nuclear reactor


nuclear reactor
reactor (def. 4). Also called nuclear pile.
[1940-45]

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Device that can initiate and control a self-sustaining series of nuclear-fission reactions.

Neutrons released in one fission reaction may strike other heavy nuclei, causing them to fission. The rate of this chain reaction is controlled by introducing materials, usually in the form of rods, that readily absorb neutrons. Typically, control rods made of cadmium or boron are gradually inserted into the core if the series of fissions begins to proceed at too great a rate, which could lead to meltdown of the core. The heat released by fission is removed from the reactor core by a coolant circulated through the core. Some of the thermal energy in the coolant is used to heat water and convert it to high-pressure steam. This steam drives a turbine, and the turbine's mechanical energy is then converted into electricity by means of a generator. Besides providing a valuable source of electric power for commercial use, nuclear reactors also serve to propel certain types of military surface vessels, submarines, and some unmanned spacecraft. Another major application of reactors is the production of radioactive isotopes that are used extensively in scientific research, medical therapy, and industry.

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device
Introduction
 any of a class of devices that can initiate and control a self-sustaining series of nuclear fissions. Such devices are used as research tools, as systems for producing radioisotopes, and most prominently as energy sources. The latter are commonly called power reactors.

      Fission is the process in which a heavy nucleus splits into two smaller fragments. A large amount of energy is released in this process, and this energy is the basis of fission power systems. The nuclear fragments are in very excited states and emit neutrons (neutron) and other forms of radiation. The neutrons can then cause new fissions, which in turn yield more neutrons, and so forth. Such a continuous self-sustaining series of fissions constitutes a fission chain reaction. For a detailed discussion of nuclear fission, see nuclear fission. (nuclear fission)

      In an atomic bomb the chain reaction is designed to increase in intensity until much of the material has fissioned. This increase is very rapid and produces the extremely sharp, tremendously energetic explosions characteristic of such bombs. In a nuclear reactor the chain reaction is maintained at a controlled, nearly constant level. Nuclear reactors are so designed that they cannot explode like atomic bombs.

      Most of the energy of fission—about 85 percent of it—is released within a very short time after the process occurs. The rest of the energy comes from the radioactive decay (radioactivity) of fission products, which is what the fragments are called after they have emitted neutrons. Radioactive decay continues when the fission chain has been stopped, and its energy must be dealt with in any proper reactor design.

Principles of operation

Chain reaction and criticality
      The course of a chain reaction is determined by the probability that a neutron released in fission will cause a subsequent fission. If on the average less than one neutron causes another fission, the rate of fission will decrease with time and ultimately drop to zero. This situation is called subcritical. When an average of one neutron from a fission causes another fission, the fission rate is steady and the reactor is critical. A critical reactor is what is usually desired. When more than one neutron causes a subsequent fission, fission rate and power increase and the situation is termed supercritical. In order to be able to increase power, reactors are designed to be slightly supercritical when all controls are removed.

Reactor control
      A parameter called reactivity is positive when a reactor is supercritical, zero at criticality, and negative when the reactor is subcritical. Reactivity can be controlled in various ways: by adding or removing fuel; by changing the fraction of neutrons that leaks from the system; or by changing the amount of an absorber that competes with the fuel for neutrons. Control is generally accomplished by varying absorbers, which are commonly in the form of movable elements—control rods—or sometimes by changing the concentration of the absorber in a reactor coolant. Leakage changes are usually automatic; for example, an increase of power may cause coolant to boil (see below), which in turn increases neutron leakage and reduces reactivity. This, and other types of negative power-reactivity feedbacks, are vital aspects of safe reactor design.

      Reactor control is facilitated by the presence of delayed neutrons. These neutrons are emitted by fission products some time after fission has occurred. The fraction of delayed neutrons is small, but there is a sufficient number of such neutrons for the types of changes needed to regulate an operating reactor, and so the chain reaction must “wait” for them before it can respond. This eases operation considerably.

Fissile (fissile material) and fertile materials
      All heavy nuclides (nuclide) can fission if they are in an excited enough state, but only a few fission readily when struck by slow (low-energy) neutrons. Such species of atoms are called fissile. The most important of these are uranium-233 (233U), uranium-235 (235U), plutonium-239 (239Pu), and plutonium-241 (241Pu). The only one that occurs in usable amounts in nature is uranium-235, which makes up a mere 0.711 percent of natural uranium by weight. Uranium-233 can be produced by neutron capture in natural thorium (232Th); that is to say, when a nucleus of thorium-232 absorbs a neutron, it becomes uranium-233. Similarly, plutonium-239 is created by neutron capture in uranium-238 (238U; the principal constituent of naturally occurring uranium), and plutonium-241 is formed when a neutron is absorbed into plutonium-240 (240Pu). Plutonium-240 builds up over time in most power reactors. Thorium-232, uranium-238, and plutonium-240 are termed fertile materials because they can be transformed into fissile materials.

      A power reactor contains both fissile and fertile materials. The fertile materials replace fissile materials that are destroyed by fission. This permits the reactor to run longer before the amount of fissile material decreases to the point where criticality can no longer be maintained.

Heat removal (cooling system)
      The energy of fission is quickly converted to heat, the bulk of which is deposited in the fuel. A coolant is therefore required to remove this heat. The most common coolant is water, but any fluid can be used. Heavy water (deuterium oxide), air, carbon dioxide, helium, liquid sodium, sodium-potassium alloy (called NaK), molten salts, and hydrocarbons have all been used in reactors or reactor experiments. Some research reactors are operated at very low power and have no need for a dedicated cooling system; in such units the small amount of heat that is generated is removed by conduction and convection to the environment. Very high power reactors must have extremely sophisticated cooling systems to remove heat quickly and reliably; otherwise, the heat will build up in the reactor fuel and melt it.

Shielding
      An operating reactor is a powerful source of radiation, since fission and subsequent radioactive decay produce neutrons and gamma rays, both of which are highly penetrating radiations. A reactor must have special shielding around it to absorb this radiation in order to protect technicians and other reactor personnel. In a popular class of research reactors known as “swimming pools,” this shielding is provided by placing the reactor in a large, deep pool of water. In other kinds of reactors, the shield consists of a thick concrete structure around the reactor system. The shield also may contain heavy metals, such as lead or steel, for more effective absorption of gamma rays, and heavy aggregates may be used in the concrete itself for the same purpose.

Critical concentration and size
      Not every arrangement of material containing fissile fuel can be brought to criticality. Even if there were no leakage of neutrons from a reactor, a critical concentration of fissile material must be present. Otherwise, absorption of neutrons by other constituents of the reactor will be too high to permit a critical chain reaction to proceed. Similarly, even if there is a high enough concentration for criticality, the reactor must be large enough so that not too many neutrons leak out before being absorbed. This imposes a critical size limit on a reactor of a given concentration.

      Although the only useful fissile material in nature, uranium-235 (uranium), is found in natural uranium, there are just a few combinations and arrangements of this and other materials that can be brought to criticality. To increase the range of feasible reactor designs, enriched uranium can be used. Most of today's power reactors employ enriched uranium fuel in which the percentage of uranium-235 has been increased to 3 to 4 percent. This is about five times the concentration in natural uranium. Large plants for enriching uranium exist in several countries; enrichment has now become a commercial enterprise (see below).

Thermal, intermediate, and fast reactors
      Reactors are conveniently classified according to the typical energies of the neutrons that cause fission. Neutrons emanating in fission are very energetic; their average energy is around two million electron volts (MeV), 80 million times higher than the energy of atoms in ordinary matter at room temperature. As the neutrons collide with nuclei in a reactor, they lose energy. The choice of reactor materials and of fissile material concentrations determines how much they are slowed down by these collisions before causing fission.

      In a thermal reactor, enough collisions are permitted to occur so that most of the neutrons reach thermal equilibrium with the atoms of the reactor at energies of a few hundredths of an electron volt. Neutrons lose energy most efficiently by colliding with light atoms such as hydrogen (mass 1), deuterium (mass 2), beryllium (mass 9), and carbon (mass 12). Materials that contain atoms of this kind—water, heavy water, beryllium metal and oxide, and graphite—are deliberately incorporated into the reactor for this reason and are known as moderators. Since water and heavy water also can function as coolants, they can do double duty in thermal reactors.

