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nuclear reactor,device for producing controlled release of nuclear energynuclear energy,
the energy stored in the nucleus of an atom and released through fission, fusion, or radioactivity. In these processes a small amount of mass is converted to energy according to the relationship E = mc2, where E is energy, m
..... Click the link for more information. . Reactors can be used for research or for power production. A research reactor is designed to produce various beams of radiation for experimental application; the heat produced is a waste product and is dissipated as efficiently as possible. In a power reactor the heat produced is of primary importance for use in driving conventional heat engines; the beams of radiation are controlled by shielding.
A fission reactor consists basically of a mass of fissionable material usually encased in shielding and provided with devices to regulate the rate of fission and an exchange system to extract the heat energy produced. A reactor is so constructed that fission of atomic nuclei produces a self-sustaining nuclear chain reactionchain reaction,
self-sustaining reaction that, once started, continues without further outside influence. Proper conditions for a chain reaction depend not only on various external factors, such as temperature, but also on the quantity and shape of the substance undergoing the
..... Click the link for more information. , in which the neutrons produced are able to split other nuclei. A chain reaction can be produced in a reactor by using uranium or plutonium in which the concentration of fissionable isotopes has been artificially increased. Even though the neutrons move at high velocities, the enriched fissionable isotope captures enough neutrons to make possible a self-sustaining chain reaction. In this type of reactor the neutrons carrying on the chain reaction are fast neutrons.
A chain reaction can also be accomplished in a reactor by employing a substance called a moderator to retard the neutrons so that they may be more easily captured by the fissionable atoms. The neutrons carrying on the chain reaction in this type of reactor are slow (or thermal) neutrons. Substances that can be used as moderators include graphite, beryllium, and heavy water (deuteriumdeuterium
, isotope of hydrogen with mass no. 2. The deuterium nucleus, called a deuteron, contains one proton and one neutron. Deuterium is also called heavy hydrogen, and water in which the hydrogen atoms are deuterium is called heavy water (deuterium oxide, D2O).
..... Click the link for more information. oxide). The moderator surrounds or is mixed with the fissionable fuel elements in the core of the reactor.
Types of Fission Reactors
A nuclear reactor is sometimes called an atomic pile because a reactor using graphite as a moderator consists of a pile of graphite blocks with rods of uranium fuel inserted into it. Reactors in which the uranium rods are immersed in a bath of heavy water are often referred to as "swimming-pool" reactors. Reactors of these types, in which discrete fuel elements are surrounded by a moderator, are called heterogeneous reactors. If the fissionable fuel elements are intimately mixed with a moderator, the system is called a homogeneous reactor (e.g., a reactor having a core of a liquid uranium compound dissolved in heavy water).
The breeder reactor is a special type used to produce more fissionable atoms than it consumes. It must first be primed with certain isotopes of uranium or plutonium that release more neutrons than are needed to continue the chain reaction at a constant rate. In an ordinary reactor, any surplus neutrons are absorbed in nonfissionable control rods made of a substance, such as boron or cadmium, that readily absorbs neutrons. In a breeder reactor, however, these surplus neutrons are used to transmute certain nonfissionable atoms into fissionable atoms. Thorium (Th-232) can be converted by neutron bombardment into fissionable U-233. Similarly, U-238, the most common isotope of uranium, can be converted by neutron bombardment into fissionable plutonium-239.
Production of Heat and Nuclear Materials
The transmutation of nonfissionable materials to fissionable materials in nuclear reactors has made possible the large-scale production of atomic energy. The excess nuclear fuel produced can be extracted and used in other reactors or in nuclear weapons. The heat energy released by fission in a reactor heats a liquid or gas coolant that circulates in and out of the reactor core, usually becoming radioactive. Outside the core, the coolant circulates through a heat exchanger where the heat is transferred to another medium. This second medium, nonradioactive since it has not circulated in the reactor core, carries the heat away from the reactor. This heat energy can be dissipated or it can be used to drive conventional heat engines that generate usable power. Submarines and surface ships propelled by nuclear reactors and nuclear-powered electric generating stations are in operation. However, nuclear accidents in 1979 at Three Mile IslandThree Mile Island,
site of a nuclear power plant 10 mi (16 km) south of Harrisburg, Pa. On Mar. 28, 1979, failure of the cooling system of the No. 2 nuclear reactor led to overheating and partial melting of its uranium core and production of hydrogen gas, which raised fears of
..... Click the link for more information. and in 1986 at ChernobylChernobyl
, Ukr. Chornobyl, abandoned city, N Ukraine, near the Belarus border, on the Pripyat River. Ten miles (16 km) to the north, in the town of Pripyat, is the Chernobyl nuclear power station, site of the worst nuclear reactor disaster in history. On Apr.
