IFMIF
维基百科，自由的百科全书
国际聚变材料放射测试设施 (International Fusion Material Irradiation Facility, IFMIF) 是一项国际科学研究项目, 目的是测试核聚变反应堆所用材料的可用性. IFMIF 将使用基于粒子加速器的中子源在适当的时间周期内产生大的但合适的中子流, 以测试在极端情况下材料的长期行为, 这些极端情况类似于反应堆内壁处的情况. IFMIF 将由两个平行的加速器构成, 每个长约 50 m, 用来产生氘核粒子束. 用这些粒子束撞击锂元素组成的标靶后, 可得到高能中子, 进而照射材料样本和被测试成分.
目录 |
[编辑] 参与者
IFMIF 的最初策划者是日本, 欧盟, 美国, 和俄罗斯, 管理者是 国际能源委员会|,
[编辑] 建设
IFMIF 的建设准备工作按预期已经在 2006年 开始, 尽管发挥其实际的测试功能至少被排在 2017年 之后.
[编辑] Background information
Developing materials for fusion reactors has long been recognized as a problem nearly as difficult and important as that of plasma confinement, but it has received only a fraction of the attention. The neutron flux in a fusion reactor is expected to be about 100 times that in existing pressurized water reactors. Each atom in the blanket of a fusion reactor is expected to be hit by a neutron and displaced about a hundred times before the material is replaced. Furthermore the high-energy neutrons will produce hydrogen and helium in various nuclear reactions that tends to form bubbles at grain boundaries and result in swelling, blistering or embrittlement. One also wishes to choose materials whose primary components and impurities do not result in long-lived radioactive wastes. Finally, the mechanical forces and temperatures are large, and there may be frequent cycling of both.
The problem is exacerbated because realistic material tests must expose samples to neutron fluxes of a similar level for a similar length of time as those expected in a fusion power plant. Such a neutron source is nearly as complicated and expensive as a fusion reactor itself would be. Proper materials testing will not be possible in ITER; the problem is due to be addressed by IFMIF.
The material of the plasma facing components (PFC) is a special problem. The PFC do not have to withstand large mechanical loads, so neutron damage is much less of an issue. They do have to withstand extremely large thermal loads, up to 10 MW/m², which is a difficult but solvable problem. Regardless of the material chosen, the heat flux can only be accommodated without melting if the distance from the front surface to the coolant is not more than a centimeter or two. The primary issue is the interaction with the plasma. One can choose either a low-Z material, typified by graphite although for some purposes beryllium might be chosen, or a high-Z material, usually tungsten with molybdenum as a second choice.
[编辑] 碳
If graphite is used, the gross erosion rates due to physical and chemical sputtering would be many meters per year, so one must rely on redeposition of the sputtered material. The location of the redeposition will not exactly coincide with the location of the sputtering, so one is still left with erosion rates that may be prohibitive. An even larger problem is the tritium co-deposited with the redeposited graphite. The tritium inventory in graphite layers and dust in a reactor could quickly build up to many kilograms, representing a waste of resources and a serious radiological hazard in case of an accident. The consensus of the fusion community seems to be that graphite, although a very attractive material for fusion experiments, cannot be the primary PFC material in a commercial reactor.
[编辑] 钨
The sputtering rate of tungsten can be orders of magnitude smaller than that of carbon, and tritium is not so easily incorporated into redeposited tungsten, making this a more attractive choice. On the other hand, tungsten impurities in a plasma are much more damaging than carbon impurities, and self-sputtering of tungsten can be high, so it will be necessary to ensure that the plasma in contact with the tungsten is not too hot (a few tens of eV rather than hundreds of eV). Tungsten also has disadvantages in terms of eddy currents and melting in off-normal events, as well as some radiological issues.
[编辑] 外部链接
Template:Fusion experiments