      One disadvantage of thermal reactors is that at low energies uranium-235 and plutonium-239 not only can be fissioned by thermal (or slow) neutrons but also can capture neutrons without undergoing fission. This destroys fissile atoms without any fission to show for it. When neutrons of higher energy cause fission, fewer of these captures occur. To achieve this, a reactor can be built to operate without a moderator. Then, depending on how many collisions take place with heavier atoms before fission occurs, the typical fission-causing neutrons can have energies in the range of 0.5 electron volt to thousands of electron volts (intermediate reactors) or several hundred thousand electron volts (fast reactors). Such reactors require higher concentrations of fissile material to reach criticality than do thermal reactors but are more efficient at converting fertile material to fissile material. Indeed, they can be designed to produce more than one new fissile atom for each fissile atom destroyed. Such reactors are called breeders. Breeder reactors (breeder reactor) may become particularly important if the world demand for nuclear power turns out to be a long-term one, since their fuel is manufactured from very abundant fertile materials.

Reactor design and components
      There are a large number of ways in which a reactor may be designed and constructed, and many types have been experimentally realized. Over the years, nuclear engineers have developed reactors with solid fuels and liquid fuels, thick reflectors and no reflectors, forced cooling circuits and natural conduction or convection heat-removal systems, and so on. Most reactors, however, have certain basic components. These are described below.

Core
      All reactors have a core, a central region that contains the fuel, fuel cladding, coolant, and, where separate from the latter, moderator. It is in the core that fission occurs and the resulting neutrons migrate.

      The fuel is usually heterogeneous—i.e., it consists of elements containing fissile material along with a diluent. This diluting agent may be fertile material or simply material that has good mechanical and chemical properties and that does not readily absorb neutrons. The diluted fissile material is enclosed in a cladding—a substance that isolates the fuel from the coolant and keeps the radioactive fission products contained.

Fuel types
      Different kinds of reactors use different types of fuel elements. For example, the light-water reactor (LWR), which is the most widely used variety for commercial power generation in the United States, employs a fuel consisting of pellets of sintered uranium dioxide loaded into cladding tubes of zirconium alloy that measure about one centimetre in diameter and roughly three to four metres long. These tubes, called pins, are bundled together into a fuel assembly, with the pins arranged in a square lattice. The uranium used in the fuel is 3- to 4-percent enriched. Since light (ordinary) water tends to absorb more neutrons than do other moderators, such enrichment is crucial. The CANDU (Canadian deuterium-uranium) reactor, which is the principal type of heavy-water reactor, uses natural uranium compacted into pellets. These pellets are inserted in tubes arranged in a lattice. Such a fuel assembly measures about one metre in length, and several assemblies are arranged end-to-end within a channel inside the reactor core.

      In a high-temperature graphite reactor the fuel is made of small spherical particles containing uranium dioxide at the centre with concentric shells of carbon, silicon carbide, and carbon around them. (These shells serve as microscopic cladding.) The particles are mixed with graphite and encased in a macroscopic graphite cladding. In a sodium-cooled fast reactor, commonly called a liquid-metal reactor (LMR), the fuel consists of dioxide pellets (French design) or uranium-plutonium-zirconium metal alloy pins (U.S. design) in steel cladding.

      The most common type of fuel used in research reactors consists of plates of a uranium- aluminum alloy with an aluminum cladding. The uranium is enriched to 20 percent, and silicon, along with aluminum, are included in the “meat” of the plate. A common variety of research reactor, known as TRIGA (from training, research, and isotope-production reactors–General Atomic), employs a fuel of mixed uranium and zirconium hydride in zirconium cladding.

Coolants and moderators
      A variety of substances, including light water, heavy water, air, carbon dioxide, helium, liquid sodium, liquid sodium-potassium alloy, and hydrocarbons (oils), have been used as coolants. Such substances are good conductors of heat and serve to carry the thermal energy produced by fission from the core to the steam-generating equipment of the nuclear power plant.

      In many cases, the same substance functions as both coolant and moderator, as in the case of light and heavy water. The moderator slows down the fast (high-energy) neutrons emitted in fission to speeds at which they are more likely to induce fission. In doing so, the moderator helps initiate and sustain a fission chain reaction.

Reflector
      A reflector is a region of unfueled material surrounding the core. Its function is to scatter neutrons that leak from the core and thereby return some of them to the core. This reduces core size and smooths out the power density. The reflector is particularly important in research reactors, since it is the region in which much of the experimental apparatus is located. Some reflectors are located inside the core as central islands in which high neutron intensities can be achieved for experimental purposes. In most types of power reactors, a reflector is less important, because the reactors are large and do not leak many neutrons. Yet, as it serves to keep the power density uniform, such an unfueled zone of moderator material is left around the core. The liquid-metal reactor represents a special case. Most sodium-cooled reactors are deliberately built to allow a large fraction of their neutrons—those not needed to maintain the chain reaction—to leak from the core. These neutrons are valuable because they can produce new fissile material if they are absorbed by fertile material. Thus, fertile material—generally depleted uranium or its dioxide—is placed around the core to catch the leaking neutrons. Such an absorbing reflector is referred to as a blanket or a breeding blanket.

Reactor control elements
      All reactors need special elements for control. Although control can be achieved by varying parameters of the coolant circuit or by varying the amount of absorber dissolved in the coolant or moderator, by far the most common method involves the use of special absorbing assemblies—namely, control rods or sometimes blades. Typically a reactor is equipped with three types of rods for different purposes: (1) safety rods for starting up and shutting down the reactor, (2) regulating rods for adjusting the reactor's power rate, and (3) shim rods for compensating for changes in reactivity as fuel is depleted by fission and capture.

      The most important function of the safety rods is to shut down the reactor, either when such a shutdown is scheduled or in case of a real or suspected emergency. These rods contain enough absorber to terminate a chain reaction under any conceivable condition. They are withdrawn before fuel is loaded and remain available in case a loading error requires their action. After the fuel is loaded, the rods are inserted, to be withdrawn again when the reactor is ready for operation. The mechanism by which they are moved is designed to be fail-safe in the sense that if there is a mechanical failure the safety rods will fall by gravity into the reactor. In some cases, moreover, the safety rods have an automatic feature, such as a fuse, which releases them by virtue of physical effects independent of electronic signals.

      Regulating rods are deliberately designed to affect reactivity only by a small degree. It is assumed that at some time the rods might be totally withdrawn by mistake, and the idea is to keep the added reactivity in such cases well within sensible limits. A well-designed regulating rod will add so little reactivity when it is removed that the delayed neutrons will continue to control the rate of power increase.

      Shim rods are designed to compensate for the effects of burnup (i.e., energy production). Reactivity changes resulting from burnup can be large, but they occur slowly over periods of days to years, as compared to the seconds-to-minutes range over which safety actions and routine regulation take place. Therefore, shim rods may control a significant amount of reactivity, but they will work perfectly well under constraints on their speed of movement. A common way in which shims are operated is by inserting or removing them as regulating rods reach the end of their most useful position range. When this happens, shim rods are moved so that the regulating rods can be reset.

      The functions of shim and safety rods are sometimes combined in rods that have low rates of withdrawal but that can be rapidly inserted. This is usually done when the effect of burnup is to decrease reactivity. The rods are only partially inserted at the outset of operation, but the reactor can be quickly shut down by lowering them all the way into the core (scramming). As operation proceeds, the rods are moved farther out so that there is a greater shutdown reactivity margin.

      The amount of shim control required can be reduced by the use of a burnable “poison.” This is a neutron-absorbing material, such as boron or gadolinium, which will burn off faster than the fissile material does. At the beginning of operation, this controls the extra reactivity that has been built into the fuel to compensate for the amount of fuel consumed. At the end of an operating period, the absorber material will have been almost completely destroyed by neutron capture.

