time can be used to monitor real-time and in situ the interaction between plasma and surfaces. The performance of the set-up is tested and illustrative spectra for Ar⁺ scattering from Si (1 0 0) surfaces that were subjected to different pre-treatments are shown.     

High speed plasma diagnostics for laser plasma interaction and fusion studies

Laser plasma interaction and fusion studies involve many high speed plasma diagnostics to determine the various parameters for explaining the physical processes taking place in plasma. Detection and analysis of short-term or transient radiations (X-ray and visible) are the bases for diagnosing the physical processes occurring during laser-plasma interaction or similar radiation-emitting processes. This paper reviews the development of various high speed plasma diagnostics which are not only applicable in determining the temporal, spatial and spectral properties of X-rays for this purpose but also have wide use in various other fields of research.     

聚变-裂变能源混合堆可行性及在我国核能发展中作用的分析

简要论述了核能在我国能源发展战略中的地位及聚变-裂变混合堆在核能持续发展中的重要作用。对以不久将来即可实现的ITER聚变装置作驱动堆芯、天然铀水冷裂变系统作包层的混合堆做了细致的分析。这种混合堆型可以实现GWe级净电功率输出,年造钚1 656 kg,支持2.68个同功率压水堆电站对易裂变燃料的需要。初步的经济评估说明,混合堆电的成本是同功率压水堆电成本的1.67倍;而在不计燃料成本的情况下,混合堆与压水堆组合系统电的成本是同功率压水堆电成本的1.18倍。考虑到一般压水堆需消耗大量的天然铀,加上铀浓缩成本,混合堆与压水堆组合系统电的成本,与压水堆电的成本是可以相比拟的。     

First-wall coatings for Tokamak fusion reactors

The coating of the first wall of a fusion reactor must withstand an intense neutron flux, the impact of energetic charged and neutral particles, and photons, all of which escape from the plasma. Penetration of material eroded from the first-wall coating into the plasma will have a detrimental effect on the plasma stability and energy balance. Therefore material selection for the first-wall coating of a fusion reactor represents one of the critical problem areas for the achievement of fusion power.Theoretical and experimental studies which have been conducted on the many facets of the problem are reviewed with special emphasis placed on the more critical aspects. Possible directions of future research in this area are outlined.     

Influence of non-axial magnetic field in a 3D3V Particle-In-Cell plasma model

In this article we report on results obtained using a newly developed self-consistent fully 3D Particle-In-Cell code for modelling of plasma-solid interaction.The model presented here involves a hollow cylindrical chamber opened to the plasma, with a thin cylindrical guard at the inlet and a strong external magnetic field limiting access of charged particles to the cylindrical wall. This model layout might provide more insight into processes taking place during magnetron deposition of thin films onto porous media. It is also a basis for probe diagnostics in fusion plasma research.The magnetic field is either parallel or slightly inclined with respect to the cylindrical axis. The results presented are axial and azimuthal ion current densities and cumulative distribution functions of ions impinging on the cylindrical surface for several angles of magnetic field inclination. They confirm the importance of proper alignment with magnetic field in certain geometries.Efficiency and possibilities of further extensions to the 3D model are also briefly discussed.     

Gyrotron‐Forschung und ‐Entwicklung am KIT

Gyrotron Research and Development at KIT — An Important Contribution to the Fusion Research at Europe In future nuclear fusion power stations shall contribute to the continuous generation of environmentally friendly energy production. Electron cyclotron resonance heating is an important system to heating up the fusion plasma, the microwave source of which are gyrotrons. Together with its European partners, Karlsruhe Institute of Technology (KIT) has developed the series gyrotrons for the stellarator W7‐X and is developing EU gyrotrons for the tokamak ITER. Both are generating 1 MW CW of RF output power at 140 GHz and 170 GHz, respectively. Future fusion power plants will require multi‐megawatt gyrotrons operating at above 200 GHz. KIT is looking at new concepts, systematic design approaches and key components to achieve this target. In this paper the design approach for gyrotrons, the developments of key components and the modular test concept will be presented. Additionally, a brief introduction into the new test stand is given. Together with its European partners and with its industrial partner Thales Electron Devices, it is possible to develop and to produce a major key device, the gyrotron, for plasma experiments and future power plants.     