..... Click the link for more information. raised concern over the safety of reactors, and these concerns were revived somewhat in 2011 after an earthquake and tsunami resulted in a nuclear disaster in FukushimaFukushima
, city (1990 pop. 277,528), capital of Fukushima prefecture, N Honshu, Japan, on the Kiso plain. A silk-textile center, it is a major commercial city of NE Japan, also producing cameras, automobiles, fruits, and bonsai trees.
Fukushima prefecture (1990 pop.
..... Click the link for more information. , Japan. Another concern over fission reactors is the storage of hazardous radioactive wasteradioactive waste,
material containing the unusable radioactive byproducts of the scientific, military, and industrial applications of nuclear energy. Since its radioactivity presents a serious health hazard (see radiation sickness), disposing of such material is a great problem.
..... Click the link for more information. . In the United States, the events at Three Mile Island made nuclear fission plants politically unacceptable and economically unattractive for many years; no new plants were approved for construction until 2012. In contrast, in France, Japan, and a few other nations nuclear fission has been used extensively for power generation. The Japanese and French adopted a more cautious approach in the aftermath of Fukushima; Germany, which has been less dependent on nuclear reactors, chose to accelerate its planned phase out of nuclear power generation.
Fusion reactors are being studied as an alternative to fission reactors. The design of nuclear fusion reactors, which are still in the experimental stage, differs considerably from that of fission reactors. In a fusion reactor, the principal problem is the containment of the plasmaplasma,
in physics, fully ionized gas of low density, containing approximately equal numbers of positive and negative ions (see electron and ion). It is electrically conductive and is affected by magnetic fields.
..... Click the link for more information. fuel, which must be at a temperature of millions of degrees in order to initiate the reaction. Magnetic fields have been used in several ways to hold the plasmas in a "magnetic bottle." If development should reach a practical stage of application, it is expected that fusion reactors would have many advantages over fission reactors. Fusion reactors, for instance, would produce less hazardous radioactive waste. Because their fuel, deuterium (an isotope of hydrogen readily separated from water), is far less expensive to obtain than enriched uranium, fusion reactors also would be far more economical to operate.
See G. I. Bell, Nuclear Reactor Theory (1970); R. J. Watts, Elementary Primer of Diffusion Theory and the Chain Reaction (1982).
reactor, nuclear:see nuclear reactornuclear reactor,
device for producing controlled release of nuclear energy. Reactors can be used for research or for power production. A research reactor is designed to produce various beams of radiation for experimental application; the heat produced is a waste product and is
..... Click the link for more information. .
a device in which a controlled nuclear chain reaction takes place accompanied by the release of energy. The first nuclear reactor was built in the United States in December 1942 under the direction of E. Fermi. In Europe the first nuclear reactor was started up in December 1946 in Moscow under the direction of I. V. Kurchatov. By 1978, there were about 1,000 nuclear reactors of different types in operation around the world. Every nuclear reactor has a core, which contains the nuclear fuel and which is usually surrounded by a neutron reflector; a coolant; a system for controlling the chain reaction; radiation shielding; and a remote control system (Figure 1). The principal characteristic of a nuclear reactor is its power. A power of 1 megawatt corresponds to a chain reaction in which 3 × 1016 fissions occur per 1 sec.
The nuclear fuel is located in the core of the reactor, where the nuclear fission chain reaction takes place and energy is released. The state of a nuclear reactor is characterized by the neutron effective multiplication factor keff or the reactivity ρ:
(1) ρ = (keff – 1)/keff
If keff > 1, the chain reaction increases with time, the reactor is in the supercritical state, and ρ > 0. If keff < 1, the reaction
decreases, the reactor is in the subcritical state, and ρ < 0. At keff = 1, ρ = 0, the reactor is in the critical state and operates at steady state, and the number of fissions is constant over time. To initiate a chain reaction upon starting up the nuclear reactor, a source of neutrons (for example, a mixture of Ra and Be or 252Cf) is usually introduced into the core, although this is not obligatory, since the spontaneous fission of uranium nuclei and cosmic rays provide a sufficient number of initial neutrons for the development of a chain reaction when keff > 1.
