Nuclear reactor system utilizing nuclear fission in a controlled and self-sustaining manner. Neutrons are used to fission the nuclear fuel, and the

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1 Engineering & Materials: Other Engineering Disciplines:Nuclear engineering Nuclear reactor 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. See also: Chain reaction (physics); Nuclear fission Nuclear reactors are used in a variety of ways as sources for energy, for nuclear irradiations, and to produce special materials by transmutation reactions. Since the first demonstration of a nuclear reactor, made beneath the west stands of Stagg Field at the University of Chicago on December 2, 1942, many hundreds of nuclear reactors have been built and operated in the United States. Extreme diversification is possible with the materials available, and reactor power may vary from a fraction of a watt to thousands of megawatts. The size of a nuclear reactor core is governed by its power level, time between refueling, fuel environment, the factors affecting the control of the neutron chain reaction, and the effectiveness of the coolant in removing the fission energy released. 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 onethird 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 also: 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 % 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. This permits using smaller amounts and lower concentrations of fissile materials. See also: Neutron; 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. Reactors have been built in all three categories. The first fast reactor was the Los Alamos (New Mexico) assembly called Clementine, which operated from 1946 to The fuel core consisted of nickel-coated rods of pure plutonium metal, contained in a 6-in.-diameter (15-cm) steel pot. Coolants for fast reactors may be steam, gas, or liquid metals. Current fast reactors utilize liquid sodium as the coolant and are being developed for breeding and power. An example of an intermediate reactor was the first propulsion reactor for the submarine USS Seawolf. The fuel core consisted of enriched uranium with beryllium as a moderator; the original coolant was sodium, and the reactor operated from 1956 to Examples of thermal reactors, currently the basis of commercial nuclear power production, are given below. Fuel composition 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. Uranium-235 Naturally occurring uranium contains only 0.7% of the fissionable isotope uranium-235, the balance being essentially uranium-238. Uranium with higher concentrations of uranium-235 is called enriched uranium.

2 Uranium metal is susceptible to irradiation damage, which limits its operating life in a reactor. The life expectancy can be improved somewhat by heat treatment, and considerably more by alloying with elements such as zirconium or molybdenum. Uranium oxide exhibits better irradiation damage resistance and, in addition, is corrosion-resistant in water. Ceramics such as uranium oxide have lower thermal conductivities and lower densities than metals, which are disadvantageous in certain applications. See also: Alloy Current light-water-cooled nuclear power reactors utilize uranium oxide as a fuel, with an enrichment of several percent uranium-235. Cylindrical rods are the most common fuel-element configuration. They can be fabricated by compacting and sintering cylindrical pellets which are then assembled into metal tubes which are sealed. Developmental programs for attaining long-lived solid-fuel elements include studies with uranium oxide, uranium carbide, and other refractory uranium compounds. See also: Uranium Plutonium-239 Plutonium-239 is produced by neutron capture in uranium-238. It is a by-product in power reactors and is becoming increasingly available as nuclear power production increases. For example, Japan has produced over 10 tons of plutonium solely as the result of its commercial reactor program. However, plutonium as a commercial fuel is still at a demonstration stage of development in Europe and Japan. The commercial recycle of plutonium from processed spent fuel was deferred indefinitely in the United States by the Carter administration in April Plutonium is hazardous to handle because of its biological toxicity and, like radium, is a carcinogen if ingested. It is fabricated in glove boxes to ensure isolation from operating personnel. It can be alloyed with other metals and fabricated into various ceramic compounds. It is normally used in conjunction with uranium-238; alloys of uranium-plutonium, and mixtures of uranium-plutonium oxides and carbides, are of most interest as fuels. Except for the additional requirements imposed by plutonium toxicity and proliferation safeguards, much of the uranium technology is applicable to plutonium. For light-water nuclear power reactors, the oxide fuel pellets are contained in a zirconium alloy (Zircaloy) tube. Stainless steel tubes are used for containing the oxide fuel for the fast breeder reactors. See also: Plutonium Uranium-233 Uranium-233, like plutonium, does not occur naturally, but is produced by neutron absorption in thorium-232, a process similar to that by which plutonium is produced from uranium-238. Interest in uranium-233 arises from its favorable nuclear properties and the abundance of thorium. However, studies of this fuel cycle are at a relatively early stage. Uranium-233 also imposes special handling problems because of proliferation concerns and the radiological toxicity of the daughter products of another uranium isotope (uranium-232) present in the fuel cycle. It does not, however, introduce new metallurgical problems. Thorium is metallurgically different, but it has favorable properties both as a metal and as a ceramic. See also: Nuclear fuels; Thorium Fuel configurations Fuel-moderator assemblies may be homogeneous or heterogeneous. Homogeneous assemblies include the aqueous-solution-type water boilers and molten-salt-solution dispersions, slurries, and suspensions. The few homogeneous reactors built have been used for limited research and for demonstration of the principles and design features. In heterogeneous assemblies, the fuel and moderator form separate solid or liquid phases, such as solid-fuel elements spaced either in a graphite matrix or in a water phase. Most power reactors utilize an arrangement of closely spaced, solid fuel rods, about in. (13 mm) in diameter and 12 ft (3.7 m) long, in water. In the arrangement shown in Fig. 1, fuel rods are arranged in a grid pattern to form a fuel assembly, and over 200 fuel assemblies are in turn arranged in a grid pattern in the reactor core. Fig. 1 Arrangement of fuel in the core of a pressurized-water reactor, a typical heterogeneous reactor. (a) Fuel rod; (b) side view (CEA = control element assembly), (c) top view, and (d) bottom view of fuel assembly; (e) cross section of reactor core showing arrangement of fuel assemblies; (f) cross section of two adjacent fuel assemblies, showing arrangement of fuel rods. 1 in. = 25 mm. (Combustion Engineering, Inc.)