Structural components
      These are the parts of a reactor system that hold the reactor together and permit it to function as a useful energy source. The most important structural component is usually the reactor vessel. In both the light-water reactor and the high-temperature gas-controlled reactor (HTGR), a pressure vessel is used so that the coolant can be contained and operated under conditions appropriate for power generation—namely, high temperature and pressure. Within the reactor vessel are structural grids for holding the reactor core and solid reflectors; coolant channels; control-rod guide channels; internal thermohydraulic components (e.g., pumps or steam circulators) in some cases; instrument tubes; and parts of safety systems.

Coolant system (cooling system)
      The function of a power reactor installation is to extract the heat of nuclear fission and convert it to useful power, generally electricity. The coolant system plays a pivotal role in performing this function. A coolant fluid enters the core at low temperature and leaves it at higher temperature. This higher temperature fluid is then directed to conventional thermodynamic components where the heat is converted into electrical power. In most light-water, heavy-water, and gas-cooled power reactors, the coolant is maintained at high pressure. Sodium and organic coolants operate at atmospheric pressure.

      Research reactors have very simple heat removal systems in which coolant is run through the reactor and the heat that is removed is transferred to ambient air or to water without going through a power cycle. In research reactors of the lowest power running at only a few kilowatts, this may involve simple heat exchange to tap water or to a pool of water cooled with ambient air. During operation at higher power levels, the heat is usually removed by means of a small natural-draft cooling tower.

Containment system
      Reactors are designed with the expectation that they will operate safely without releasing radioactivity to their surroundings. It is, however, recognized that accidents can occur. An approach using multiple barriers has been adopted to deal with such accidents. These barriers are, successively, the fuel cladding, primary vessel, and thick shielding. As a final barrier, the reactor is housed in a containment structure. This consists basically of the reactor building, which is designed and tested to prevent any radioactivity that escapes from the reactor from being released to the environment. As a consequence, the containment structure must be at least nominally airtight. In practice, it must be able to maintain its integrity under circumstances of a drastic nature, such as accidents in which most of the contents of the reactor core are released to the building. It has to withstand pressure buildups and damage from debris propelled by an explosion within the reactor, and it must pass a test to demonstrate that it will not leak more than a small fraction of its contents over a period of several days, even when its internal pressure is well above that of the surrounding air. The most common form of containment building is a cylindrical structure with a spherical dome, which is characteristic of LWR systems. This is much more typical of nuclear plants than the large cooling tower that is often used as a symbol for nuclear power. (It should be noted that cooling towers are found at large modern coal- and oil-fired power stations as well.)

      Reactors other than those of the LWR type also have containment structures, but they vary in shape and construction. When it can be justified that major pressure buildups are not to be expected, the containment can be any form of airtight structure. In the United States, containment structures are required for all commercial power reactors and all high-power research reactors. In general, low-power research reactors are exempt, based on the common assumption that an accident in such systems will not lead to a widespread release of radioactivity. Reactors operated by the U.S. Department of Energy and by the armed services also are exempt, a matter which has caused considerable controversy. Some of these have containment structures, while others do not.

      The concept of containment originated in the United States during the 1950s and has been generally accepted throughout much of the world. The Soviet bloc countries, however, did not concur with this view, and when containment was provided it was generally not up to Western standards. For example, Chernobyl (Chernobyl accident) Unit 4, which suffered a catastrophic explosive accident and fire in 1986, merely had an internal structure that could only withstand the loss of function of a single pressure tube. Though called containment, this was a misnomer by Western standards.

      The most severe test of a containment system occurred during an accident in the United States in 1979 at Three Mile Island Unit 2, near Harrisburg, Pa. In this installation, a stoppage of core cooling resulted in the destruction, including partial melting, of the entire core and the release of a large part of its radioactivity to the enclosure around the reactor. In spite of a hydrogen deflagration that also occurred during the accident, the containment structure prevented all but a very small amount of radioactivity from entering the environment and must be credited with having prevented a major radioactive release and its consequences.

Types of reactors
      Most of the world's existing reactors are power reactors. There also are many research reactors, and the navies of many nations include submarines and surface ships driven by propulsion reactors. There are several types of power reactors, but only one, the light-water reactor, is widely used. Accordingly, this variety is discussed in considerable detail here. Other significant types are briefly described, as are research and propulsion reactors. Some attention is also given to the prospective uses of reactors for space travel and for certain industrial purposes.

Power reactors
Light-water reactor
 As noted above, LWRs are power reactors that are cooled and moderated with ordinary water. There are two basic types: the pressurized-water reactor (PWR) and the boiling-water reactor (BWR). In the first type, high-pressure, high-temperature water removes heat from the core and is then passed to a steam generator (boiler). Here the heat of the coolant is transferred to a stream of water in the generator (the secondary loop in the Figure, B—>), causing the water to boil and slightly superheat. The steam generated by this serves as the working fluid in a steam-turbine cycle.

      In a boiling-water reactor, water passing through the core is allowed to boil at intermediate pressure, and the steam from the reactor is used directly in the power cycle. Although the BWR seems simpler, the PWR has advantages with regard to fuel utilization and power density, and the two concepts have been economically competitive with each other since the 1960s. Both these light-water reactors are fueled with uranium dioxide pellets in zirconium alloy cladding (see above). The BWR fuel is slightly less enriched, but the PWR fuel produces more energy before being discharged, and so these two aspects balance each other out economically. Because the BWR operates at lower pressure, it has a thinner pressure vessel than the PWR; however, because its power density is somewhat lower, the BWR's vessel has a larger diameter for the same reactor power. The internal system of a BWR is more complex, since there are internal recirculation pumps and complex steam separation and drying equipment within its vessel. Though the internals of the PWR are simpler, a BWR power plant is smaller because it has no steam generators. In fact, the steam generators—there are usually four of them in a big PWR plant—are larger than the reactor vessel itself. The control rods of a typical PWR are inserted from the top (through the reactor head), while those of a BWR are inserted from the bottom.

      Light-water reactors are refueled by removing the reactor head—after lowering and unlatching the safety rods in the case of a PWR. This exposes the reactor to visual observation. The pressure vessel is filled to the top with water, and, since the core is near the bottom of the vessel, the water acts as a shield for this operation. Then, the fuel assemblies to be removed are lifted up into a shielded cask within which they are transferred to a storage pool for cooling while they are still highly radioactive. Many of the remaining assemblies are then shifted within the core, and finally fresh fuel is loaded into the empty fuel positions. The purpose of shifting fuel at the time of reload is to achieve an optimal reactivity and power distribution for the next cycle of operation. Reloading is a time-consuming operation. In principle, it could be accomplished in three weeks, but in practice the plant undergoes maintenance during reload, which can take considerably more time—up to a few months. Utilities schedule maintenance and reload during the spring and fall when electricity demand is lowest and the system usually has reserve capacity.

      The discharged fuel stored in the storage pool is not only highly radioactive but also continues to produce energy. This energy is removed by natural circulation of the water in the storage pool. Originally it was expected that this spent fuel could be shipped out for reprocessing within two years, but this option is currently practiced only in France. In the United States, storage pools have continued to receive spent fuel, and some of the pools are filling up. Options available to nuclear plant operators are to store the spent fuel more densely than originally planned, to build new pools, or to store the oldest, no longer very hot fuel in above-ground silos (dry storage). Ultimately this fuel will be transferred to the U.S. Department of Energy for reprocessing or waste disposal or both, but this may not happen until the year 2003 or perhaps later if a viable disposal program is not established.

      During the 1970s light-water reactors represented the cheapest source of new electricity (electric power) in most parts of the world, and it still is economical in Japan, Korea, Taiwan, and France and many other European countries. In the United States, however, strict regulation of light-water reactors during the 1980s, coupled with a decrease in reactor research and development activity, have made the competitive nature of new light-water reactor installations problematic. Plants that have been exceptionally well managed during construction and operation remain competitive; unfortunately, these are not the rule. New designs, developed abroad, may alter this situation, however.

