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9月11、12日 日本九州大学TETSUO TANABE教授系列学术报告
2017-09-07|点击: | 【 【打印】【关闭】

  报告1:Tritium: Fuel of Fusion Reactors

  报告人:Tetsuo Tanabe(九州大学)

  时  间:2017年9月11日(星期一)上午9:00

  地  点:601会议室

  摘要:

  ITER (International Tokomak Experimental Reactor) will soon be ready to demonstrate the fusion reaction producing energy. Nevertheless, to realize a fusion reactor as energy source, many engineering/technological issues remain to be solved.  It is unnecessary to say that the most important point is the fusion reactor must be economically-beneficial.  The initial or capital cost to construct the fusion reactor is unavoidable, and most of the economical estimations so far performed indicate that a fusion reactor could be economical and worth repay its capital cost.  But additional costs related to tritium fuel, or to sustain fuel self-sufficiency in a D-T fusion reactor, and the trade-offs between tritium breeding and electricity production (or energy conversion) and tritium safety, do not seem small and significant effort will be required to reduce these costs. Owing to significant effort to improve plasma confinement, burning DT plasma will be realized in ITER and people are expecting a fusion reactor as an energy source.  To realize this, still lots of effort are required.

  One of the key issues is fueling and recycling of tritium (T) due to its limited resource and radioactivity.  T inventory in a reactor must be kept as small as possible for safety reason. In addition, T recovery and breeding must be efficient enough to overcome limited resources of T.  In the divertor area, fuel particle flux to the plasma facing surface is so high to implant large amount of T fuels to be immovable. In addition, at plasma shadowed area, huge amount of tritium is likely codeposited with materials originated from erosion of plasma facing materials and the T inventory in the reactor could be easily exceed the safety limitation.  Therefore, T management in a reactor is one of the most important issues to be solved.

  Although tritium handling system has been established, the established system might not be directly applied to tritium system for a fusion reactor, because the amount of tritium to be handled in the latter is much larger than the former.  In addition, owing to its limited resource, full recovery of T handled in any T sub-systems in the fusion reactor is mandatory.  For tritium safety, T removal is always required in any existing T handling systems and removed T can be disposed, whereas loss of T in the fusion reactor directly influences its fuel self-efficiency.

  The lecture consist of two parts, (1) introduction of characteristics of a fusion reactor burning tritium and (2) current status of R&D on fuel cycle (T fueling, recovering and breeding) with consideration of T safety.

  报告2:Plasma materials interactions in a fusion reactor

  报告人:Tetsuo Tanabe(九州大学)

  时  间:2017年9月12日(星期二)上午9:00

  地  点:601会议室

  摘要:

  Owing to significant effort to improve plasma confinement, burning DT plasma will be realized in ITER and people are expecting a fusion reactor as an energy source.  To realize this, still lots of effort are required.  Tritium issues will be discussed in the first lecture, here will be discussed plasma wall interactions including tolerance to high power and particle load, erosion, deposition, and hydrogen recycling.

  In a fusion reactor, energy output (17.6 MeV) is mostly carried by a neutron (n : 14 MeV) and helium (He : 3.6 MeV).  Energy carried by n is converted to heat for electric power generation in blanket and that by He is used for plasma heating.  The power used for plasma heating, i.e. nearly 1/4 of the fusion power, should be removed or exhausted from burning plasma.  Accordingly the divertor area is subjected to extraordinary high heat load by particles and radiation, which cause large erosion, evaporation and possible melting of plasma facing materials.  Materials torrent to such high heat load are very limited and only carbon and tungsten are candidate plasma facing materials of divertor area at present.  Historically, employment carbon materials as plasma facing surface gave significant improvement of plasma confinement. However concerns of large hydrogen retention in the carbon materials excluded the carbon materials as plasma facing material of ITER.

  In this lecture, properties of carbon and tungsten as the plasma facing materials and research and development of them are summarized and future prospects or criteria for the selection of plasma facing materials will be given.

  ~欢迎感兴趣的同志参加~

 
 
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