3 Heat Removal 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. Heat is not generated uniformly in a reactor core. The heat flux generally decreases axially and radially from a peak near the center of the reactor. In addition, local perturbations in heat generation occur because of the arrangement of the reactor fuel, its burn-up, various neutron "poisons" used to shape the power distribution, and inhomogeneities in the reactor structure. These variations impose special considerations in the design of reactor cooling systems, including the need for establishing variations in coolant flow rate through the reactor core to achieve uniform temperature rise in the coolant, avoiding local hot-spot conditions, and avoiding local thermal stresses and distortions in the structural members of the reactor. Nuclear reactors have the unique thermal characteristic that heat generation continues after shutdown because of fission and radioactive decay of fission products. Significant fission heat generation occurs only for a few seconds after shutdown. Radioactive-decay heating, however, varies with the decay characteristics of the mixture of fission products and persists at low but significant power levels for many days. See also: Radioactivity Accurate analysis of fission heat generation as a function of time immediately after reactor shutdown requires detailed knowledge of the speed and reactivity worth of the control rods. The longer-term fission-product-decay heating, on the other hand, depends upon the time and power level of prior reactor operation and the isotopic composition of the fuel. Typical values of the total heat generation after shutdown (as percent of operating power) are 10-20% after 1 s, 5-10% after 10 s, approximately 2% after 10 min, 1.5% after 1 h, and 0.7% after 1 day. These rates are important in reactor safety since 0.7% of the thermal power (2564 MW) of a 1000-MW(e) commercial nuclear power plant is approximately 18 MW of heat still being generated 1 day after the reactor is shut down. Reactor coolants Coolants are selected for specific applications on the basis of their heat-transfer capability, physical properties, and nuclear properties. Water 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; Fig. 2), the water boils directly in the reactor core to make steam that is piped to the turbine. In a pressurized-water reactor (PWR; Fig. 3), 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. Figure 4 shows the relation of the core and heat removal systems to the condenser, electric power system, and waste management system in the Prairie Island (Minnesota) nuclear plant, which is typical of plants using pressurized-water reactors.