      Most recent light-water reactors have had electric capacity ratings of 1,000 megawatts or more. These are not very suitable for the utility industry, which has had only a slow growth in base-load demand since about 1975. Therefore, as of 1989, advanced light-water reactors in the 600-megawatt capacity range were also being considered.

High-temperature gas-cooled reactor
      The HTGR, as mentioned above, is fueled with a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: (1) a German system that uses spherical fuel elements of tennis-ball size loaded into a graphite silo and (2) an American version in which the fuel is loaded into precisely located graphite hexagonal prisms. In both variants, the coolant consists of helium pressurized to about 100 bars. In the German system the helium passes through interstices in the bed of the spherical fuel elements, while in the American system it passes through holes in the graphite prisms. Both are capable of operating at very high temperature, since graphite has an extremely high sublimation temperature and helium is completely inert chemically. The hot helium can be used directly as the working fluid in a high-temperature gas turbine, or its heat can be utilized to generate steam for a water cycle. Experimental prototypes of both the American and German designs have been built, but no commercial plants were on order as of the early 1990s.

Liquid-metal reactors
      Sodium-cooled, fast-neutron-spectrum reactors received much attention during the 1960s and '70s when it appeared that their breeding capabilities would soon be needed to supply fissile material to a rapidly expanding nuclear industry. When it became clear in the 1980s that this was not a realistic expectation, enthusiasm slackened. The developmental work of the previous decades, however, resulted in the construction of a number of liquid-metal reactors around the world—in the United States, the former Soviet Union, France, Britain, Japan, and Germany. Most liquid-metal reactors are fueled with uranium dioxide or mixed uranium–plutonium dioxides. In the United States, however, the greatest success has been with metal fuels. While some liquid-metal reactors are of the loop type, equipped with heat exchangers (heat exchanger) and pumps outside the primary reactor vessel, others are of the pool variety, featuring a large volume of primary sodium in a pool that also contains the primary pumps and primary-to-secondary heat exchanger. In all types, the heat extracted from the core by primary sodium is transferred to a secondary, nonradioactive sodium loop, which serves as the heat source for a steam generator and turbine. The pool type seems to have some advantage in terms of safety in that the large volume of primary sodium heats up only slowly even if no power is extracted; thus, the reactor is effectively isolated from upsets in the balance of the plant. The reactor core in all such systems is a tightly packed bundle of fuel in steel cladding through which the sodium coolant flows to extract the heat. Most liquid-metal reactors are breeders or are capable of breeding, which is to say that they all produce more fissile material than they consume.

CANDU reactor
      Canada focused its developmental efforts on reactors that would utilize abundant domestic natural uranium as fuel without having to resort to enrichment services that could be supplied only by other countries. The result of this policy was CANDU—the line of natural uranium-fueled reactors moderated and cooled by heavy water. A reactor of this kind consists of a tank, or calandria vessel, containing cold heavy water at normal pressure. The calandria is pierced by pressure tubes made of zirconium alloy, in which the natural uranium fuel is placed and the heavy water coolant is circulated. Power is obtained by transferring the heat from the exiting hot pressurized heavy water to a steam generator and then running the steam from the latter through a conventional turbine cycle. The fuel assembly of a CANDU reactor, which consists of a bundle of short zirconium alloy-clad tubes containing natural uranium dioxide pellets, can be changed while the system is running. A new assembly is simply pushed into one end of a pressure tube and the old one collected as it drops out at the other end. This feature has given the CANDU higher capacity factors than other reactor types. Several countries have purchased CANDU reactors for the same reason that they were developed by Canada—to be independent of imported enrichment services.

Advanced gas-cooled reactor
      The advanced gas-cooled reactor (AGR) was developed in Britain as the successor to reactors of the Calder Hall class, which combined plutonium production and power generation. Calder Hall was the first nuclear station to feed an appreciable amount of power into a civilian network. It was fueled with slugs of natural uranium metal canned in aluminum, cooled with carbon dioxide, and employed a moderator consisting of a block of graphite pierced by fuel channels. In the advanced gas-cooled reactor, fuel pins clad in Zircaloy (trademark for alloys of zirconium having low percentages of chromium, nickel, iron, and tin) and loaded with 2-percent enriched uranium dioxide are placed into zirconium-alloy channels that pierce a graphite moderator block. The enriched fuel permits operation to economic levels of fuel burnup. A coolant of carbon dioxide transports heat to a steam generator, activating a steam-turbine cycle. Although a number of advanced gas-cooled reactors have been built in Britain, they have been less trouble-free and more costly than expected, and no new ones are planned.

Other power reactor types
      A large variety of reactor types have been built and operated on an experimental basis. A few examples include organic liquid-cooled and -moderated reactors that can operate like a pressurized-water reactor without requiring high pressures in the primary circuit; sodium-cooled, graphite-moderated reactors; and heavy-water reactors built in a pressure-vessel design.

Research reactors
Water-cooled, plate-fuel reactor
      This is the most common type of research reactor. It uses enriched uranium fuel in plate assemblies (see above) and is cooled with water. Water-cooled, plate-fuel reactors operate over a wide range of thermal power levels, from a few kilowatts to hundreds of megawatts. The systems with the lowest power ratings are usually operated at universities and used primarily for teaching, while those with the highest are used by major research laboratories chiefly for materials testing and research.

      A common form of the water-cooled, plate-fuel reactor is the pool reactor, in which the reactor core is positioned at the bottom of a large, deep pool of water. This has the advantage of simplifying both observation and the placement of channels from which beams of neutrons can be extracted. At lower thermal power levels, no pumping is required and the cooling water circulates by natural convection. A heat exchanger is usually located at the top of the pool, where the hottest water is stratified. At higher operating power levels, pumping becomes necessary to augment the natural circulation.

      Most pool reactors use the water of the pool as a reflector (see above), but some have blocks of a solid moderator (canned graphite or beryllium metal) around the core that serves as an inner reflector. Graphite and beryllium create a large peak in slow neutron intensity a short distance from the core, which is an advantage when beams of slow neutrons are to be extracted or when such neutrons are used for irradiating materials.

      At higher power levels, it becomes more convenient to employ a tank-type reactor because it is simpler to control the flow path of pumped water in such a system. Low-power teaching reactors also are available in the tank form. The core and reflector arrangement in tank-type, plate-fuel research reactors is the same as in the pool-type systems and has the same variations; however, solid concrete shielding is employed around the sides instead of the water shield characteristic of the latter.

TRIGA reactors
      The TRIGA system is an increasingly popular variety of research reactor. It is another tank-type, water-cooled system, but its fuel differs from that employed by the above-mentioned research reactors. The fuel assembly of the TRIGA consists of zirconium-clad (zirconium) rods of mixed uranium and zirconium hydrides. The virtue of this fuel is that it exhibits an extremely large negative power-reactivity coefficient—so large that the reactor can be made strongly supercritical for an instant, causing its power to rise very rapidly, after which it quickly shuts itself down. The resulting power pulse is useful for a number of dynamic experiments. The total energy released in a pulse is not a problem, since the automatic shutdown occurs very quickly and the energy release is proportional to both peak power and pulse duration.

Other research reactors
      As in the case of power reactors, a number of different reactor types have seen service as research reactors, and some are still in operation. The variety is so great as to defy cataloging. There have been homogeneous (fueled solution cores), fast, graphite-moderated, heavy-water-moderated, and beryllium-moderated reactors, as well as those adapted to use fuels left over from power reactor experiments. The design of research reactors is much more fluid and sensitive to a greater variety of special research demands than is design for other applications.

ship propulsion reactors
      The original, and still the major, naval application of nuclear energy is the propulsion of submarines (submarine). The chief advantage of using nuclear reactors for submarine propulsion is that they, unlike fossil-fuel combustion systems, require no air for power generation. Consequently, a nuclear-powered submarine can remain underwater indefinitely, whereas a conventional diesel-powered submarine must surface periodically to run its engines in air. Nuclear power confers a strategic advantage on naval surface vessels (naval ship) as well because it eliminates their dependence on refueling from vulnerable tankers.

