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Diagnostics of first wall Materials in a Magnetically Confined Fusion Device by Polarization-Resolved Laser-Induced Breakdown Spectroscopy

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Plasma Science and Technology 2014年第2期16卷 149-154页

作者：赵栋烨纳扎海然吴鼎丁洪斌 School of Physics and Optical Engineering Key Laboratory of Materials Modification by LaserIon and Electron BeamsChinese Ministry of EducationDalian University of Technology

Laser-induced breakdown spectroscopy （LIBS） is a powerful analytical tool for real- time diagnostics and detection of multiple elements deposited at the first wall of magnetically confined plasma ft^sion devices. Recently, we have tested LIBS in our laboratory for application to in situ real-time diagnostics in the fusion device EAST. In this study, we applied polarization- resolved LIBS （PR-LIBS） to reduce the background continuum and enhance the resolution and sensitivity of LIBS. We used aluminium （A1）（as a substitute for Be） and the first wall materials tungsten （W） and molybdenum （Mo） to investigate polarized continuum emission and signal-to- background ratio （SBR）. A Nd：YAG laser with first, second and third harmonics was used to produce plasma. The effects of the laser polarization plane, environmental pressure and polarizer detection angle were investigated. The spectra obtained without using a polarizer （i.e. LIBS） were compared with those obtained with a polarizer （PR-LIBS）. Distribution of emission spectral intensity was observed to follow Malus＇ law with respect to variation in the angle of detection of the polarizer. The spectra obtained by PR-LIBS had a higher SBR and greater stability than those obtained by LIBS, thereby enhancing the reliability of LIBS for quantitative analyses. A comparison of A1, Mo and W showed that W exhibited a higher continuum with stronger polarization than the low-Z elements.

关键词： LIBS PR-LIBS EAST first wall laser-produced plasma

Analysis of Plasma-Driven Tritium Permeation Through the first wall of DFLL-TBM in ITER

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Plasma Science and Technology 2009年第6期11卷 730-733页

作者：宋勇黄群英倪木一 Institute of Plasma Physics Chinese Academy of Sciences School of Nuclear Science and Technology University of Science and Technology of China

Tritium permeation through the first wall （FW） from the plasma into helium coolant is evaluated for a dual-functional lithium-lead test blanket module （DFLL-TBM）. The effect of the surface conditions on the plasma facing and coolant sides, both temperature gradient and beryllium layer clad on the plasma facing side, as well as trapping in defects on the tritium permeation is considered. The results show that most of the tritium implanted in FW re-entered the plasma. The plasma-driven tritium permeation is very sensitive to the surface conditions on the plasma facing side. With a higher sticking coefficient on the plasma-facing side, the tritium permeation into helium coolant is significantly reduced. The tritium permeation is strongly reduced with a beryllium layer clad on the front side of FW. The plasma driven tritium permeation will not seriously impact the tritium safety of DFLL-TBM. Based on tritium safety, it is reasonable to clothe the beryllium layer on FW and keep the surface clean to reduce the plasma driven tritium permeation.

关键词： plasma tritium permeation first wall TBM

Effects of a liquid lithium curtain as the first wall in a fusion reactor plasma

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Chinese Physics B 2007年第11期16卷 3312-3318页

作者：李承跃邓柏权 J.P.Allain School of Physical Electronics University of Electronic Science and Technology of ChinaChengdu610054China Argonne National Laboratory 9700 South Cass AvenueArgonneIL 60439USA Southwestern Institute of Physics Chengdu 610041China

This paper explores the effect of a liquid lithium curtain on fusion reactor plasma, such curtain is utilized as the first wall for the engineering outline design of the Fusion Experimental Breeder （FEB-E）. The relationships between the surface temperature of a liquid lithium curtain and the effective plasma charge, fuel dilution and fusion power production have been derived. Results indicate that under normal operation, the evaporation of liquid lithium does not seriously affect the effective plasma charge, but effects on fuel dilution and fusion power are more sensitive. As an example, it has investigated the relationships between the liquid lithium curtain flow velocity and the rise of surface temperature based on operation scenario II of the FEB-E design with reversed shear configuration and high power density. Results show that even if the liquid lithium curtain flow velocity is as low as 0.5 m/s, the effects of evaporation from the liquid lithium curtain on plasma are negligible. In the present design, the sputtering of liquid lithium curtain and the particle removal effects of the divertor are not yet considered in detail. Further studies are in progress, and in this work implication of lithium erosion and divertor physics on fusion reactor operation are discussed.