4 Coolant intake water is pumped through hundreds of 1-in.-diameter (25-mm) tubes in the condenser, and the warm water from the condenser is then pumped over cooling towers and returned to the plant. See also: Cooling tower; Radioactive waste management; Vapor condenser Fig. 2 Boiling-water reactor. (Atomic Industrial Forum, Inc.) Fig. 3 Pressurized-water reactor. (Atomic Industrial Forum, Inc.) Fig. 4 Nuclear plant, using pressurized-water reactors. (Northern States Power Company)

5 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 approximate 1/500 that for light water that makes it possible to operate reactors using heavy water with natural uranium fuel. There is no serious neutron-activation problem with pure water; 16 N, formed by the (n,p) reaction with 16 O (absorption of a neutron followed by emission of a proton), is a major source of activity, but its 7.5-s half-life minimizes this problem since the radioactivity, never very high to begin with, quickly decays away. The most serious limitation of water as a coolant for power reactors is its high vapor pressure. A coolant temperature of 550 F (288 C) requires a system pressure of at least 1100 psi (7.3 megapascals). This temperature is below fossil-fuel power station practice, for which steam temperatures near 1000 F (538 C) have become common. Lower thermal efficiencies result from lower temperatures. Boiling-water reactors operate at about 70 atm (7 MPa), and pressurizedwater reactors at 150 atm (15 MPa). The high pressure necessary for water-cooled power reactors determines much of the plant design. This will be discussed below. See also: Nuclear reaction Gases 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 (Fig. 5), in which the gas transfers heat from the reactor core to a steam generator. The British advanced gas reactor (AGR), however, uses carbon dioxide (CO 2 ). Gases are capable of operation at extremely high temperature, and they are being considered for special process applications and direct-cycle gas-turbine applications. Hydrogen was used as the coolant for the reactor developed in the Nuclear Engine Rocket Vehicle Application (NERVA) Program, now terminated. Heated gas discharging through the nozzle developed the propulsive thrust. Fig. 5 High-temperature gas-cooled reactor. (Atomic Industrial Forum, Inc.)

6 Liquid metals The alkali metals, in particular, have excellent heat-transfer properties and extremely low vapor pressures at temperatures of interest for power generation. Sodium vapor pressure is less than 17.6 lb/in. 2 (120 kilopascals) at 1650 F (900 C). 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; Fig. 6), 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. Fig. 6 Loop-type liquid-metal fast breeder reactor. In some designs, referred to as pool-type, the heat exchanger and primary sodium pump are located with the core inside the pressure vessel. (Atomic Industrial Forum, Inc.) Sodium-24, formed by the absorption of a neutron, is an energetic gamma emitter with a 15-h halflife. The primary system is surrounded by biological shielding, and approximately 2 weeks is required for decay of 24 Na activity prior to access to the system for repair or maintenance. Another sodium isotope, 22 Na, has a 2.6-year half-life and builds up slowly with plant operation until eventually the radioactivity reaches a level where it is necessary to drain the sodium before maintenance can be performed. Sodium does not decompose, and no makeup is required. However, sodium reacts violently if mixed with water. This requires extreme care in the design and fabrication of sodium-to-water steam generators and backup systems to cope with occasional leaks. The poor lubricating properties of sodium and its reaction with air further specify the mechanical design requirements of sodium-