      The design of U.S. naval (United States Navy, The) nuclear power plants is classified for defense security purposes, and so only general information pertaining to them has been published. It is known that such power plants are fueled with highly enriched uranium and moderated and cooled with light water. The design of the first nuclear submarine power plant, that of the USS Nautilus, was heavily influenced by high-power research reactor design. Special features include the incorporation of a very large reactivity margin to accommodate long burnups without refueling and to permit restart after shutdown. For submarine use, the power plant also must be extremely quiet to avoid sonic detection. Various models have been developed to fit the specific requirements of different classes of submarines.

      The nuclear power plants for U.S. aircraft carriers (aircraft carrier) are believed to have been derived from the power plant designs for the largest submarines, but again the particulars of their design have not been published.

      Besides the United States, Britain, France, Russia, and several other countries have nuclear submarines. In each case, the design was developed in secret, but it is generally believed that they are all rather similar; the demands of the application usually lead to similar solutions. Russia also has a small fleet of nuclear-powered icebreakers, whose power plants are thought to be essentially the same as those in their earliest submarines. As with naval vessels, the ability to operate without refueling is an enormous advantage for Arctic icebreakers.

      Prototypes of nuclear-powered commercial cargo ships were built and operated by the United States and West Germany but have now been decommissioned. These vessels did not operate very economically, and opposition to their docking in a number of major ports also was a factor in their decommissioning. The prototypes were powered by reactors of the pressurized-water type.

Production reactors
      The very first nuclear reactors were built for the express purpose of manufacturing plutonium for nuclear weapons (nuclear weapon), and the euphemism of calling them production reactors has persisted to this day. At present, most of the material produced by such systems is tritium (3H, or T), the fuel for hydrogen bombs. Plutonium has a long half-life, and so countries with arsenals of nuclear weapons using plutonium as fissile material generally have more than they expect to need. On the other hand, tritium has a half-life of only about 12 years; thus stocks of this radioactive hydrogen isotope have to be continuously replenished. The United States, for example, operates several reactors moderated and cooled by heavy water that produce tritium at the Savannah River facility in South Carolina.

      The plutonium isotope that is most desirable for sophisticated nuclear weapons is plutonium-239. If plutonium-239 is left in a reactor for a long time after production, plutonium-240 builds up as an undesirable contaminant. Accordingly, a major feature of a production reactor is its capability for quick throughput of fuel at a low energy-production level. Any reactor that can be operated this way is a potential production reactor.

      The world's first plutonium production reactors, built by the United States at Hanford, Wash., were fueled with natural uranium, moderated by graphite, and cooled by light water. It is believed that the early Soviet production reactors were the same sort, and the French and British versions differed only in that they were cooled with gas. As was noted above, the first significant power reactor, the Calder Hall reactor, was actually a dual-purpose production reactor.

Specialized reactors
      Nuclear reactors have been developed to provide electric power and steam heat in far-removed, isolated areas. Russia (Union of Soviet Socialist Republics), for instance, operates smaller power reactors specially designed to supply both electricity and steam for heating to accommodate the needs of a number of remote Arctic communities. Independent developmental work on small automatically operated reactors with similar capabilities has been undertaken by Sweden and Canada.

      Reactors have been developed to supply power and propulsion in space. The Soviet Union deployed small intermediate reactors in satellites for powering equipment and telemetry during the 1970s and '80s, but this policy became a target for criticism because at least one reactor-powered spacecraft reentered the atmosphere and deposited radioactive debris in Canada. Developmental activity in the United States has been directed largely toward reactor applications for the Strategic Defense Initiative (SDI) and for such deep-space missions as manned exploration of other planets or the establishment of a permanent lunar base. Reactors for these applications would necessarily be high-temperature systems based on either the HTGR or the LMR design but would use enriched fuel. A power cycle in space must be run at a very high temperature to minimize the size of the radiator from which heat is to be rejected. A reactor for space applications also has to be compact so that it can be shielded with a minimum amount of material.

      Small pressurized-water reactors have been used in the past to provide power for remote bases in Greenland and Antarctica. Though they have been replaced with oil-fired power plants, it still appears feasible to employ nuclear power for such applications or even for more exotic ones, such as supplying power to permanent undersea camps.

      Finally, concepts have been developed, notably in Germany, for employing HTGR systems as sources of high-temperature heat for chemical process industries. An idea that has drawn particular attention involves the use of reactor-generated heat at the mouth of a coal mine to convert the coal into clean gas for delivery by pipeline. Such processes remain economically unattractive at present but may ultimately became feasible as natural sources of fluid fuels are exhausted.

Reactor safety
      Nuclear reactors contain very large amounts of radioactive isotopes—mostly fission products but also such heavy elements as plutonium. If this radioactivity were to escape the reactor, its effects on the people in the vicinity would be severe. The deleterious effects of exposure to high levels of ionizing radiation would include increased rates of cancer and genetic defects, an increased number of developmental abnormalities in children exposed in the womb, and even death within a period of several days to months when irradiation is extreme (see radiation: Major types of radiation injury (radiation)). For this reason, a major consideration in reactor design is ensuring that a significant release of radioactivity (fallout) does not occur. This is accomplished by a combination of preventive measures and mitigating measures. Preventive measures are those that are taken to avoid accidents, and mitigating measures are those that decrease the adverse consequences. Essentially, preventive measures are the set of design and operating rules that are intended to make certain that the reactor is operated safely, while mitigating measures are systems and structures that prevent such accidents as do occur from proceeding to a catastrophic conclusion. Among the most well-known preventive measures are the reports and inspections for double-checking that a plant is properly constructed; rules of operation; and qualification tests for operating personnel to ensure that they know their jobs. The mitigating measures include safety rod systems for quickly shutting down a reactor to prevent a runaway chain reaction; emergency cooling systems for removing the heat of radioactive decay in the event that normal cooling capability is lost; and the containment structure for confining any radioactivity that might escape the primary reactor system. An extreme mitigating measure is the exercising of plans to evacuate personnel who might otherwise be heavily exposed in a reactor installation.

Preventive measures
      Since no human activity can be shown to be absolutely safe, all these measures cannot reduce the risks to zero, but it is the aim of the rules and safety systems to minimize the risk to the point where a reasonable individual would conclude they are trivial. What this de minimis risk value is, and whether it has been achieved by the nuclear industry, is a subject of bitter controversy, but it is generally accepted that independent regulatory agencies—the United States Nuclear Regulatory Commission (NRC) and similar agencies around the world—are the proper judges of such matters.

      To help evaluate the risks from nuclear power plants, the U.S. Atomic Energy Commission (AEC) authorized a major safety study in 1972 (the AEC was disbanded in 1974 and its functions have been assumed by the NRC). The study was conducted with major assistance from a number of laboratories, and it involved the application of probabilistic risk assessment (PRA) techniques for the first time on a system as complex as a large nuclear power reactor. This work resulted in the publication in 1975 of a report titled Reactor Safety Study, also known as WASH-1400. The most useful aspect of the study was its delineation of components and accident sequences (scenarios) that were determined to be the most significant contributors to severe accidents.

      The Reactor Safety Study concluded that the risks of an accident that would injure a large number of people were extremely low for the light-water reactor systems analyzed. This conclusion, however, was subject to very large quantitative uncertainties and was challenged.

      One basic problem with probabilistic risk assessment is that it cannot easily be confirmed by experience when the level of risk has been reduced to low values. That is to say, if probabilistic risk assessment predicts that a reactor is subject to, say, one failure in 10,000 years, there is no way to prove that statement with only a few, or even with 10,000, years of experience. Thus, the results of the Reactor Safety Study as to risk levels were not confirmable.

      There matters stood until 1979, when Three Mile Island Unit 2 suffered a severe accident. Through a combination of operator errors, coupled with the failure of an important valve to operate correctly, cooling water to the core was lost, parts of the core were melted and the rest of it destroyed, and a large quantity of fission products was released from the primary reactor system to the interior of the containment structure. The containment vessel of the reactor building fulfilled its function, and only a small amount of radioactivity was released, demonstrating the wisdom of having this component. Still, a severe accident had occurred.

