International Fusion Materials Irradiation Facility
The International Fusion Materials 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.
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.
Preparation for IFMIF construction is expected to have begun around 2006, although operational 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, such as DEMO.
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.
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.
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.
Taylor, N. P.; Pampin, R. (2006), "Activation Properties of Tungsten as a First Wall Protection in Fusion Power Plants", Fusion Engineering and Design 81 (8-14): 1333–1338, doi:10.1016/j.fusengdes.2005.05.010, retrieved 2013-03-07