235U is used as the fissionable material in most nuclear reactors. If the core contains a neutron moderator (graphite, water, or other substance that contains light nuclei) in addition to the nuclear fuel (natural or enriched uranium), most of the fissions are induced by thermal neutrons (a thermal reactor). Natural uranium that is not enriched with 235U can be used in thermal reactors, as was done in the first reactors. If there is no moderator in the core, most of the fissions are induced by fast neutrons, with energies ℰn > 10 keV (fast reactor). Reactors using intermediate neutrons, with energies of 1–1,000 eV, are also possible. (SeeTHERMAL REACTOR and FAST REACTOR.)
Structurally, nuclear reactors are subdivided into heterogeneous reactors, in which the nuclear fuel is discretely distributed through the core in the form of blocks separated by a neutron moderator, and homogeneous reactors, in which the nuclear fuel and moderator are a uniform mixture—-solution or suspension (seeHETEROGENEOUS REACTOR and HOMOGENEOUS REACTOR). The blocks containing the nuclear fuel in a heterogeneous reactor are called fuel elements and are arranged in a regular lattice (seeFUEL ELEMENT). The volume allotted for one fuel element is called a cell. Nuclear reactors are subdivided according to use, into power reactors and research reactors (see). A single reactor can often perform several functions (seeDUAL-PURPOSE REACTOR).
Criticality condition. The condition of criticality of a nuclear reactor takes the form
(1) Keff = k∞ · P = 1
where 1 – P is the probability of escape (leakage) of neutrons from the reactor core, k∞ is the multiplication factor in an infinite reactor core, defined for thermal reactors by the four-factor formula:
(2) K∞ = v∊φθ
where v is the average number of secondary (fast) neutrons that arise upon the fission of a 235U nucleus by thermal neutrons, ∊ is the fast fission factor (the increase in the number of neutrons as a result of the fission of nuclei, mainly 238U nuclei, by fast neutrons), φ is the probability that a neutron, in the course of being moderated, will escape capture by a 238U nucleus, and θ is the thermal utilization factor. The quantity η = v/(1 + α) is often used, where α is the ratio of the radiative capture cross section σr to the fission cross section σf.
Condition (1) determines the size of the nuclear reactor. For instance, in the case of a natural-uranium, graphite-moderated reactor, v = 2.4, ∊ ≈ 1.03, and ∊φθ ≈ 0.44, whence k∞ = 1.08. This means that for k∞ > 1, P must be less than 0.93, which corresponds to reactor core dimensions of about 5–10 m (as reactor theory shows). The volume of a modern power reactor reaches hundreds of cubic meters and is determined mainly by the possibilities of heat removal rather than by conditions of criticality. The volume of the reactor core in the critical state is called the critical volume, and the mass of fissionable material is called the critical mass. Nuclear reactors with fuel in the form of solutions of salts of pure fissionable isotopes in water and with a water neutron reflector have the smallest critical mass, which is 0.8 kg for 235U and 0.5 kg for 239Pu. 251Cf has the lowest critical mass (theoretically 10 g). The critical parameters of a graphite-moderated nuclear reactor that uses natural uranium as fuel are as follows: mass of uranium, 45 tons, and graphite volume, 450 cu m. To reduce neutron leakage, the core is spherical or close to spherical, for example, a cylinder, with a height of the order of the diameter, or a cube (minimum ratio of surface to volume).
|Table 1. Values of v and η for thermal neutrons1|
The quantity v is known for thermal neutrons to an accuracy of 0.3 percent (Table 1). Upon an increase in the energy ℰn of a neutron that has induced fission, v increases according to the law v = vth + 0.15ℰn, where vth corresponds to fission by thermal neutrons and ℰn is measured in megaelectron volts.
Although the quantity (∊ – 1) is usually only a few percent, the part played by fission induced by fast neutrons is still considerable, since for large reactors (k∞ – 1) ≪ 1 (the natural-uranium, graphite-moderated reactors, in which a chain reaction was first carried out, could not have been built if there had been no fission by fast neutrons).
The maximum possible value of θ is reached in a nuclear reactor that contains only fissionable nuclei. Nuclear power reactors use low-enriched uranium (concentration of 235U, about 3–5 percent), and the 238U nuclei absorb a considerable fraction of the neutrons. For example, the maximum value of vθ = 1.32 for a natural mixture of uranium isotopes. Neutron absorption in the moderator and structural materials usually does not exceed 5–20 percent of the absorption by all isotopes of the nuclear fuel. Of all the moderators, heavy water has the lowest neutron absorption, and among structural materials, Al and Zr have the lowest neutron absorption.