      Many investigations of the Three Mile Island accident followed. Recommendations differed among them, but a common thread was that the human element was a much more important factor in safe operation than had been theretofore recognized. The human element pertained not only to the operating staff but also to the managements of nuclear plants and even to the NRC itself. Following the accident, therefore, many changes in operator training and in technical and inspectorate staffing were implemented, just as a number of hardware enhancements were introduced. It is generally believed that these changes have been effective in reducing the likelihood of the occurrence of accidents as severe as that at Three Mile Island. As a side issue to this, however, the operating costs of nuclear power plants have escalated sharply as more and more highly trained people have been added to the operating staffs.

      One area where probabilistic risk assessment has proved useful is with regard to the licensing of new plants, either light-water reactor installations or those of less common reactor types. PRA has the virtue of comparing systems fairly reliably. With better computer hardware and software than were available in 1975, it has become feasible to do PRA analyses of individual plants and compare them. A standard protocol for the NRC in licensing new, and particularly new types of, plants has therefore been that they must demonstrate lower risks than light-water reactors, which have been accepted as the norm.

      The significance of the human element, particularly as it relates to plant management and high-level regulatory decision making, was borne out again by the Chernobyl (Chernobyl accident) catastrophe of 1986. One of the four reactors in a nuclear power station about 100 kilometres north of Kiev exploded and caught fire as the result of an ill-conceived experiment (a test to see how long the steam turbines would run while coasting to a stop if the reactor would be abruptly shut down). Before the situation had been brought under control, 31 people had died (two from the blast and 29 from radiation exposure), an estimated 25 percent of the radioactive contents of the reactor had been released in a high cloud plume, 135,000 people had to be evacuated, and a large area surrounding the plant received fallout so great that it could not be farmed or pastured. Significant radiation was detected as far north as Scandinavia and as far west as Switzerland. It has been estimated that between 4,000 and 40,000 cases of cancer would ultimately result from this accident (besides the initial several hundred victims), mostly within Ukraine but some in areas far removed from there. Investigation of the accident placed the largest blame, as with the Three Mile Island mishap, on poor management both at the plant and within the government bureaucracy.

      Because all such nuclear plant accidents have basically resulted from human failings rather than from some intrinsic factor, most experts believe that nuclear energy can be a safe source of power. A review of the overall performance record shows that there had been, as of 1989, several thousand “reactor-years” of safe power-reactor operation in the Western world, with health effects less damaging than those associated with the extraction of an equal amount of power from coal. Incorporating the lessons learned from past accidents should certainly make future operations safer. There is, however, a condition on the conclusion that nuclear power is by and large a safe form of power. The facilities for generating this power must be designed, built, and operated to high standards by knowledgeable, well-trained professionals; and a regulatory mechanism capable of enforcing these standards must be in place.

Mitigating measures
      Two of the principal safety measures, the safety rods and the containment structure, have already been described. Other major safety systems are the emergency core cooling system, which makes it possible to cool the reactor if normal cooling is disrupted, and the emergency power system, which is designed to supply electrical power in case the normal supply is disrupted so that detectors and vital pumps and valves can continue to be operated. An important part of the safety system is the strict adherence to design rules, some of which have been mentioned—namely, the reactor should have a negative power-reactivity coefficient; the safety rods must be injectable under all circumstances; and no single regulating rod should be able to add substantial reactivity rapidly. Another important design rule is that the structural materials used in the reactor must retain acceptable physical properties over their expected service life. Finally, construction is to be covered by stringent quality assurance rules, and both design and construction must be in accordance with standards set by major engineering societies and accepted by the NRC.

      According to probabilistic risk assessment studies, three kinds of events are most responsible for the risks associated with light-water reactors—namely, station blackout, transient without scram, and loss of cooling. The nature of each of these mishaps is delineated, as are the proposed countermeasures and the anticipated risks.

      In station blackout, a failure in the power line to which the station is connected is postulated. The proposed emergency defense is a secondary electrical system, typically a combination of diesel generators big enough to drive the pumps and a battery supply sufficient to run the instruments. The risk would be that of the emergency generators not accepting load when they are started up. In transient without scram, the event is insertion of reactivity, for example, by an unchecked withdrawal of shim rods. The protective response is the rapid and automatic insertion of the safety rods. The risk would be the safety rods not functioning properly. In loss of cooling, the event is a failure of the normal cooling system to operate, either because of a break in a coolant line or because of an operator error. The emergency response is activation of the emergency core cooling system, and the risk would be that the system fails to operate. The ultimate event in the chain that led to the Three Mile Island accident was loss of emergency cooling by operator action owing to a misinterpretation of what sort of accident was occurring. In all these cases, proper operator action as well as proper functioning of the appropriate backup system are important aspects of emergency response. A final backup capability that is coming into play is the use of computers in an advisory mode to help the operator understand what is happening and suggest proper responses.

      Different reactor types pose different types of risk. For example, neither the pool-type liquid-metal reactor nor the high-temperature gas-cooled reactor are at major risk with regard to loss of cooling and perhaps not with regard to station blackout. However, the LMR, and perhaps the HTGR, are at some risk from events that might cause air or water to enter the coolant system. The hazard is that reactor materials, sodium or graphite, could chemically react with air and water. The hazard is greater with sodium in the LMR than it is with graphite in the HTGR.

      Another type of risk arises from external events, such as the possibility that earthquakes might initiate one or another major accident. The earthquake risk is minimized by building plants away from faults and by making use of earthquake-resistant mechanical design and construction.

Nuclear fuel cycle
      No discussion of nuclear power can be complete without a brief exposition of the nuclear fuel cycle. The whole point of a reactor is, after all, to cause fission in nuclear fuel. Moreover, it has turned out that low cost of fueling is the chief reason for the economic competitiveness of nuclear power. The principal steps of the fuel cycle are uranium mining and extraction from its ore (milling), uranium enrichment, fuel fabrication, loading and irradiation in the reactor (fuel management), unloading and cooling, reprocessing, waste packaging, and waste disposal.

uranium mining
      Uranium is mined from ores whose uranium content is on the order of 0.1 percent (one part per thousand). Most ore deposits are at or near the surface, and whether they are mined by open-pit or deep-mining techniques depends on the depth of the deposit and whether it slopes downward. The ore is crushed and the uranium chemically extracted from it at the mouth of the mine. The residue remains radioactive as it contains long-lived radioactive daughter nuclei of uranium and has to be carefully managed to minimize the release of radioactive contaminants into the environment. The uranium concentrate, which consists of uranium compounds (typically 75 to 95 percent), is shipped to a chemical plant for further purification and chemical conversion.

Enrichment
      There are several possible enrichment methods, but the only two that are used on a large scale are gaseous diffusion and gas centrifuging. In gaseous diffusion, natural uranium in the form of uranium hexafluoride gas (UF6), a product of chemical conversion, is allowed to seep through a porous barrier. The molecules of 235UF6 penetrate the barrier slightly faster than those of 238UF6. Since the percentage of 235U increases by only a very small amount after traversal of the barrier, the process must be repeated over and over in a large number of stages to obtain the desired amount of enrichment.

      In gas centrifuging, the uranium hexafluoride gas is fed into a high-speed centrifuge. The lighter species of this mixture of gaseous molecules including 235U tend to concentrate away from the wall, while the heavier ones accumulate along the wall. The degree of enrichment per stage in a centrifuge is greater than that obtained in a gaseous diffusion chamber, but the centrifuge is a more expensive piece of equipment.

Fabrication
      This step involves the conversion of the suitably enriched product material to the chemical form desired for reactor fuel. As of the late 1980s the only fuel fabricated on a large scale was that for light-water reactors.

