nuclear reactor246206737 reactors-basics

nuclear reactor246206737 reactors-basics

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  • Fuel Cycle •How a Reactor Works • High Temperature Gas Cooled Reactors • Boiling Water Reactors • Pressurized Water Reactors • Fast Breeder Reactor Fuel Cycle The steps required to transform raw uranium ore into nuclear fuel are known collectively as the fuel cycle. Uranium is a relatively abundant element in the earths crust. There are several types of uranium ore, uranite and pitchblend are two forms of the same mineral, both are uranium dioxide. Gummite is another common ore containing uranium oxide and water. Orbernite is a rarer ore of uranium that usually does not occur in massive enough quantities to make mining it economical but in Zaire it forms large masses. It is a hydrous copper uranium phosphate. Aytunite is hydrous calcium uranium phosphate and is very abundant worldwide. Carnotite is hydrous potassium uranium vanadate found in sandstone deposits in the Southwestern United States and elsewhere. Uranophane is a somewhat less common mineral found in several areas of the US, Saxony, Czechoslovakia and Zaire. It is hydrated calcium uranium silicate. According to Washington Nuclear Corporation “The percentage of U in uranium ores varies widely: Namibian ore is about 0.035 percent U3O8, US ore ranges from 0.05 to 0.3 percent, and in Northern Saskatchewan the U3O8 reaches 12.0 percent. Major deposits are also mined in Australia and the former Soviet Union. Uranium is an abundant but not concentrated material. Material mined in the various producing areas in 1997 amounted to 35,692 metric tons, of that amount Canada produced 12,029 metric tons and the United States produced 2,170 metric tons. Canada produced 35% of the world’s production. Uranium may be mined in several ways depending upon where the ore is located and how concentrated it is. Open pit mining is used for deposits near the surface and underground mining is used for deeper high- grade ore. In locations where the uranium ore is in sandstone below the water table of a confined aquifer it may be mined by the in situ leach method by injecting a mildly alkaline solution in to the rock to dissolve the uranium and then pumping it out. The next step is milling, except for uranium mined by the insitu leach method which does not require milling. The ore is brought out of the mine and ground finely, then a dilute solution of acid or alkali is added to the powder to leach out the uranium. The resulting fluid is treated to extract and concentrate the uranium. The liquor, as it’s called, is dried to produce a product known as yellowcake. The yellowcake is 70 - 90 % Uranium oxide, U3O8. During conversion the yellowcake is chemically converted to Uranium hexafluoride or Hex, UF6. This product is a white solid which transforms into a gas at 56 C. There is only one Uranium Hexafluoride Production Facility in the United States, Allied - Signal Inc. at Metropolis, Illinois. Hex is highly toxic and corrosive so it must be handled carefully. It is approximately 99.28% Uranium238 and 0.71% Uranium235 this is the same as natural ore. U238 does not fission, therefore, the percentage of U235 must be increased in order to sustain a chain reaction. In the United
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  States enrichment isdone at two Gaseous Diffusion Plants, Paducah, Kentucky and Portsmouth, Ohio. Both plants are operated by United States Enrichment Corporation, a formerly government owned company which is now operated by a contractor. This process is begun by heating the UF6 slightly to produce a gas which is them passed through a series of diffusion barriers to increase the percentage of U235 to 3-5%. The gas must pass through hundreds of semi permeable membranes with holes less than one millionth of a centimeter in diameter. The membranes are in a series of pipes. The U235 is lighter and passes through the barriers first. The U238 is heavier and passes more slowly through the barriers so it can be captured and drawn off. Many sets of pipes are arranged in sets. Each set is known as a cascade. The process uses a large amount of electricity and is not terribly efficient. The enriched UF6 gas is cooled after the final cascade. The final product is solid UF6 with is then converted into UO2 or Uranium Oxide. This power is compacted into pellets which are sintered (heated) to produce ceramic UO2 pellets. These are loaded into tubes made from a Zirconium and Tin alloy called Zircaloy or occasionally Stainless Steel. The air is removed and replaced with Helium. The tubes are then sealed at each end with weld plugs. Sealing is intended to contain fission products produced while in operation within the fuel rods. The tubes are loaded into racks called assemblies and the finished assemblies are loaded into the reactor. The individual tubes that make up the assembly are called rods. The number of fuel rods, their size and the arrangement of the fuel assemblies varies according to the type of reactor. There are seven commercial fuel fabrication plants in the United States, they are: ABB Combustion Engineering in Hematite, Missouri; General Electric in Wilmington, North Carolina; Westinghouse Electric in Columbia, South Carolina; Nuclear Fuel Services in Erwin, Tennessee; Babcock and Wilcox Fuel Company in Lynchburg, Virginia; and Siemens Nuclear Power Corporation in Richland, Washington. The average Boiling Water Reactor assembly