核电专业英语第四章

Unit 4UNIT 4—4.2It is often convenient to consider an infinitely large reactor, as this enables us neglect neutron diffusion and leakage. The multiplication factor is then referred to as the infinite multiplication factor, and is expressed by important four factor formula: The relationship between the two multiplication factors is:It was pointed out that for certain types of reactor enriched uranium is necessary to achieve criticality. The most important example is the pressurized water reactor which requires slightly enriched uranium with 2 to 3 percent of U. Some current British graphite moderated reactors use fuel of a similar composition. Uranium fueled fast reactors require highly enriched uranium with 25 to 50 percent of U.The process of enriching uranium involves a partial separation of the U and U so that the product has a higher concentration of U than that of natural uranium. The waste, known as the tails, has a lower concentration. two processes are available for uranium enrichment on a commercial scale. In both of them; natural uranium is converted to gaseous compound uranium hexafluoride UF and two isotopes of uranium produce two gases of slightly different density, the UF being slightly more dense than the UF. Both processes make use of slight difference in density to achieve separation of the isotopes. In the gaseous diffusion process the UF gas diffuses through a series of semi-permeable membranes, and in the centrifuge process the UF gas is spun at high rotational speed in a centrifuge.4.3 Conversion and BreedingOne important point to emerge from earlier section of this lesson is that in thermal reactors fueled with uranium, either natural or enriched, practically all the fission occurs in U. In fast reactors, which contain no moderator and in which neutron energies are much higher than in thermal reactor, U fission occurs to a small extent, but even in this type of reactor, it is U fission which predominates and sustains the chain reaction. It is nearly correct to say, therefore, that only the U in natural uranium contributes energy directly from its own fission.Although U cannot itself be used as the fuel in a nuclear reactor, it does have a vital role to play as an isotope from which new fissile fuel can be created. In a uranium fueled reactor, a significant fraction of the neutrons produced by fission, possible 30~40 percent, are captured in U and produce U by an reaction. U is the start of a radioactive decay chain which produces NP, also radioactive and PU, which has a very long half-life and almost be regarded as a stable isotope. The processes involved are:PU, as already pointed out, is fissile, its characteristics as far as fission is concerned are similar to those of U, and it can be used as the fuel in both fast and thermal reactors.Another isotope which has characteristics similar to U is TH, the only naturally occurring isotope of the element thorium. This isotope can only undergo fission with neutrons of energy greater than about 1.4MeV, so it cannot sustain a chain reactionand be used directly as a nuclear fuel. However, as a result of neutron capture in TH the following process take places:U is fission with neutrons of all energies and like U and PU it can be used as the fuel for nuclear reactors.The importance of U and TH lies in their ability to act as fertile materials from which, as a result of neutron capture in nuclear reactors, the fissile isotopes PU and U are produced. This process is known as breeding or conversion, and it provides the method whereby U and TH can be used as sources of energy through fission.An important characteristic of a nuclear reactor, particularly one which is designed to produce new fissile material by one of the processes just described as well as power, is the Ratio of the rate at which fertile material and fissile material are used up. This ratio is the breeding ratio B and is defined as the number of new fissile atoms produced in a reactor per atom of existing fissile fuel consumed by fission and neutron capture.If the breeding ratio is exactly 1 then the quantity of fissile fuel remains constant, if it is greater than l the quantity of fissile fuel increase. And if it is less than 1 the quantity of fissile fuel decreases. In order to utilize all the world's resources of U and TH it is essential that some, though not necessarily all, of the world's reactors are designed so that their breeding ratios are greater than 1 .The following simplified argument illustrates this point. Consider a reactor (or a number of reactors) whose breeding ratio is B, fueled with natural uranium, the total mass of U being M tons. If all the U is burned up, the quantity of U converted to Pu is approximately BMM tons. If this plutonium is used to provide the second charge of fuel for the reactors, and if the value of B is the same for a plutonium fueled reactor as for a uranium fueled reactor (which is not strictly true as we shall see), then the use of BM tons of 239Pu results in another B2M tons of U being converted to 239Pu, and so on. The total amount of uranium used up would be M+BM+B2M+..., If B<1, then this is equal to M/(l-B). If this results were exact, which it is not, the value of B necessary to use up all the natural uranium originally in the reactors would be about 0.993, or for all practical purposes 1. In fact it is necessary that the value of B should be slightly greater than l as the foregoing argument neglects the inevitable losses of plutonium and uranium during chemical processing, separation of fission products and manufacture of fuel elements.If the value of B is much less than l then only a fraction of the available uranium is used up, for example if B=0.75, then theoretically 2.86 percent of the natural uranium can be used, still a very small fraction which in practice would be even smaller.The dependence of the breeding ratio on other reactor parameters can be deduced from the following argument. When a neutron is absorbed in an atom of fissile fuel, that atom is consumed (in the sense used in the definition of the breeding ratio) and neutrons are produced, where is the average number of neutrons produced per neutron absorbed in the fissile fuel. For a steady chain reaction in a reactor one of these neutrons must be absorbed in