Numerical investigation of buoyancy effects in vertical ITER-TF-CICC at normal operating conditions

Increasing the size of an experimental fusion device is an important factor towards achieving plasma ignition conditions; this could however generate new problems such as the build up of buoyancy flow in long, vertical parts of cable-in-conduit conductors (CICC). In this article, the impact of buoyancy effects to the conductor cooling behaviour is simulated for a conductor part at an inner leg of an ITER TF coil. The MAGS code system had to be extended by a two-channel thermohydraulic model based on the model of the MITHRANDIR code. Additionally, gravity was included. Using realistic nuclear heating assumptions, buoyancy flow shows no significant impact to the coil operation temperature margins, even if the thermal coupling is restricted to convective heat transfer between hole and bundle. The time constant of the buoyancy flow build up is of the same order of magnitude of a typical ITER cycle. An example with artificially enhanced nuclear heating by a factor of four is also included for the illustration of consequences of buoyancy effects to CICC vertical flow.     

Gas sources in fusion devices

Fusion plasmas are routinely produced from pure gas injected into the vacuum vessel of fusion devices at pressures in the range of millitorr. The large surface to volume ratio, characteristic of these fully instrumented devices, makes plasma density and impurity control a difficult task. The problem is further aggravated due to the large fluxes of energetic species produced by the imperfect confinement of particles and energy into the plasma and, additionally, due to the large efficiency of the plasma in pumping the gas released by the interaction of such particles with the vessel walls. Many of these processes have a direct impact in the control of the tritium inventory, of high relevance in the operation of future fusion reactors, and they are presently the focus of very active research.In the present work, the problems associated to the control of gas sources in fusion devices are addressed. First, the impact of the gas release on the particle and impurity control of the fusion plasmas is discussed. Secondly, the different types of sources that are present either before or during plasma formation are identified and finally, the techniques oriented to control the interchange of particles between the plasma and its surroundings are described.     

聚变堆中面向等离子体材料的研究进展

受控热核聚变能是公认的可以有效解决人类未来能源需求的主要途径之一,经过多年的努力,其研究已经取得很大进展,进入了从物理可行性向工程可行性的验证阶段.决定核聚变能未来发展的一个关键问题是相关的材料问题,尤其是面向等离子体材料的发展.评述了国内外目前核聚变实验装置中面向等离子体材料的研究进展.     

Near-wall electron instability of a plasma flux

Experimental data are obtained which indicate some characteristic features of the plasma-surface interaction leading to the emission of electrons from the walls of the accelerating channel and the evolution of a specific beam instability of the plasma flux. Pis’ma Zh. Tekh. Fiz. 23, 47–52 (May 26, 1997)     

中国的超导托卡马克装置HT—7U用炭基面对等离子体材料的研究

聚变能作为一种潜在的替代能源,可以为将来提清洁和无穷尽的能源供给,引起中国政府的高度重视。一项国家大科学工程“超导托卡马克HT－7U的建造”目前正在中国科学院合肥等离子体物理研究所实施。考虑到HT－7U将运行在高热负荷和近稳态的工作条件下,研究和选取面对等离子体候选材料就显得极为重要。中国科学院山西煤炭化学研究所和合肥等离子体物理研究所合作致力于HT－7U用炭基面对等离子体材料的研究和开发。本文主要介绍了聚变能的优点,有关聚变研究的一些基本概念如托卡马克,等离子体,面对等离子体材料,壁材料和等离子体间的相互作用行为等。给出了HT－7U炭基面对等离子体材料的主要结果。相关的具体研究结果将在后续文章中给予报道。