weighs 319.9kg, of which 183.3kg is Uranium, measures 4.470m in length, and contains 63 fuel rods. The average Pressurized Water Reactor assembly weighs 657.9kg, of which 461.4kg is Uranium, measures 4.059m in length, and contains 264 fuel rods. The fuel assemblies are loaded into the reactor using a gantry crane while the reactor is shut down. There may be 750 assemblies in a BWR, 150 in a PWR. Of the 110 reactors operating in the United States, 73 are PWRs and 35 are BWRs. While the reactor is operating the fuel reaches quite high temperatures. It tends to heat the Uranium pellets unevenly. On the surface the pellets reach approximately 1200o F, while the core reaches 2300o F. Both figures are substantially lower than the melting point of UO2 which is 5072o F but the differential heating tends to crack the pellets. The fuel assemblies are rotated through the reactor, starting when first loaded on the outside and being rotated toward the center later. The average lifespan of a fuel assembly is 4 to 6 years. Since the reactor must be shutdown to load and unload fuel, utilities are changing from an annual cycle which requires 3 refuelings in a 3 year period, to an 18 month cycle which requires only 2 refuelings in the same period. The fuel removed from the reactor is placed in a storage pool next to the reactor. Currently there is no reprocessing of used commercial fuel rods in the United States, although a number of European countries reprocess fuel from commercial plants. Back to menu How a Reactor Works A nuclear reactor consists of a containment vessel which surrounds the reactor vessel, a number of Uranium fuel assemblies inside the reactor vessel, a loop of pipe that carries water
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  from the reactorto a steam generator and back to the reactor by means of a pump, another loop of pipe to take steam from the steam generator to the turbine generator and then take water back to the steam generator to be made into steam again by means of a pump. The key to the process is the heat generated in the reactor by the fissioning of Uranium235. The reactor is started by slowly withdrawing the control rods from the core to get the nuclear chain reaction started. The fuel begins to fission, each atom of U235 that is struck by a free neutron in turn produces free neutrons, which strike an atom of U235 to continue the chain reaction. The water in the reactor acts as a moderator to slow the neutrons and make it more likely that they will cause fissioning. The control rods can be moved in or out of the reactor to slow down or speed up the fission reaction. The control rods contain material that absorbs neutrons, such as cadmium or boron. When enough neutrons are absorbed, the reaction stops. In addition to moderation the reaction the water acts as coolant to control the temperature of the core and prevent the fuel from melting. The system operates under pressure, something like a kitchen pressure cooker. This allows the water to reach much higher temperatures, nearly 300°F, than it otherwise could without boiling. When this superheated water reaches the steam generator the cool water in the secondary loop is immediately brought to a boil and converted into steam to turn the blades of the turbine and generate electricity. This is the same principle that is used in plants that burn oil, coal or gas, the heat is used to boil water and turn a turbine, the only difference is the source of the heat. There are a number of different reactor designs in use, but most of the reactors in the United States are either Boiling Water (BWR) or Pressurized Water (PWR) reactors. Worldwide there are several other types including Gas Cooled, Pressurized Heavy Water, Graphite and Water, Fast Breeders, Light Water Breeders, Magnox Gas Cooled. Boiling Heavy Water, Light Water Cooled Heavy Water, and others. The various configurations and moderator/coolant combinations all operate on the basic principal of heat produced by a nuclear chain reaction being used to turn a turbine and generate power. There have been individual examples of several of these other types of reactors built in the US. Some of these have had operating problems or accidents and most are now closed. Back to menu High Temperature Gas Cooled Reactors Fort St. Vrain and Peach Bottom were high temperature gas cooled reactors, both are now decommissioned . Peach Bottom was an experimental 40 megawatt reactor which went on line in January of 1966 in Lancaster County Pennsylvania and closed in October of 1974. It was operated by Peco Energy Company of Philadelphia. The Fort St. Vrain reactor was a 330 megawatt reactor ordered in 1965 from General Atomic Co., which went on line in December of 1973 in Platteville Colorado and closed in August of 1989. It was operated by Public Service Company of Colorado. High temperature gas cooled reactors such as the Fort St. Vrain plant are graphite moderated and cooled with helium. The helium coolant permits higher operating temperatures of the fluid used, 760°–1430°F The steam generators of the plant produce steam under pressure, 2400 psia and 1000°F Fort St. Vrain used a different type of fuel, "microspheres of thorium and fully enriched uranium embedded in a carbonaceous binder" The Thorium232 is transformed into Uranium233 in the core by neutron bombardment and the U233 is fissionable, therefore the reactor could run longer before refueling. The plant shut down prematurely due to "failed control rod drives and degradation of [the] steam generator ring header". Back to menu