7 cooled reactors. Nevertheless, sodium-cooled reactors have operated with good reliability and relatively high operating availability. The other alkali metals exhibit similar characteristics but appear to be less attractive than sodium. The eutectic alloy of sodium with potassium (NaK), however, has the advantage that it remains liquid at room temperature, but adversely affects the properties of steel used in system components. Mercury has also been used as a coolant but its overall properties are less favorable than sodium. Plant balance 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 (the energy of motion) 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 also: Enthalpy; Generator; Prime mover; Steam turbine After the steam expands to spin the rotor in the turbine, it exits into the condenser where it is cooled to produce condensate water, which is then fed back to the core or to the steam generator, and the cycle is repeated. The condenser is a very large and important part of the plant. Roughly twice the amount of heat that has been converted in the turbine for the production of electric power is removed in the condenser for further rejection to the environment. The second law of thermodynamics defines the fraction of useful energy or work that can be attained by a thermal engine. Large amounts of water circulate through the condenser to carry the waste heat to its ultimate sink, which may be the sea, a river, a lake, or the atmosphere itself through cooling towers. See also: Cooling tower; Steam condenser; Thermodynamic principles The energy conversion part of the power plant (that is, excluding the reactor itself with its main components and system) is often called balance of plant (BOP). It incorporates a large number of components and systems. It represents over four-fifths of the plant's total cost, and is important for the efficient and safe operation of the plant. 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. See also: Fluid flow; Hydrodynamics 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 also: Conduction (heat); Heat capacity; Viscosity Secondary considerations include other physical properties of the coolant, particularly its vapor pressure. If the vapor pressure is high at the operating temperature, local or bulk boiling of the fluid may occur, unless the system pressure is maintained at a higher pressure at all times. This, in turn, must be considered in establishing the heat transfer coefficient for the fluid. See also: Vapor pressure Because the coolant absorbs and scatters neutrons, variations in coolant density also affect reactor performance and control. This is particularly significant in reactors in which the coolant exists in two phases for example, the liquid and vapor phases in boiling systems. Gases, of course, do not undergo phase change, nor do liquids operating at temperatures well below their boiling point; however, the fuel density does change with temperature and may have an important effect upon the reactor. Power generation and, therefore, the heat removal rate are not uniform throughout the reactor. If the mass flow rate of the coolant is uniform through the reactor core, then unequal temperature rise of the coolant results. This becomes particularly significant in power reactors in which it is desired to achieve the highest possible coolant outlet temperature to attain maximum thermal efficiency of the power cycle. The performance limit of the coolant is set by the temperature in the hottest region or channel of the reactor. Unless the coolant flow rate is adjusted in the other regions of the reactor, the coolant will leave these regions at a lower temperature and thus will reduce the average coolant outlet temperature. In power reactors, this effect is reduced by adjusting the rate of flow to each region of the reactor in proportion to its heat generation rate. This involves very careful design and

8 analysis of the system. In the boiling-type reactor, this effect upon coolant temperature does not occur because the exit temperature of the coolant is constant at the saturation temperature for the system. However, the variation in power generation in the reactor is reflected by a difference in the amount of steam generated in the various zones, and orificing is still required to achieve most effective use of coolant flow. In some power reactors, the flow rate and consequent pressure drop of the coolant are sufficient to create large mechanical forces in the system. It is possible for the pressure drop through the fuel assemblies to exceed the weight of the fuel elements in the reactor, with a resulting hydraulic lifting force on the fuel elements. Often this requires a design arrangement to hold the fuel elements down. Although this problem can be overcome by employing downward flow through the system, it is often undesirable to do so because of shutdown-cooling considerations. It is desirable in most systems to accomplish shutdown cooling by natural-convection circulation of the coolant. If downflow is employed for forced circulation, then shutdown cooling by natural-convection circulation requires a flow reversal, which can introduce new problems. Therefore, hydraulic forces are overcome by use of support plates, seismic restraints, and fuel spacers. Thermal stress considerations The temperature of the reactor coolant increases as it circulates through the reactor core. This increase in temperature is constant at steady-state conditions. Fluctuations in power level or in coolant flow rate result in variations in the temperature rise. These are reflected as temperature changes in the coolant core exit temperature, which in turn result in temperature changes in the coolant system. 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. The consequent temperature stresses induced in the various parts of the system are superimposed upon the thermal stresses that exist under normal steady-state operations and produce conditions known as thermal shock or thermal cycling. In some systems, it is not uncommon for the thermal stresses to be significant. In these cases, careful attention must be given to the transient stresses, and thermal shielding (such as thermal sleeves on pipes and baffles) is commonly employed in critical sections of the system. Normally, this consists of a thermal barrier which, by virtue of its heat capacity and insulating effect, delays the transfer of heat, thereby reducing the rate of change of temperature and protecting critical system components from thermal stresses. Thermal stresses are also important in the design of reactor fuel elements. Metals that possess dissimilar thermal-expansion coefficients are frequently required. Heating of such systems gives rise to distortions, which in turn can result in flow restrictions in coolant passages. Careful analysis and experimental verification are often required to avoid such circumstances. 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 also: Centrifugal pump; Electromagnetic pump In addition to the variety of special pumps developed for reactor coolant systems, there is a variety of piping system components and heating exchange components. As in all flow systems, flowregulating devices such as valves are required, as well as flow instrumentation to measure and thereby control the systems. Here again, leak tightness has necessitated the development of valves with special seals such as metallic bellows around the valve stem to ensure system integrity. Measurement of flow and pressure has also required the development of sensing instrumentation that is reliable and leak-tight. See also: Flow measurement; Pressure measurement; Valve Many of these developments have borrowed from other technologies where toxic or flammable fluids are frequently pumped. In many cases, however, special equipment has been developed specifically to meet the requirements of the reactor systems. An example of this type of