The probability of the resonance capture of neutrons by 238U nuclei in the course of moderation (1 – φ) is considerably reduced in heterogeneous reactors. The reduction of (1 – φ) is due to the fact that the number of neutrons with energies close to resonance energies drops sharply inside the fuel block, and only the outer layer of the block participates in resonance absorption. The heterogeneous structure of the nuclear reactor makes it possible to use natural uranium in a chain reaction process. Although the structure reduces the θ, the loss in reactivity is much less than the gain as a result of the reduction in resonance absorption.
Calculation of the parameters of thermal reactors necessitates the determination of the thermal neutron spectrum. If neutron absorption is very weak and a neutron has time to collide many times with moderator nuclei before absorption, then thermodynamic equilibrium is established between the moderating medium and the neutron gas (thermalization of neutrons) and the spectrum of thermal neutrons is described by the Maxwellian distribution. In reality, neutron absorption in the reactor core is fairly high. This leads to a deviation from the Maxwellian distribution—the average energies of neutrons are greater than the average energies of the molecules of the medium. The motions of nuclei, the chemical bonds of atoms, and other factors influence thermalization.
Fuel burnup and production. As the nuclear reactor operates, the fuel composition changes as a result of the accumulation of fission fragments and the formation of transuranium elements, chiefly Pu isotopes. The influence of fission fragments on reactor reactivity is called poisoning (for radioactive fragments) and slagging (for stable fragments). Poisoning is caused chiefly by 135Xe, which has the highest neutron absorption cross section (2.6 × 106 barns). Its half-life is T½ = 9.2 hours, and the fission yield is 6–7 percent. Most of the 135Xe is formed as a result of the decay of 135I (T½ = 6.8 hours). The keff changes by 1–3 percent upon poisoning. The large absorption cross section of l35Xe and the presence of the intermediate isotope l35I give rise to two important effects. First, there is an increase in the concentration of 135Xe and, consequently, a reduction in the reactivity after shutdown or loss of reactor power (“iodine well”). This either necessitates having an additional reserve of reactivity in the control devices or precludes short-term shutdowns or power fluctuations. The depth and duration of the iodine well depend on the neutron flux φ: at φ = 5 × 1013 neutrons/cm2 · sec, the duration of the iodine well is about 30 hours, and the depth is twice the steady-state change in keff caused by poisoning with 135Xe. Second, poisoning may cause space-time fluctuations in the neutron flux φ) and, consequently, in the reactor power as well. These fluctuations arise at φ > 1013 neutrons/cm2 · sec and in the case of large reactors. The periods of fluctuations are about ten hours.
A large number of different stable fragments arise upon the fission of nuclei. A distinction is made between fission fragments with absorption cross sections that are larger than the absorption cross section of the fissioning isotope and fission fragments with absorption cross sections that are lower. The concentration of the former reaches saturation within the first few days of a nuclear reactor’s operation (principally l49Sm, which changes keff by 1 percent). The concentration of the latter and the negative reactivity that they cause increase linearly with time.
The formation of transuranium elements in a nuclear reactor follows the reaction schemes
Here, c denotes neutron capture, and the number under the arrow is the half-life.
The accumulation of 239Pu (nuclear fuel) in the initial operation of the nuclear reactor proceeds linearly with time, the rate being faster (with fixed burnup of 235U) the lower the enrichment of the uranium. The concentration of 239Pu approaches a constant value, which does not depend on the degree of enrichment but is determined by the ratio of the neutron capture cross sections of 238U and 239Pu. The characteristic time of reaching equilibrium concentration of 239Pu is about 3/φ) years (φ in units of 1013neutrons/cm2 · sec). The isotopes 240Pu and 241Pu reach the equilibrium concentration only upon repeated burning of the fuel in the nuclear reactor after the reprocessing of the nuclear fuel.
The burnup of nuclear fuel is characterized by the total energy released in the nuclear reactor per ton of fuel. For nuclear reactors using natural uranium, maximum burnup is about 10 gigawatt days per ton (heavy-water-moderated reactors). A burnup of about 20–30 gigawatt days per ton is attained in reactors using weakly enriched uranium (2–3 percent 235U), and in fast reactors it reaches 100 gigawatt days per ton. A burnup of 1 gigawatt day per ton corresponds to the combustion of 0.1 percent of the nuclear fuel.