      The chemical form prepared for the light-water reactor is uranium dioxide. Produced in the form of a ceramic powder, this compound is ground into a very fine flour and inserted into a die, where it is pressed into a pellet shape. Next the pellet is sintered in a furnace at 1,500–1,800° C. This sintering, similar to the firing of other ceramic ware, produces a dense ceramic pellet. Such pellets are loaded into prefabricated zirconium alloy cladding tubes, which are then filled with an inert gas and welded shut. These tubes, or pins, are bundled together with proper spacing assured by top and bottom grid plates through which the ends of the pins pass. Together with other necessary hardware, the bundle constitutes a fuel assembly.

Fuel management
      Fuel is loaded into a reactor in a careful pattern so as to obtain the most energy production from it before it becomes no longer usable. Fresh fuel is more reactive than old fuel, and this reactivity is used to keep the reactor critical. Typically, a reactor is fueled in cycles, each cycle lasting one to two years, and a fuel batch is kept in the reactor for three or four cycles. At the end of each cycle, the oldest fuel is removed and fresh fuel loaded. The partially burned fuel that remains, however, is shuffled before the fresh fuel is installed. The objective of this procedure is to achieve a loading of maximum reactivity while keeping the power distribution among the different fuel assemblies within technical specifications.

      Fuel burnup—that is, energy production—is limited by two factors. After significant burnup has occurred, the physical properties of the fuel become degraded and it is not prudent to continue to keep it in the reactor. Also, after some burnup, the old fuel no longer contributes useful reactivity to the reactor. The fuel design, including its initial enrichment, is such that these two limits are made to approximately coincide.

Unloading and cooling
      Spent reactor fuel is extremely radioactive, and its radioactivity also makes it a source of heat. When the spent fuel is removed from the reactor, it must continue to be both shielded and cooled. This is accomplished by placing the spent fuel in a water storage pool located next to the reactor. The water in the pool contains a large amount of dissolved boric acid, which is a heavy absorber of neutrons; this assures that the fuel assemblies in the pool will not go critical. (Pool water is also a common source of emergency cooling water for the reactor.) Pools vary in size; the older ones are able to accommodate only about 10 years worth of spent fuel. As the pools fill up, more spent fuel storage is needed. As noted earlier, additional storage space can be gained by loading spent fuel into the pool more densely than originally planned, by building a new pool, or by removing the oldest fuel assemblies from the existing pool and storing them in air-cooled concrete and steel silos located above ground. This last method becomes feasible after fuel has been stored for two or three years because radioactivity and heat generation decrease rapidly over this period. Dense storage in existing pools and silo storage both seem to be less expensive than building a new pool.

Reprocessing (recycling)
      Both the converted plutonium and residual uranium-235 in spent fuel can be recycled. Such materials can be recovered by chemically reprocessing the fuel. Equally as significant, reprocessing can reduce the volume and radioactivity of the waste material, which must ultimately be eliminated by some method of permanent disposal. Until 1975 it was generally assumed that after two to five years spent fuel would be delivered to a reprocessing plant. By that time, however, the cost of reprocessing had escalated to a point where its economics became questionable. Also, during the period 1976–81, it was U.S. policy, by presidential directive, not to reprocess. The directive has since been rescinded, but reprocessing is still not done commercially in the United States.

      Policy and institutional arrangements are different in France and Britain (United Kingdom). Commercial reprocessing plants exist in both countries and are processing spent fuel not only from nuclear plants in the host countries but also from those in others. The reprocessed plutonium can be used not only as fuel for planned future liquid-metal reactors but also to help fuel existing light-water reactors. In the latter application, the plutonium is utilized in mixed oxide form—a combination of uranium and plutonium dioxides having 3 to 6 percent plutonium.

      Reprocessing is accomplished by dissolving the spent fuel in nitric acid and contacting the acid solution with oil in which tributyl phosphate (TBP) is dissolved. TBP is a complexing agent for uranium and plutonium, forming compounds with them that bring them into the oil solution. A physical separation of the (immiscible) oil and acid serves to remove the desired products from the nitric acid solution, which still contains all the fission products. The uranium and plutonium can then be washed out of the TBP back into a water solution and separated from each other to the degree desired by means of various techniques. Thus, reprocessing produces three product streams: (1) a purified uranium product, (2) a plutonium product that may be either pure or mixed with uranium, and (3) a waste stream of fission products dissolved in nitric acid.

Waste conditioning
      In the absence of reprocessing, the spent fuel is considered to be waste and must be prepared for disposal. This operation is to be performed in a separate facility, for which the Department of Energy has responsibility in the United States. As of 1998, the department is to begin receiving spent fuel from utilities largely on an “oldest-fuel-first” schedule. After brief storage, the fuel pins would be removed from their assemblies. End pieces that contain no fuel would be removed and the pins repacked into a dense lattice emplaced in a corrosion-resistant steel canister. A cover would be welded on and the canister covered with an overpack. This would represent the basic waste form for spent-fuel disposal.

      Some waste exists in the form of the fission-product solution that arises from reprocessing. Reprocessed fuel from production reactors also generates this type of waste. The waste solution is completely evaporated, leaving behind the fission products in the solid residue, which is heated until all the constituent nitrate salts are converted to oxides. These oxides are then put into a glass-forming oven and mixed with materials that will produce a borosilicate glass (Pyrex). The fission-product oxides dissolve in the glass as it forms. The glass melt is subsequently poured into a steel canister, 200–400 millimetres in diameter and about one metre high, where it solidifies into a solid glass block. Once covered with an overpack of bentonite clay, the solid canister-like block is ready for disposal.

      The glassmaking process for waste conditioning described here is operational on an industrial scale in France and has been tested in many other countries, including the United States.

Waste disposal (refuse disposal system)
Proposed method
      The waste disposal method currently being planned by all countries with nuclear power plants is called geologic disposal. This means that all conditioned nuclear wastes are to be deposited in mined cavities deep underground. Shafts are to be sunk into a solid rock stratum, with tunnel corridors extending horizontally from the central shaft region and tunnel “rooms” laterally from the corridors. The waste would be emplaced (probably by remotely controlled or robotic devices) in holes drilled into the floors of these rooms, after which the boreholes would be sealed and the rooms and corridors backfilled. When the entire operation is completed (perhaps after about 30 years of operation), the shafts too would be backfilled and sealed.

Risks of nuclear waste disposal
      When a holistic view is taken of the nuclear waste disposal process, the risks seem extremely small, yet among the general public these risks are one of the most feared aspects of the nuclear fuel cycle. A great deal of suspicion about the process arises from the numerous incidents of mismanagement of other types of waste, and these fears have been encouraged by antinuclear activists. A number of basic observations on the process of geologic disposal point to the difficulty of resolving differences that are founded on perceptual discrepancies.

      Nuclear waste retains its very intense level of radioactivity for several hundred years, but after 1,000 years have passed the remaining radioactivity, while persistent, is at a level comparable to, but greater than, that of a body of natural uranium ore. This separates the safety problem into two time periods: a first millennium during which it is crucial to ensure tight retention of the wastes in the repository, and a subsequent period during which it is only necessary to ensure that any release that occurs is small and slow.

      The impingement of groundwater and subsequent corrosion of the waste canisters, followed by dissolution of the waste, provides a possible route for the emergence of the waste in the surface environment. Water migrates slowly in most rock formations. Contrary to the popular belief that any dissolution of the waste and discharge of the resulting solution to the environment will quickly lead to high-level contamination, only a low level is projected, even in worst-case scenarios.

      Migration of radioactive species that has been observed at shallow burial sites for low-level radioactive waste is not an indication that similar migration can be expected in a deep underground repository. In addition to the near insolubility of the waste material, waste form engineering, particularly of corrosion-resistant containers, provides extra protection against such dispersal. Moreover, most of the dispersal problem in shallow disposal sites is caused by biochemical products that do not exist in deep formations; water found at depth is sterile.

      Finally, a great deal of care is to be expended in selecting the site of the repository. Site selection is probably the biggest problem, both politically and technically. Various conditions are mandatory: the repository must not be near a populated area; the rock stratum selected must be deep (300 metres or more) and, as much as possible, naturally sealed from aquifers; and any discharge of the water table into the surface waters should be slow. Furthermore, the site must be in a tectonically inactive zone so that earthquakes will not break that seal.