another atom of fissile fuel to keep the reaction going, and according to the definition of the breeding ratio, B neutrons must be captured in the fertile material. Some neutrons will inevitably be captured in non-fuel materials,and some will leak out of the reactor. The sum of these two processes is represented by neutrons per neutron absorbed in the fissile fuel. These processes are shown in Fig.4.1. It is clear that in order to maintain a steady chain reaction the following items should be meeted.Clearly if B is to be greater than l, 77f must exceed 2 by an amount that allows for the term, whose value is likely to be about 0.2. Values of for the three fissile isotopes for fission caused by thermal and high energy neutrons are given in Table 4.1 .It will be seen from Table 4.1 that, allowing for reasonable non-fuel neutron capture and leakage, only a 233U fueled thermal reactor can achieve a breeding ratio greater than l .In fast reactors there is a considerable improvement, and reactors of this type are capable of giving breeding ratios greater than l . The value of B in a fast reactor is actually greater than that given by equation due to the effect of fast fission in 238U or 232Th, which is very slight in a thermal reactor but may be considerable in a fast reactor in which as much as 20 percent of the fission may be in the 238U(less in 232Th).4.4 Fuel CyclesFig.4.2 shows a thermal react or fueled initially with natural or enriched uranium, it is typical of the vast majority of reactors built up to the present. The breeding ratio of such a reactor is less than I so the amount of 239Pu produced is less than the amount of 235U used. When the reactor is refueled, although the 239Pu might be recycled with the depleted uranium (which contains much less than 0.715 percent of 235U), some additional fissile fuel is needed to make up the deficit. Practically all the plutonium produced in this way has been stockpiled for nuclear weapons and future reactor, so that up to the present nearly all nuclear reactors have been fueled with natural or enriched uranium and stocks of plutonium and depleted uranium are steadily accumulating in several countries.Future reactors will be fueled with 239Pu and depleted uranium, which can be regarded as pure 238U, and in such reactors 238U will be converted t0 239Pu, but only in fast reactors will the breeding ratio be greater than I and in such reactors more 239Pu will be produced than is destroyed. Fig.4.3 shows the fuel cycle of a fast reactor fueled with 239Pu and 238U in which the breeding ratio is greater than l .When the first fuel charge is unloaded from the reactor it contains more 239Pu and less 238U than it did when new, and it is contaminated with fission products which must be removed during reprocessing. Some 239Pu is available for other uses such as fueling thermal reactors, while the rest recycled. The 238U is recycled but there is less of this isotope than in the original charge, so an additional supply 238U is required for the second and subsequent fuel loadings. The important point about this fuel cycle is that the reactor is kept going by fresh supplies of non-fissile 238U, and in this way 238U can be completely converted t0 239Pu and used as a source of energy.If the reactor of Fig.4.3 had been a thermal reactor in which the breeding ratio is less than 1, it would require a continuous supply of 239Pu as well as 238U to keep it going. It is obvious that in a planned long-term nuclear power program a combination of fast and thermal reactors will be able to utilize all the world's resources of uranium.Similar ideas apply to reactors making use of the 232Th-233U breeding process.A reactor fueled initially with either 235U or 239Pu as the fissile material, and 232Th as the fertile material will produce 233U. When sufficient 233U is produced it can be used as the fuel charge in either a fast or thermal reactor with 232Th, and in such a reactor a breeding ratio greater than 1 is possible. In such a reactor the 233U can be recycled, and the only fuel requirement is 232Th, which can thus be completely used for the production of energy.In reactors which are designed for breeding as well as power production the core may be subdivided into two regions. In this type of reactor the inner region, called the core, contains nearly all the fissile material (239Pu), and it is in this part of the reactor that most of the energy is released by fission. The outer region is called the blanket and contains the fertile material (238U). At the start of the reactor's life there is very little fission in the blanket, However neutrons produced by fission in the core may diffuse into the blanket and be captured in 238U to produce 239Pu. As the operation of this reactor proceeds, fissile material builds up in the blanket and provision must be made to remove the energy released by fission from the blanket as well as from the core.Another point worthy of mention is that in any reactor designed for breeding, neutron capture in the moderator, structural materials, etc. , and neutron leakage from the core should be reduced to a minimum. This capture and leakage is the term, and it is clear from this equation that any increase of the term reduces the possible value of the breeding ratio. As an example of this, it is generally true to say that the breeding ratio in a water-moderated reactor is less than in a graphite or heavy water-moderated reactor due to the rather high capture cross-section of water. In a reactor with a reasonably high value of the breeding ratio (say between 0.8 and l.0) new fissile fuel is being produced almost as fast as it is being consumed, and the fuel in such a reactor can be used to a very high burn-up, thus prolonging the periods between refueling and reducing fuel costs.In an expanding program of nuclear power using breeder reactors to produce new fuel for later reactors, an important parameter is the doubling time, Td, which is defined as the time required for the quantity of fissile fuel in a breeder reactor to double itself. Clearly, it can be interpreted as the time required for a breeder reactor to produce enough new fuel to provide the first fuel charge for another identical reactor, and this time will control the rate at which a breeder reactor program can be expanded.。