The reactivity of the reactor decreases as the fuel burns out (in reactors employing natural uranium, some increase in reactivity occurs in the case of low burnup). Spent fuel can be replaced all at once throughout the core, or it can be replaced gradually, one fuel element at a time, in which case fuel elements of all ages are present in the core (the continuous overloading mode). (Intermediate versions are also possible.) In the first case, the reactor with fresh fuel has excess reactivity that must be compensated. In the second case, such compensation is needed only upon the initial loading and until the continuous overloading mode has been established. Continuous overloading increases burnup, since the reactivity is determined by the average concentrations of fissionable nuclides (the fuel elements with minimum concentration of fissionable nuclides are removed). Table 2 gives the composition of the nuclear fuel removed from a 3-gigawatt water-moderated
|Table 2. Composition of unloaded fuel|
|Isotopes||Amount (kg)||Isotopes||Amount (kg)|
|1Including 1.585 kg of recovere d235U|
water-cooled reactor (seeWATER-MODERATED WATER-COOLED REACTOR). The entire core is removed after three years of operation of the nuclear reactor and 3/φ) years of “aging” (φ = 3 × 1013 neutrons/cm2 · sec). The initial composition is as follows; 238U, 77,350 kg; 235U, 2,630 kg; and 234U, 20 kg. The total mass of the loaded fuel is 3 kg greater than that of the unloaded fuel (the energy released “weighs” 3 kg). After the reactor is shut down, the fuel continues to release energy, at first mainly because of fission by delayed neutrons and then, after 1–2 min, mainly because of the beta and gamma emissions of fission fragments and transuranium elements. If the nuclear reactor has been in operation for a fairly long time prior to shutdown, within two minutes after shutdown the energy release is 3 percent of the preshutdown level, with corresponding values of 1 percent after one hour, 0.4 percent after one day, and 0.05 percent after one year.
The conversion ratio Kc is the ratio of the quantity of fissionable Pu isotopes generated in the reactor to the amount of 235U consumed. In table 2, Kc = 0.25. Kc increases with a reduction in enrichment and burnup. Thus, in a heavy-water-moderated reactor using natural uranium as fuel, Kc = 0.55 for a burnup of 10 gigawatt days per ton, whereas for very low burnups (in which case, Kc is called the initial plutonium ratio), Kc = 0.8. If the nuclear reactor consumes and produces the same isotopes (breeder reactor), the ratio of the rate of production to the rate of consumption is called the breeding ratio Kb. In thermal reactors, Kb < 1, whereas in fast reactors, Kb may reach 1.4–1.5. The increase in Kb for fast reactors can be attributed mostly to the fact that for fast neutrons, v increases and α decreases, especially for 239Pu (seeBREEDER REACTOR).
Reactor control. To regulate the operation of a reactor, it is important that some of the neutrons escape from the fission fragments with delay. The percentage of such delayed neutrons is small—0.68 percent for 235U and 0.22 percent for 239Pu (in Table 1, v is the sum of the number v0 of prompt neutrons and the number vdel of delayed neutrons). The delay time Tdel ranges from 0.2 to 55 sec. If keff – 1 <̃ vdel/v0 or less, the number of fissions in the reactor increases (keff > 1) or decreases (keff < 1) with a characteristic time ~Tdel. Without delayed neutrons, this time would be several orders of magnitude less, which would considerably complicate reactor control.
The reactor control system includes (1) emergency devices, which reduce reactivity, that is, devices that introduce negative reactivity into the reactor upon the appearance of an emergency signal; (2) automatic regulators, which maintain a constant neutron flux φ, thereby keeping the power constant as well; and (3) compensating devices, which compensate for the effects of poisoning, burnup, and temperature. In most cases, these are in the form of rods, which are raised or lowered in the reactor core and which are made of materials with strong neutron-absorbing properties, such as cadmium or boron. The insertion and withdrawal of the rods are controlled by mechanisms actuated by signals from instruments that are sensitive to the magnitude of neutron flux. The elements used for burnup compensation may be either burnable absorbers, whose efficiency decreases as they capture neutrons (cadmium, boron, and rare-earth elements), or solutions of absorbing material in the moderator. A negative temperature coefficient of reactivity (ρ decreases with increasing temperature) is conducive to stable reactor operation. If the temperature coefficient is positive, the operation of the reactor control system is considerably complicated.
The nuclear reactor is equipped with instrumentation that provides information to the operator on the state of the reactor, such as information on the neutron flux in different areas of the core, coolant flow rate and temperature, the level of ionizing radiation in different parts of the reactor and in auxiliary rooms, and the position of control devices in the reactor control system. The information received from the instruments is fed to a computer, which may present the data to the operator in processed form (report mode), recommend necessary changes in reactor operating conditions based on mathematical processing of the information (advisory mode), or actuate reactor control within certain limits without participation of the operator (control mode).