      The risk of high-level waste burial is almost certainly smaller than the risks of reactor accidents and even than the risks arising from improperly managed mine tailings. Nonetheless, the siting of a repository must be handled with political sensitivity, and the confirmation of acceptable hydrologic and geologic conditions must have a high degree of validity. There are many acceptable sites in principle, but confirming acceptability for any one of them is a large and expensive technical undertaking.

History of reactor development
      Soon after the discovery of nuclear fission was announced in 1939, it was also determined that the fissile isotope involved in the reaction was uranium-238 and that neutrons were emitted in the process. Newspaper articles reporting the discovery mentioned the possibility that a fission chain reaction could be exploited as a source of power. World War II, however, began in Europe in September of 1939, and physicists in fission research turned their thoughts to using the chain reaction in a bomb. It was quickly recognized that a high concentration of fissile material would be needed to accomplish this.

      Inasmuch as fission had been first discovered in Germany, there was great fear, particularly among refugee physicists from Europe who had fled to America, France, and Britain, that Nazi Germany might develop just such a bomb. As a result, these three countries began working toward the development of atomic bombs (atomic bomb), which at that point was still speculation. The most successful program was established in the United States, where President Franklin D. Roosevelt was persuaded by a letter from Albert Einstein to initiate a secret project devoted to this purpose. In early 1940 the U.S. government made funds available for research that eventually evolved into the Manhattan Project. After the fall of France to the German armies (1940), leading French researchers escaped to England and joined the ongoing British project. After the entry of the United States into the war in 1941, the British effort was transferred to the safer confines of North America. Though the British group participated in American research, it was chiefly concerned with initiating a research program in Canada.

      The Manhattan Project included work on uranium enrichment to procure uranium-235 in high concentrations and also research on reactor development. The goal was twofold: to learn more about the chain reaction for bomb design and to develop a way of producing a new element, plutonium, which was expected to be fissile and could be isolated from uranium chemically.

      Reactor development was placed under the supervision of the leading experimental nuclear physicist of the era, Enrico Fermi (Fermi, Enrico). Fermi's project, begun at Columbia University and first demonstrated at the University of Chicago, centred on the design of a graphite-moderated reactor. It was soon recognized that heavy water was a better moderator and would be more easily used in a reactor, and this possibility was assigned to the Canadian research team since heavy-water production facilities already existed in Canada. Fermi's work led the way, and on Dec. 2, 1942, he reported having produced the first self-sustaining chain reaction. His reactor, later called Chicago Pile No. 1 (CP-1), was made of pure graphite in which uranium metal slugs were loaded toward the centre with uranium oxide lumps around the edges. This device had no cooling system, as it was expected to be operated for purely experimental purposes at very low power. CP-1 was subsequently dismantled and reconstructed at a new laboratory site in the suburbs of Chicago, the original headquarters of what is now Argonne National Laboratory. The device saw continued service as a research reactor until it was finally decommissioned in 1953.

       Notable early nuclear reactorsOn the heels of the successful CP-1 experiment, plans were quickly drafted for the construction of the first production reactors. These were the early Hanford reactors, which were graphite-moderated, natural uranium-fueled, water-cooled devices. As a backup project, a production reactor of air-cooled design was built at Oak Ridge, Tenn.; when the Hanford facilities proved successful, this reactor was completed to serve as the X-10 reactor at what is now Oak Ridge National Laboratory. Shortly after the end of World War II, the Canadian project succeeded in building a zero-power, natural uranium-fueled research reactor, the so-called ZEEP (Zero-Energy Experimental Pile). The first enriched-fuel research reactor was completed at Los Alamos, N.M., at about this time as enriched uranium-235 became available for research purposes (see Table 3 (Notable early nuclear reactors)). In 1947 a 100-kilowatt reactor with a graphite moderator and uranium metal fuel was constructed in England, and a similar one was built in France the following year.

      In 1953 President Dwight D. Eisenhower of the United States announced the Atoms for Peace program. This program established the groundwork for a formal U.S. nuclear power program and expedited international cooperation on nuclear power.

      The earliest U.S. nuclear power project had been started in 1946 at Oak Ridge, but the program was abandoned in 1948, with most of its personnel being transferred to the naval reactor program that produced the first nuclear-powered submarine, the Nautilus. After 1953 the U.S. nuclear power program was devoted to the development of several reactor types, of which three ultimately proved to be successful in the sense that they remain as commercial reactor types or as systems scheduled for future commercial use. These three were the fast breeder reactor (now called LMR); the pressurized-water reactor; and the boiling-water reactor. The first LMR was the Experimental Breeder Reactor, EBR-I, which was designed at Argonne National Laboratory and constructed at what is now the Idaho National Engineering Laboratory near Idaho Falls, Idaho. EBR-I was an early experiment to demonstrate breeding, and in 1951 it produced electricity from nuclear heat for the first time. As part of the U.S. nuclear power program, a much larger experimental breeder, EBR-II was developed and put into service (with power generation) in 1963. The principle of the boiling-water reactor was first demonstrated in a research reactor in Oak Ridge, but development of this reactor type was also assigned to Argonne, which built a series of experimental systems designated BORAX in Idaho. One of these, BORAX-III, became the first U.S. reactor to put power into a utility line on a continuous basis. A true prototype, the Experimental Boiling Water Reactor, was commissioned in 1957. The principle of the pressurized-water reactor had already been demonstrated in naval reactors, and the Bettis Atomic Power Laboratory of the naval reactor program was assigned to build a civilian prototype at Shippingport, Pa. This reactor, the largest of the power-reactor prototypes, is often hailed as the first commercial-scale reactor in the United States.

      During the late 1950s and early 1960s a number of true commercial prototype nuclear power plants were built. Of these, the most successful was the light-water reactor system, although the advanced gas-cooled type remained the British standard for many years and the CANDU system prevailed in Canada. From the mid-1960s, larger units were ordered in the expectation of an ever-increasing commercial utilization of nuclear power, and by the early 1970s nuclear plant orders were coming in at such a rapid pace that the unit sizes were increased so as to reduce the number of separate projects that each vendor would have to staff for. By the later years of the decade, however, the surfeit of orders in the United States was followed by a large number of project cancellations. This phenomenon was the result of a sharp decrease over what had been projected as the rate of increase in base-load electricity demand for which the large nuclear plants were designed. The new plants were not needed. Moreover, the cost of new nuclear plants had begun to escalate to the point where their economics became questionable. Public fears of nuclear power, stimulated by the Three Mile Island accident, also were a factor.

      Similar scenarios have slowed the deployment of nuclear power in several countries besides the United States. On the other hand, France, Japan, South Korea, and Taiwan, which all have few alternative fuel resources, have continued building up their nuclear power capacity.

Bernard I. Spinrad

Additional Reading
Richard Rhodes, The Making of the Atomic Bomb (1986), chronicles developments leading to the first reactor and first atomic bomb. An elementary text covering reactor concepts, radiation, nuclear fuel cycles, reactor systems, safety and safeguards, and fusion concepts is Ronald Allen Knief, Nuclear Energy Technology: Theory and Practice of Commercial Nuclear Power (1981); the same concepts are treated at a more advanced mathematical level in John R. Lamarsh, Introduction to Nuclear Engineering, 2nd ed. (1983). James J. Duderstadt and Louis J. Hamilton, Nuclear Reactor Analysis (1976), discusses the theory of neutron behaviour in matter, criticality, neutron spectrum, and reactor core design and control, with emphasis on methods of calculation. Manson Benedict, Thomas H. Pigford, and Hans Wolfgang Levi, Nuclear Chemical Engineering, 2nd ed. (1981), includes coverage of fuel cycles, the chemistry of uranium and heavy elements, the theory of multistage systems, enrichment processes and theory, the reprocessing of nuclear fuel, and nuclear waste management. Current developments in domestic and international nuclear power, safety, research, and opinion are published in Nuclear News (monthly), the newsletter of the American Nuclear Society. Bernard I. Spinrad

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Universalium. 2010.

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