9 development involves the measurement of flow in liquid-metal piping systems. The simple principle of a moving conductor in a magnetic field is employed by placing a magnet around the pipe and measuring the voltage generated by the moving conductor (coolant) in terms of flow rate. Temperature compensation is required, and calibration is important. Although the development of nuclear power reactors has introduced many new technologies, no method has yet displaced the conventional steam cycle for converting thermal energy to mechanical energy. Steam is generated either directly in the reactor (direct-cycle boiling reactor) or in auxilliary steam generation equipment, in which steam is generated by transfer of heat to water from the reactor coolant. These steam generators require a very special design, particularly when dissimilar fluids are involved. Typical of the latter problem is the sodium-to-water steam generators in which integrity is essential because of the potentially violent chemical reaction between sodium and water. Core Design and Materials A typical reactor core for a power reactor consists of the fuel element rods supported by a grid-type structure inside a vessel (Fig. 1). The primary function of the vessel is to contain the coolant. Its design and materials are determined by such factors as the nature of the coolant (corrosive properties) and quantity and configuration of fuel. The vessel has several large nozzles for coolant entrance and exit and smaller nozzles used for controlling reactor operation (control-rod drive mechanisms and instruments). The top of the vessel unbolts for refueling operations. The pressure vessel design takes account of thermal stresses caused by temperature differences in the system. An exceptionally high degree of integrity is demanded of this equipment. Reactors are designed to permit removal of the internals from the vessel, which is then periodically inspected by automatic devices that can detect any cracks which might have developed during operation. These in-service inspections are required by codes and regulations on a fixed time schedule. Structural materials 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. Some requirements are established because of secondary effects; for example, corrosion limits may be established by the rate of deposition of coolant-entrained corrosion products on critical surfaces rather than by the rate of corrosion of the base material. See also: Corrosion 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. See also: Aluminum; Stainless steel; Zirconium The 18-8 series stainless steels have been used for structural members in both water-cooled reactors and sodium-cooled reactors because of their corrosion resistance and favorable physical properties at high temperatures. Type 304, 316, and 347 stainless steel have been used the most extensively because of their weldability, machinability, and physical properties, although other ironnickel alloys, such as Inconel, are also used. To increase reliability and to reduce cost, heavy-walled pressure vessels are normally fabricated from carbon steels and clad on the internal surfaces with a thin layer of stainless steel to provide the necessary corrosion resistance. See also: Iron alloys; Nickel alloys; Steel As the size of power reactors has increased, it has become necessary in some instances to fieldfabricate reactor vessels. A notable example is the on-site fabrication of the large pressure vessel required for the liquid-metal fast breeder reactor at Creys-Malville, France. This involved fieldwelding of wall sections and subsequent stress relieving. Both steel and prestressed concrete vessels for gas-cooled reactors are also field-fabricated. Research reactors operating at low temperatures and pressures introduced special experimental considerations. The primary objective is to provide the maximum volume of unperturbed neutron flux for experimentation. It is desirable, therefore, to extend the experimental irradiation facilities beyond the vessel wall. This has introduced the need for vessels constructed of materials having a low cross section for neutron capture. Relatively large low-pressure aluminum reactor vessels with wall sections as thin as practicable have been manufactured for research reactors. Fuel cladding 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;