关键词： liquid lithium curtain first wall fuel dilution effective plasma charge,plasma-wall interaction

Study of the Effects of Liquid Lithium Curtain as first wall on Plasma

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中国核科技报告 2002年 919-926页

作者： DENG Baiquan HUANG Jinhua YAN Jiancheng PENG Lilin

The final goal of fusion energy research is to make it economically competitive and the cost of electricity (COE) as low as acceptable by the energy market. Therefore the fusion plasma has to be operating with high po... 详细信息

关键词： Liquid lithium curtain first wall Fuel dilution Effective plasma charge

Effect of the ITER FW Manufacturing Process on the Microstructure and Properties of a CuCrZr Alloy

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Plasma Science and Technology 2015年第10期17卷 887-892页

作者：刘丹华王平怀宋义李前谌继明 Southwestern Institute of Physics

The first wall （FW） is one of the core components in ITER. As the heat sink material, the CuCrZr alloy shall be properly jointed with beryllium and stainless steel. At present, the grains of CuCrZr are prone to coarsen seriously in the thermal cycle process of FW manufacturing, which has become a critical issue for ITER parties. To investigate the mirostructure and mechanical properties of the optimized CuCrZr alloy in the first wall fabricating thermal cycle, simulative experiments have been done in this study. The alloy ingot was forged and hot rolled into plates, and then solid solution annealed, cold rolled and aged for strengthening. Several heat treatments were done to the CuCrZr samples, and the changes of microstructure, micro-hardness and tensile strength were investigated. The results indicated that the original elongated grains had changed into equiaxed ones, and the vickers hardness had declined to about 60 after experiencing the process of CuCrZr/316L（N） bi-metallic plate manufacturing, either by hot isostatic pressing at a higher temperature or by explosion welding followed by solution annealing. Joining Be/CuCrZr by hot isostatic pressing acts as an aging process for CuCrZr, so after the simulated heat treatment, the hardness of the alloy increased to about 110 HV and the tensile yield strength at 250℃ rose to about 170 MPa. Meanwhile, the average grain size was controlled below 200μm.

关键词： ITER first wall CuCrZr alloy hot isostatic pressing explosive weldiag

Deuterium Retention and Physical Sputtering of Low Activation Ferritic Steel

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Plasma Science and Technology 2005年第2期7卷 2737-2740页

作者： T.Hino K.Yamaguchi Y.Yamauchi Y.Hirohata K.Tsuzuki Y.Kusama Department of Nuclear Engineering Hokkaido University Sapporo060-8628 JapanDepartment of Nuclear Engineering Hokkaido UniversitySapporo060-8628 JapanDepartment of Nuclear Engineering Hokkaido UniversitySapporo060-8628 JapanDepartment of Nuclear Engineering Hokkaido UniversitySapporo060-8628 JapanJapan Atomic Energy Research Institute NakaIbaraki-ken311-0193 JapanJapan Atomic Energy Research Institute NakaIbaraki-ken311-0193 Japan

Low activation materials have to be developed toward fusion demonstrationreactors. Ferritic steel, vanadium alloy and SiC/SiC composite are candidate materials of the firstwall, vacuum vessel and blanket components, respectively. Although changes of mechanical-thermalproperties owing to neutron irradiation have been investigated so far, there is little data for theplasma material interactions, such as fuel hydrogen retention and erosion. In the present study,deuterium retention and physical sputtering of low activation ferritic steel, F82H, wereinvestigated by using deuterium ion irradiation apparatus. After a ferritic steel sample wasirradiated by 1.7 keV D^+ ions, the weight loss was measured to obtain the physical sputteringyield. The sputtering yield was 0.04, comparable to that of stainless steel. In order to obtain theretained amount of deuterium, technique of thermal desorption spectroscopy (TDS) was employed to theirradiated sample. The retained deuterium desorbed at temperature ranging from 450 K to 700 K, inthe forms of DHO, D_2, D_2O and hydrocarbons. Hence, the deuterium retained can be reduced by bakingwith a relatively low temperature. The fluence dependence of retained amount of deuterium wasmeasured by changing the ion fluence. In the ferritic steel without mechanical polish, the retainedamount was large even when the fluence was low. In such a case, a large amount of deuterium wastrapped in the surface oxide layer containing O and C. When the fluence was large, the thickness ofsurface oxide layer was reduced by the ion sputtering, and then the retained amount in the oxidelayer decreased. In the case of a high fluence, the retained amount of deuterium became comparableto that of ferritic steel with mechanical polish or SS 316 L, and one order of magnitude smallerthan that of graphite. When the ferritic steel is used, it is required to remove the surface oxidelayer for reduction of fuel hydrogen retention. Ferritic steel sample was exposed to the envi

关键词： ferritic steel fuel hydrogen retention sputtering yield first wall fusionreactor