Classification of reactors. Nuclear reactors are subdivided according to purpose and power into test reactors, research reactors, isotopic reactors, and power reactors. Test reactors (critical assembly) are used to study different physical quantities of importance for the design and use of nuclear reactors; the power of such reactors does not exceed a few kilowatts. In research reactors, the fluxes of neutrons and gamma quanta that are generated in the core are used for conducting studies in nuclear physics, solid-state physics, radiation chemistry, and biology, for testing materials to be used in intense neutron fluxes (including reactor components), and for producing isotopes. The power of such reactors does not exceed 100 megawatts, and the energy released is not, as a rule, utilized. The pulse reactor is a kind of research reactor (see PULSE REACTOR). In isotopic reactors, neutron fluxes are used to produce isotopes, including Pu and 3H for military purposes. In power reactors, the energy that is released upon the fission of nuclei is used for the production of electrical energy, district heating, and desalinization of seawater, as well as in propulsion units on ships. The thermal power of modern power reactors reaches 3–5 gigawatts.
Nuclear reactors may also be classified on the basis of the type of nuclear fuel used (natural uranium, low-enriched uranium, pure fissionable isotope), the chemical composition of the fuel (metallic uranium, UO2, UC), the type of coolant (H2O, gas, D2O, organic liquids, liquid metals), and the type of moderator (C, H2O, D2O, Be, BeO, metal hydrides, without moderator). The most widely used facilities are heterogeneous thermal reactors with H2O, C, and D2O as the moderators and with H2O, gas, and D2O as the coolants. Fast reactors will be intensively developed in the next few decades. Such reactors will “burn” 238U, which will improve the utilization of nuclear fuel by a factor of tens over thermal reactors. This considerably increases the resources of the nuclear power industry.
REFERENCESWeinberg, A., and E. Wigner. Fizicheskaia teoriia iadernykh reaktorov. Moscow, 1961. (Translated from English.)
Kramerov, A. Ia., and Ia. V. Shevelev. Inzhenernye raschety iadernykh reaktorov. Moscow, 1964.
Bat’, G. A., A. S. Kochenov, and L. P. Kabanov. Issledovatel’skie iadernye reaktory. Moscow, 1972.
Bell, G., and S. Glasstone. Teoriia iadernykh reaktorov. Moscow, 1974. (Translated from English.)
Goncharov, V. V. “30-letie pervogo sovetskogo iadernogo reaktora.” Atomnaia energiia, 1977, vol. 42, issue 2.
A. D. GALANIN
nuclear reactor[′nü·klē·ər rē′ak·tər]
A system utilizing nuclear fission in a controlled and self-sustaining manner. Neutrons are used to fission the nuclear fuel, and the fission reaction produces not only energy and radiation but also additional neutrons. Thus a neutron chain reaction ensues. A nuclear reactor provides the assembly of materials to sustain and control the neutron chain reaction, to appropriately transport the heat produced from the fission reactions, and to provide the necessary safety features to cope with the radiation and radioactive materials produced by its operation.
Nuclear reactors are used in a variety of ways as sources for energy, for nuclear irradiations, and to produce special materials by transmutation reactions. The generation of electrical energy by a nuclear power plant makes use of heat to produce steam or to heat gases to drive turbogenerators. Direct conversion of the fission energy into useful work is possible, but an efficient process has not yet been realized to accomplish this. Thus, in its operation the nuclear power plant is similar to the conventional coal-fired plant, except that the nuclear reactor is substituted for the conventional boiler as the source of heat.
The rating of a reactor is usually given in kilowatts (kW) or megawatts-thermal [MW(th)], representing the heat generation rate. The net output of electricity of a nuclear plant is about one-third of the thermal output. Significant economic gains have been achieved by building improved nuclear reactors with outputs of about 3300 MW(th) and about 1000 MW-electrical [MW(e)]. See Electric power generation, Nuclear power
Fuel and moderator
The fission neutrons are released at high energies and are called fast neutrons. The average kinetic energy is 2 MeV, with a corresponding neutron speed of 1/15 the speed of light. Neutrons slow down through collisions with nuclei of the surrounding material. This slowing-down process is made more effective by the introduction of materials of low atomic weight, called moderators, such as heavy water (deuterium oxide), ordinary (light) water, graphite, beryllium, beryllium oxide, hydrides, and organic materials (hydrocarbons). Neutrons that have slowed down to an energy state in equilibrium with the surrounding materials are called thermal neutrons, moving at 0.0006% of the speed of light. The probability that a neutron will cause the fuel material to fission is greatly enhanced at thermal energies, and thus most reactors utilize a moderator for the conversion of fast neutrons to thermal neutrons. See Thermal neutrons
With suitable concentrations of the fuel material, neutron chain reactions also can be sustained at higher neutron energy levels. The energy range between fast and thermal is designated as intermediate. Fast reactors do not have moderators and are relatively small.
