ITER极向场线圈销轴表面硬化工艺研发

ITER极向场线圈(PF)磁体支撑销轴承载极向场线圈和部分真空室的质量,因使用环境非常恶劣,所以对其表面硬度及耐磨性能有很高的需求。按照ITER极向场线圈销轴的设计要求,研发了两种表面硬化处理方式,包括无界面N离子注入361LN强化处理以及采用在不锈钢表面堆焊一层马氏体不锈钢ER414L的方式分别不同程度地增加基底材料的表面硬度以及耐磨性能,可以满足小尺寸及超大型部件的表面硬化设计需求。     

基于XTP的ITER广域网链路优化方案研究

为了满足国际热核聚变实验堆(ITER)计划全球网络不断增长的性能需求,提出了一种结合高速协议和改进传输机制的广域网链路优化解决方案.通过分析和研究快速传输协议XTP,根据XTP提供的丰富功能接口,调整burst/rate模型有效避免拥塞产生,标记分组序号改进go-back-n重传机制,同时采用协议欺骗技术减少响应延迟,分析相关协议字段实施精确QoS策略,充分利用协议的特性对网络传输实现了灵活改进.测试结果表明该方案能够显著提高ITER中法国际链路的性能.     

ITER中国液态锂铅实验包层模块活化特性分析与废料处理

使用中子学程序系统VisualBUS和活化数据库EAF-99对DFLL-TBM的高级子模块DLL-TBM的活化特性进行了计算和分析,包括DLL-TBM各部件在不同停堆时间的活度、衰变余热和剂量率.活化计算所需要的三维中子能谱通过MCNP/4C中子/光子输运程序和国际原子能机构发布的FEND1.0数据库计算得到.在活化计算分析的基础上,参照欧洲聚变堆安全和环境评估(SEAFP)策略中有关核废料的处理标准评估了TBM各区材料在退役后的废料处理工作,包括核废料应该采用何种适当的方式进行处理及其被完全清除干净的可行性.     

Options and methods for instrumentation of Test Blanket Systems for experiment control and scientific mission

Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required for their operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives.     

Status of ITER TBM port plug conceptual design and analyses

The test blanket module port plug (TBM PP) consists of a TBM frame and two TBM-sets. However, at any time of the ITER operation, a TBM set can be replaced by a dummy TBM. The frame provides a standardized interface with the vacuum vessel (VV)/port structure and provides thermal isolation from the shield blanket. As one of the plasma-facing components, it shall withstand heat loads while at the same time provide adequate neutron shielding for the VV and magnet coils. The frame design shall provide a stable engineering solution to hold TBM-sets and also provide a mean for rapid remote handling replacement and refurbishment. This paper presents main design features of the conceptual design of TBM PP with two dummy TBMs. Also analysis results are summarized to evaluate shielding, hydraulic, and thermal and structural performances of the TBM PP design.     

Development of the bus joint for the ITER Central Solenoid

The terminations of the Central Solenoid (CS) modules are connected to the bus extensions by joints located outside the CS in the gap between the CS and torodial field (TF) assemblies. These joints have very strict space limitations. Low resistance is a common requirement for all ITER joints. In addition, the CS bus joints will experience and must be designed to withstand significant variation in the magnetic field of several tenths of a Tesla per second during initiation of plasma. The joint resistance is specified to be less than 4 nΩ.The joints also have to be soldered in the field and designed with the possibility to be installed and dismantled in order to allow cold testing in the cold test facility. We have developed coaxial joints that meet these requirements and have demonstrated the feasibility to fabricate and assemble them in the vertical configuration. We introduced a coupling cylinder with superconducting strands soldered to the surface of the cable that can be installed in the ITER assembly hall and at the cold test facility. This cylinder serves as a transition area between the CS module and the bus extension.We made two racetrack samples and tested four bus joints in our Joint Test Apparatus. Resistance of the bus joints was measured by a decay method and by a microvoltmeter; the value of the current was measured by the Hall probes. This measurement method was verified in the previous tests. The resistance of the joints varied insignificantly from 1.5 to 2 nΩ.One of the challenges associated with a soldered joint is the inability to use corrosive chemicals that are difficult to clean. This paper describes our development work on cable preparation, chrome removal, compaction, soldering, and final assembly and presents the test results.     

Laser-assisted cleaning of beryllium-containing mirror samples from JET and PISCES-B

A set of seven polycrystalline mirror samples retrieved from the JET tokamak has been cleaned in vacuum using a pulsed laser system. The surfaces of samples exposed to plasma during 2008–2009 campaigns as part of the second phase of a comprehensive first mirror test contained a mixture of carbon, beryllium and tritium. For this reason, the samples were treated in a vacuum chamber constructed specially for this purpose. In some cases mirrors show an increase of the specular reflectivity after cleaning, though beryllium and carbon deposits were not fully removed. Additionally, three samples coated in PISCES-B with a 110–120 nm beryllium layer were subjected to laser cleaning tests as well.     

Determination of deuterium site occupancy in ZrCoD3 and its role in improved durability of Zr–Co–Ni deuterides against disproportionation

In the present study, we have for the first time reported the occupancy of deuterium in a new interstitial site of ZrCoD₃ which explain the hydrogen induced disproportionation behavior of ZrCo alloy. We have also reported the effect of Ni substitution on interstitial site occupancy of deuterium in ZrCo_1−xNi_xD₃, which in turn explains the improved durability of these Ni substituted deuterides against disproportionation. The crystal structure of the ZrCo_1−xNi_x (x = 0.0, 0.1, 0.2, 0.3) deuterides was investigated by X-ray powder diffraction and neutron diffraction methods. The XRD data reveals a single phase formation for all deuterides with varying Ni content (x). The neutron diffraction study shows that deuterium occupies a new site 8e in addition to 4c₂ and 8f₁. Additionally, the Zr–D distance in 8e site is shorter than that in ZrD₂. Therefore, increase in 8e site occupancy will in turn decreases the durability against disproportionation and vice-versa. Furthermore, the neutron diffraction reveals that occupancy of new 8e site decreases and its Zr–D distance increases with increase in Ni content, which explicate the higher durability against disproportionation for Ni rich compound.     

Laser damage thresholds of ITER mirror materials and first results on in situ laser cleaning of stainless steel mirrors

A laser ablation system has been constructed and used to determine the damage threshold of stainless steel, rhodium and single-, poly- and nanocrystalline molybdenum in vacuum, at a number of wavelengths between 220 nm and 1064 nm using 5 ns pulses. All materials show an increase of the damage threshold with decreasing wavelength below 400 nm. Tests in a nitrogen atmosphere showed a decrease of the damage threshold by a factor of 2–3. Cleaning tests have been performed in vacuum on stainless steel samples after applying mixed Al/W/C/D coatings using magnetron sputtering. In situ XPS analysis during the cleaning process as well ex situ reflectivity measurements demonstrate near complete removal of the coating and a substantial recovery of the reflectivity. The first results also show that the reflectivity obtained through cleaning at 532 nm may be further increased by additional exposure to UV light, in this case 230 nm, an effect which is attributed to the removal of tungsten dust from the surface.     

Overview and status of ITER internal components

The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R&D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine.