High Heat Flux Components in Fusion Devices

聚变装置中的高热通量组件

• 批准号: 13107201
• 负责人: NODA Nobuaki
• 金额: $ 3.71万
• 依托单位: National Institute for Fusion Science
• 依托单位国家: 日本
• 项目类别: Grant-in-Aid for Scientific Research on Priority Areas
• 财政年份: 2001
• 资助国家: 日本
• 起止时间: 2001 至 2004
• 项目状态: 已结题
• 来源: https://kaken.nii.ac.jp/en/grant/KAKENHI-PROJECT-13107201/
• 关键词: plasma facing materials; tungsten; plasma-spla; coating; heat load; brazing; rhenium inter layer; ブラズマスプレイ; ろう付け接合

As, a plasma facing materials in future Fusion reactors, tungsten and high Z materials draw people's attention because its erosion resistance and low tritium inventory. Development and test fabrication of the tungsten coating have been conducted in this work. They have been tested against heat loads with electron beam facilities. Precise investigations have been carried out through irradiation in linear plasma devices. Main results are summarized as follows.(1) Plasma spray coating has been successfully applied to C/C composite. Typical thickness of the tungsten layer achieved is 0.5-1.0 mm. A problem of mismatching in thermal expansion coefficient has been solved by castellation of the surface of the C/C composite. Maximum size was 50 mm x 60 mm up to now.(2) Brazing of the tungsten coated carbon and bulk tungsten to oxygen free copper has been tried. The tungsten armors up to 5 mm were not destroyed up to 15 MW/m2. However, micro-cracks were found the top surface of the tungsten. More detail analyses and further development are needed to optimize the brazing.(3) Multi layered rhenium was inserted between tungsten and carbon. It has been confirmed that the layer acts as a diffusion barrier up to 1300 degree C.(4) Deuterium and helium irradiation to the tungsten surface makes change in surface morphology, in particular, hydrogen and /or helium blisters were formed in some conditions. It is confirmed that the blister formation is strongly correlated with tritium/helium inventory. More systematic studies will be necessary for this problem in future.

作为未来聚变反应堆中的等离子体面对材料，钨和高Z材料因其耐腐蚀性和低氚库存而受到人们的关注。本工作进行了钨涂层的研制和试制。他们已经测试了对热负荷与电子束设施。通过线性等离子体装置中的辐照进行了精确的研究。主要结果总结如下。(1)等离子喷涂已成功应用于C/C复合材料。钨层的典型厚度为0.5-1.0 mm。通过C/C复合材料表面的城堡化解决了热膨胀系数不匹配的问题。最大尺寸为50 mm x 60 mm。(2)对钨包碳和块体钨与无氧铜的钎焊进行了试验。5毫米以下的钨合金装甲在15兆瓦/平方米的功率下没有被破坏。然而，在钨的顶面发现了微裂纹。需要更详细的分析和进一步的发展，以优化钎焊。(3)在钨和碳之间插入了多层金属。已经证实，该层在高达1300摄氏度下充当扩散阻挡层。(4)氘和氦辐照钨表面会改变钨表面的形貌，特别是在某些条件下会形成氢和/或氦气泡。证实了气泡的形成与氚/氦存量密切相关。今后有必要对这一问题进行更系统的研究。

项目成果

Evaluation of the integrity of divertor models of tungsten or SiC/SiC composites joined with copper

钨或与铜连接的 SiC/SiC 复合材料偏滤器模型的完整性评估

• 发表时间: 2003
• 期刊: Journal of Nuclear Materials Vol.313-316
• 作者: A.Kurumada;Y.Imamura;Y.Kubota;N.Noda
• 通讯作者: N.Noda

K.Tokunaga, R.P.Doerner, R.Seraydarian, N.Noda et al.: "Modification of tungsten coated carbon by low energy and high flux deuterium irradiation"Journal of Nuclear Materials. Vol.307-311. 126-129 (2003)

K.Tokunaga、R.P.Doerner、R.Seraydarian、N.Noda 等人：“低能高通量氘辐照对钨涂层碳的改性”核材料杂志。

Modification of tungsten coated carbon by low energy and high flux deuterium irradiation

• DOI: 10.1016/s0022-3115(02)01093-0
• 发表时间: 2002-12
• 期刊: Journal of Nuclear Materials
• 影响因子: 3.1
• 作者: K. Tokunaga;R. Doerner;R. Seraydarian;N. Noda;N. Yoshida;T. Sogabe;T. Kato;B. Schedler

Blister formation and deuterium retention on tungsten exposed low energy and high flux deuterium plasma

钨暴露的低能高通量氘等离子体中的气泡形成和氘保留

• 发表时间: 2005
• 期刊: Journal of Nuclear Materials VoL337-339, Part 1
• 作者: K.Tokunaga;M.J.Baldwin;R.P.Doerner et al.
• 通讯作者: R.P.Doerner et al.

Inhomogeneous heat loading to high-Z test limiters depending upon the limiter materials

高阻抗测试限制器的热负荷不均匀，具体取决于限制器材料

• 发表时间: 2003
• 期刊: Journal of Nuclear Materials Vol.313-316
• 作者: M.Wada;T.Hirai;T.Ohgo;T.Tanabe et al.
• 通讯作者: T.Tanabe et al.