Only three isotopes—uranium-235, uranium-233, and plutonium-239—are feasible as fission fuels, but a wide selection of materials incorporating these isotopes is available.
The major portion of the energy released by the fissioning of the fuel is in the form of kinetic energy of the fission fragments, which in turn is converted into heat through the slowing down and stopping of the fragments. For the heterogeneous reactors this heating occurs within the fuel elements. Heating also arises through the release and absorption of the radiation from the fission process and from the radioactive materials formed. The heat generated in a reactor is removed by a primary coolant flowing through it.
Coolants are selected for specific applications on the basis of their heat-transfer capability, physical properties, and nuclear properties.
Water has many desirable characteristics. It was employed as the coolant in many of the first production reactors, and most power reactors still utilize water as the coolant. In a boiling-water reactor (BWR; see illustration), the water boils directly in the reactor core to make steam that is piped to the turbine. In a pressurized-water reactor (PWR), the coolant water is kept under increased pressure to prevent boiling. It transfers heat to a separate stream of feed water in a steam generator, changing that water to steam.
For both boiling-water and pressurized-water reactors, the water serves as the moderator as well as the coolant. Both light water and heavy water are excellent neutron moderators, although heavy water (deuterium oxide) has a neutron-absorption cross section approximately 1/500 that for light water that makes it possible to operate reactors using heavy water with natural uranium fuel. The high pressure necessary for water-cooled power reactors determines much of the plant design.
Gases are inherently poor heat-transfer fluids as compared with liquids because of their low density. This situation can be improved by increasing the gas pressure; however, this introduces other problems and costs. Helium is the most attractive gas (it is chemically inert and has good thermodynamic and nuclear properties) and has been selected as the coolant for the development of high-temperature gas-cooled reactor (HTGR) systems, in which the gas transfers heat from the reactor core to a steam generator. The British advanced gas reactor (AGR), however, uses carbon dioxide (CO2). Gases are capable of operation at extremely high temperature, and they are being considered for special process applications and direct-cycle gas-turbine applications.
The alkali metals, in particular, have excellent heat-transfer properties and extremely low vapor pressures at temperatures of interest for power generation. Sodium is attractive because of its relatively low melting point (208°F or 98°C) and high heat-transfer coefficient. It is also abundant, commercially available in acceptable purity, and relatively inexpensive. It is not particularly corrosive, provided low oxygen concentration is maintained. Its nuclear properties are excellent for fast reactors. In the liquid-metal fast breeder reactor (LMFBR), sodium in the primary loop collects the heat generated in the core and transfers it to a secondary sodium loop in the heat exchanger, from which it is carried to the steam generator in which water is boiled to make steam.
The nuclear chain reaction in the reactor core produces energy in the form of heat, as the fission fragments slow down and dissipate their kinetic energy in the fuel. This heat must be removed efficiently and at the same rate it is being generated in order to prevent overheating of the core and to transport the energy outside the core, where it can be converted to a convenient form for further utilization. The energy transferred to the coolant, as it flows past the fuel element, is stored in it in the form of sensible heat and pressure and is called the enthalpy of the fluid. In an electric power plant, the energy stored in the fuel is further converted to kinetic energy through a device called a prime mover which, in the case of nuclear reactors, is predominantly a steam turbine. Another conversion takes place in the electric generator, where kinetic energy is converted into electric power as the final energy form to be distributed to the consumers through the power grid and distribution system. See Generator, Prime mover, Steam turbine
Fluid flow and hydrodynamics
Because heat removal must be accomplished as efficiently as possible, considerable attention must be given to fluid-flow and hydrodynamic characteristics of the system.
