Fast-neutron reactor

A fast-neutron reactor (FNR) or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons (carrying energies of 5 MeV or greater), as opposed to thermal neutrons used in thermal-neutron reactors. Such a reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor.No fission productshave a half-lifein the range of100–210 k years ...... nor beyond 15.7 M yearsLegend for superscript symbols₡  has thermal neutron capture cross section in the range of 8–50 barnsƒ  fissilem  metastable isomer№  primarily a naturally occurring radioactive material (NORM)þ  neutron poison (thermal neutron capture cross section greater than 3k barns)†  range 4–97 y: Medium-lived fission product‡  over 200,000 y: Long-lived fission productduring the past 15 years there has been stagnation in the development of fast reactors in the industrialized countries that were involved, earlier, in intensive development of this area. All studies on fast reactors have been stopped in countries such as Germany, Italy, the United Kingdom and the United States of America and the only work being carried out is related to the decommissioning of fast reactors. Many specialists who were involved in the studies and development work in this area in these countries have already retired or are close to retirement. In countries such as France, Japan and the Russian Federation that are still actively pursuing the evolution of fast reactor technology, the situation is aggravated by the lack of young scientists and engineers moving into this branch of nuclear power. A fast-neutron reactor (FNR) or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons (carrying energies of 5 MeV or greater), as opposed to thermal neutrons used in thermal-neutron reactors. Such a reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor. Natural uranium consists mostly of three isotopes: 238U, 235U, and trace quantities of 234U, a decay product of 238U. 238U accounts for roughly 99.3% of natural uranium and undergoes fission only by fast neutrons. About 0.7% of natural uranium is 235U, which undergoes fission by neutrons of any energy, but particularly by lower-energy neutrons. When either of these isotopes undergoes fission, it releases neutrons with an energy distribution peaking around 1 to 2 MeV. The flux of higher-energy fission neutrons (> 2 MeV) is too low to create sufficient fission in 238U, and the flux of lower-energy fission neutrons (< 2 MeV) is too low to do so easily in 235U. The common solution to this problem is to slow the neutrons using a neutron moderator, which interacts with the neutrons to slow them. The most common moderator is water, which acts by elastic scattering until the neutrons reach thermal equilibrium with the water. The key to reactor design is to carefully lay out the fuel and water so the neutrons have time to slow enough to become highly reactive with the 235U, but not so far as to allow them to escape the reactor core. Although 238U does not undergo fission by the neutrons released in fission, thermal neutrons can be captured by the nucleus to transmute the uranium into 239Pu. 239Pu has a neutron cross section similar to that of 235U, and most of the atoms created this way will undergo fission from the thermal neutrons. In most reactors this accounts for as much as ⅓ of generated energy. Some 239Pu remains, and the leftover, along with leftover 238U, can be recycled during nuclear reprocessing. Water has disadvantages as a moderator. It can absorb a neutron and remove it from the reaction. It does this just enough that the concentration of 235U in natural uranium is too low to sustain the chain reaction; the neutrons lost through absorption in the water and 238U, along with those lost to the environment, results in too few left in the fuel. The most common solution to this problem is to slightly concentrate the amount of 235U in the fuel to produce enriched uranium, with the leftover 238U known as depleted uranium. Other designs use different moderators, like heavy water, that are much less likely to absorb neutrons, allowing them to run on unenriched fuel. In either case, the reactor's neutron economy is based on thermal neutrons. Although 235U and 239Pu are less sensitive to higher-energy neutrons, they still remain somewhat reactive well into the MeV range. If the fuel is enriched, eventually a threshold will be reached where there are enough fissile atoms in the fuel to maintain a chain reaction even with fast neutrons.