International Fusion Materials Irradiation Facility

International Fusion Materials Irradiation Facility

The International Fusion Material Irradiation Facility, also known as IFMIF, is an international scientific research program designed to test materials for suitability for use in a fusion reactor. The IFMIF, planned by Japan, the European Union, the United States, and Russia, and managed by the International Energy Agency, will use a particle accelerator-based neutron source to produce a large neutron flux, in a suitable quantity and time period to test the long-term behavior of materials under conditions similar to those expected at the inner wall of a fusion reactor.

Construction

The IFMIF will consist of two parallel accelerators, each about 50 m long, producing beams of deuterium nuclei. These, on contact with a lithium target, will be converted into high-energy neutrons and used to irradiate materials specimens and test components.Fact|date=February 2007

Preparation for IFMIF construction is expected to have begun around 2006, although operation testing of materials is not scheduled until roughly 2017. IFMIF is unlikely, therefore, to be useful in the construction of the first-generation ITER reactor, but will provide important construction information for commercial fusion reactors after ITER.

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.

Carbon

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.

Tungsten

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.

External links

* [http://www.frascati.enea.it/ifmif/ IFMIF home page]
* [http://www.iter.org/a/n1/downloads/construction_schedule.pdf ITER schedule]


Wikimedia Foundation. 2010.

Игры ⚽ Поможем сделать НИР

Look at other dictionaries:

  • Fusion power — The Sun is a natural fusion reactor. Fusion power is the power generated by nuclear fusion processes. In fusion reactions two light atomic nuclei fuse together to form a heavier nucleus (in contrast with fission power). In doing so they release a …   Wikipedia

  • International Thermonuclear Experimental Reactor — 43°41′15″N 5°45′42″E / 43.6875, 5.76167 …   Wikipédia en Français

  • National Ignition Facility — NIF s basic layout. The laser pulse is generated in the room just right of center, and is sent into the beamlines (blue) on either side. After several passes through the beamlines the light is sent into the switchyard (red) where it is aimed into …   Wikipedia

  • Nuclear fusion — Nuclear physics Radioactive decay Nuclear fission Nuclear fusion Classical dec …   Wikipedia

  • PACER (fusion) — This article is about a 1970s nuclear power experiment. For other meanings, see Pacer (disambiguation). The PACER project, carried out at Los Alamos National Laboratory in the mid 1970s, explored the possibility of a fusion power system that… …   Wikipedia

  • DIII-D (fusion reactor) — DIII D is the name of a tokamak machine developed in the 1980s by General Atomics in San Diego, USA, as part of the ongoing effort to achieve magnetically confined fusion. DIII D pioneered new technology including the use of beams of neutral… …   Wikipedia

  • ITER — International Thermonuclear Experimental Reactor Coupe du Tokamak ITER ITER (en anglais : International Thermonuclear Experimental Reactor, en français : « réacteur expérimental thermonucléaire international ») est un… …   Wikipédia en Français

  • Iter — International Thermonuclear Experimental Reactor Coupe du Tokamak ITER ITER (en anglais : International Thermonuclear Experimental Reactor, en français : « réacteur expérimental thermonucléaire international ») est un… …   Wikipédia en Français

  • ITER — is an international tokamak (magnetic confinement fusion) research/engineering proposal for an experimental project that will help to make the transition from today s studies of plasma physics to future electricity producing fusion power plants.… …   Wikipedia

  • Nova (laser) — View down Nova s laser bay between two banks of beamlines. The blue boxes contain the amplifiers and their flashtube pumps , the tubes between the banks of amplifiers are the spatial filters. Nova was a high power laser built at the Lawrence… …   Wikipedia

Share the article and excerpts

Direct link
Do a right-click on the link above
and select “Copy Link”