The heat capacity and thermal conductivity of the fluid at the temperature of operation have a fundamental effect upon the design of the reactor system. The heat capacity determines the mass flow of the coolant required. The fluid properties (thermal conductivity, viscosity, density, and specific heat) are important in determining the surface area required for the fuel—in particular, the number and arrangement of the fuel elements. These factors combine to establish the pumping characteristics of the system because the pressure drop and coolant temperature rise in the core are directly related. See Conduction (heat), Heat capacity
The temperature of the reactor coolant increases as it circulates through the reactor core. Fluctuations in power level or in coolant flow rate result in variations in the temperature rise. A reactor is capable of very rapid changes in power level, particularly reduction in power level, which is a safety feature of the plant. Reactors are equipped with mechanisms (reactor scram systems) to ensure rapid shutdown of the system in the event of leaks, failure of power conversion systems, or other operational abnormalities. Therefore, reactor coolant systems must be designed to accommodate the temperature transients that may occur because of rapid power changes. In addition, they must be designed to accommodate temperature transients that might occur as a result of a coolant system malfunction, such as pump stoppage.
Coolant system components
The development of reactor systems has led to the development of special components for reactor component systems. Because of the hazard of radioactivity, leak-tight systems and components are a prerequisite to safe, reliable operation, and maintenance. Special problems are introduced by many of the fluids employed as reactor coolants.
More extensive component developments have been required for sodium, which is chemically active and is an extremely poor lubricant. Centrifugal pumps employing unique bearings and seals have been specially designed. Sodium is an excellent electrical conductor and, in some special cases, electromagnetic-type pumps have been used. These pumps are completely sealed, contain no moving parts, and derive their pumping action from electromagnetic forces imposed directly on the fluid. See Centrifugal pump
A typical reactor core for a power reactor consists of the fuel element rods supported by a grid-type structure inside a vessel.
Structural materials employed in reactor systems must possess suitable nuclear and physical properties and must be compatible with the reactor coolant under the conditions of operation. The most common structural materials employed in reactor systems are stainless steel and zirconium alloys. Zirconium alloys have favorable nuclear and physical properties, whereas stainless steel has favorable physical properties. Aluminum is widely used in low-temperature test and research reactors; zirconium and stainless steel are used in high-temperature power reactors. Zirconium is relatively expensive, and its use is therefore confined to applications in the reactor core where neutron absorption is important.
Reactors maintain a separation of fuel and coolant by cladding the fuel. The cladding is designed to prevent the release of radioactivity from the fuel. The cladding material must be compatible with both the fuel and the coolant.
The cladding materials must also have favorable nuclear properties. The neutron-capture cross section is most significant because the unwanted absorption of neutrons by these materials reduces the efficiency of the nuclear fission process. Aluminum is a very desirable material in this respect; however, its physical strength and corrosion resistance in water decrease very rapidly above about 300°F (149°C).
Zirconium has favorable neutron properties, and in addition is corrosion-resistant in high-temperature water. It has found extensive use in water-cooled power reactors. Stainless steel is used for the fuel cladding in fast reactors, in some light-water reactors for which neutron captures are less important.
A reactor is critical when the rate of production of neutrons equals the rate of absorption in the system. The control of reactors requires the continuing measurement and adjustment of the critical condition. The neutrons are produced by the fission process and are consumed in a variety of ways, including absorption to cause fission, nonfission capture in fissionable materials, capture in fertile materials, capture in structure or coolant, and leakage from the reactor to the shielding. A reactor is subcritical (power level decreasing) if the number of neutrons produced is less than the number consumed. The reactor is supercritical (power level increasing) if the number of neutrons produced exceeds the number consumed. See Reactor physics
Reactors are controlled by adjusting the balance between neutron production and neutron consumption. Normally, neutron consumption is controlled by varying the absorption or leakage of neutrons; however, the neutron generation rate also can be controlled by varying the amount of fissionable material in the system.
The reactor control system requires the movement of neutron-absorbing rods (control rods) in the reactor under carefully controlled conditions. They must be arranged to increase reactivity (increase neutron population) slowly and under good control. They must be capable of reducing reactivity, both rapidly and slowly.
The control drives can be operated by the reactor operator or by automatic control systems. Reactor scram (rapid reactor shutdown) can be initiated automatically by a wide variety of system scram-safety signals, or it can be started by the operator depressing a scram button in the control room.
Control drives are electromechanical or hydraulic devices that impart in-and-out motion to the control rods. They are usually equipped with a relatively slow-speed reversible drive system for normal operational control. Scram is usually effected by a high-speed overriding drive accompanied by disconnecting the main drive system.
Reactor applications include mobile, stationary, and packaged power plants; production of fissionable fuels (plutonium and uranium-233) for military and commercial applications; research, testing, teaching-demonstration, and experimental facilities; space and process heat; dual-purpose design; and special applications. The potential use of reactor radiation or radioisotopes produced for sterilization of food and other products, steam for chemical processes, and gas for high-temperature applications has been recognized. See Nuclear fuel cycle, Nuclear fuels reprocessing