10 however, its physical strength and corrosion resistance in water decrease very rapidly above about 300 F (149 C), and it is therefore used only in test and research reactors that are not used to produce power. Zirconium has favorable neutron properties, and in addition is corrosion-resistant in hightemperature water. It has found extensive use in water-cooled power reactors. The technology of zirconium production and the use of zirconium-based alloys, such as Zircaloy, have advanced tremendously under the impetus of the various reactor development programs. Stainless steel is used for the fuel cladding in fast reactors, in some light-water reactors for which neutron captures are less important. Control and Instrumentation A reactor is critical when the rate of production of neutrons equals the rate of absorption in the system plus the rate of leakage out of the core. 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 also: 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. It is necessary for orderly startup and control of a reactor that the neutron flux be sufficiently high to permit rapid power increase; in large reactors, too slow a startup is erratic and uneconomical. During reactor startup, a source of neutrons is useful, therefore, for control, and aids in the instrumentation of reactor systems. Neutrons are obtained from the photoneutron effect in materials such as beryllium. Neutron sources consist of a photon (gamma-ray) source and beryllium, such as antimony-beryllium. Antimony sources are particularly convenient for use in reactors because the antimony is reactivated by the reactor neutrons each time the reactor operates. Control drives and systems 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. Power reactors have inherent negative temperature coefficients that make it practical to use stepwise control of rod motions but with smooth and limited charges in power level. 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 scramsafety 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. Allowing the control rods to drop into the core by means of gravity is one common method. To enhance reliability of the scram system, its operation can be initiated by deenergizing appropriate electrical circuits. This also automatically produces reactor scram in the event of a system power failure. Hydraulic or pneumatic drive systems, as well as a variety of electromechanical systems, have also been developed. In addition to the actuating motions required, control-rod drive systems must also provide indication of the rod positions at all times. Various types of sensors, as well as arrangements of switch indicators, are employed as postion indicators. Instrumentation Reactor control involves continuing measurement of the condition of the reactor. Neutron-sensitive ion chambers may be located outside the reactor core, and the flux measurements from the detectors are combined to measure a flux that is proportional to the average neutron density in the reactor. The chamber current is calibrated against a thermal power measurement and then is applied over a wide range of reactor power levels. The neutron-sensitive detector system is capable of measuring the lowest neutron flux in the system, including that produced by the neutron source when the reactor itself is subcritical. See also: Ionization chamber Normally, several ranges of instrument sensitivities are required to cover the entire operating range. A range is required for low-level operation, beginning at the source level, whereas others are required for the intermediate- and high-power levels. Three ranges of detectors are common in

11 power reactor systems, and some systems contain a larger number. The total range to be covered is 7-10 decades (factors of 10) of power level. The chamber current of a neutron detector, suitably amplified, can be employed as a signal to operate automatic control system devices as well as to actuate reactor scram. In addition to power level, rate of change of power level is an important measurement which is recorded and employed to actuate various alarm and trip circuits. The normal range for the current ion chambers is approximately to 10-4 A. This current is suitably amplified in logarithmic and period amplifiers. The power monitors are part of a reactor's plant protection system. Applications 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 also: Nuclear fuel cycle; Nuclear fuels reprocessing; Radioactivity and radiation applications; Ship nuclear propulsion Frank J. Rahn How to cite this article Suggested citation for this article: Frank J. Rahn, "Nuclear reactor", in AccessScience@McGraw-Hill, DOI / , last modified: August 14, For Further Study Topic Page: Engineering & Materials: Other Engineering Disciplines:Nuclear engineering Bibliography J. Douglas, The nuclear option, EPRI Journal, December 1994 S. Glasstone and A. Sesonke, Nuclear Reactor Engineering, 4th ed., 1993 International Atomic Energy Association, Nuclear Power Reactors in the World, annually R. A. Knief, Nuclear Engineering: Theory and Technology of Commercial Nuclear Power, 1992 A. Nero, A Guidebook to Nuclear Reactors, 1980 F. J. Rahn et al., A Guide to Nuclear Power Technology, 1984, reprint 1992 S. C. Stulz and J. B, Kitto, Steam: Its Generation and Use, 40th ed., 1992 Customer Privacy Notice Copyright The McGraw-Hill Companies. All rights reserved. Any use is subject to the Terms of Use and Notice. Additional credits and copyright information. For further information about this site contact AccessScience@romnet.com. Last modified: Sep 30, 2003.