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  Boiling Water Reactors BoilingWater Reactors are the second most common type in the United States, there are 35 operational BWRs at this time. The design of the BWR has two variants, Dual Cycle and Single Cycle. In a dual cycle BWR there are two loops for the coolant, in a single cycle there is only one. The vast majority of operating reactors are single cycle. The primary difference between the two is that with a single cycle, radiation precautions must be taken around the turbines since the water used to generate the steam used to turn the turbines has passed through the core of the reactor. In a single cycle BWR steam is generated directly in the core of the reactor at 1000 psia. A steam separator at the top of the reactor vessel directs the steam to the turbines and the water is recirculated through the core by recirculation pumps. "Orifices at the base of each fuel assembly control the coolant flow to the individual fuel assemblies". The water serves as coolant and moderator. The dual cycle system also produces steam in the core, but rather than being channeled to the turbine it goes to a steam generator where its heat is used to produce steam in the secondary loop which is then used to turn the turbines. All BWRs in the US have been designed and manufactured by General Electric. The predominant single cycle type is favored because it "allows a lower capital investment for piping, heat exchangers, pumps etc. It also has a somewhat better thermal efficiency because of the direct generation of steam at the maximum cycle temperature". The BWR also has comparatively low operating costs. Two problems have been identified with the BWR design used in the US, the General Electric designed Mark 1 containment was deemed faulty and a modification was added to prevent breech of containment in the event of a buildup of pressure within it. A manually operated vent was added with a 300 foot stack which can be activated by operators to release pressure from the containment directly to the environment. This prevents the complete loss of containment by intentionally exposing the public to potentially harmful radiation. The other difficulty is with cracking of the welds in the core shroud which surrounds the fuel core in the reactor. This is made of circumferentially welded stainless steel plates. It has been determined that the rate of failure in these welds has been abnormally high. When they fail the fuel bundles are allowed to shift which can affect the ability of the control rods to operate normally. Back to menu Pressurized Water Reactors Pressurized Water Reactors are the most common type of reactor used in the United States, there are 73 operational PWRs at this time. Over half of these were designed and produced by Westinghouse, the rest by either Combustion Engineering or Babcock and Wilcox. The fundamental principals governing the PWR are similar to the dual cycle BWR with the exception that the high pressure, 2250psi, in the primary loop prevents steam formation despite temperatures of 600°F Steam is produced in the secondary loop within the steam generators. Steam pressure in the secondary loop is lower, 1000psi. The steam is directed to the turbines then collects in the condenser where it becomes water again and is pumped back to the steam generator by the feed pump. The water in the primary loop acts as both coolant and moderator. The dual cycle system means that only the water in the primary loop is radioactively contaminated, the secondary loop as well as the turbines stay clean so radiation
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  precautions are notrequired near the turbines. The initial capital investment in a PWR is higher than a BWR but operating costs are lower. Workers are exposed to less radiation due to the duel loop design. A common problem associated with PWRs , leaks in steam generator tubes can be caused by corrosion and high pressure. This cracking is a safety issue since a multiple tube rupture "could result in a rapid loss of coolant accident in the reactor beyond the ability of the Emergency Core Cooling System to control". Trojan Nuclear Power Plant in Rainier, Oregon closed prematurely due to repeated leaks in steam generator tubes. An additional problem is reactor pressure vessel embrittlement caused by neutron bombardment. The metal of the reactor vessel undergoes changes as a result of the radioactivity which cause it to lose ductility, the metals ability to expand and contract in response to variations in temperature and pressure is lost. It is possible for an embrittled pressure vessel to lose integrity in the event of the activation of the Emergency Core Cooling System, like pouring cold water into a hot glass dish. Notices to plant operators on both of these issues have been sent by the NRC. Back to menu Fast Breeder Reactor Natural uranium consists of 99.3% Uranium-238 and 0.7% Uranium-235. Of these two forms only Uranium-235 can be used as a nuclear fuel. In a conventional thermal reactor, during operation some Uranium-238 is transformed into Plutonium-239 which can also be used as a nuclear fuel. By recycling this plutonium to make new fuel it may be possible to exploit at most about 2% of the potential fuel value of the world's natural uranium resources. However, a fast breeder reactor can convert Uranium-238 into Plutonium-239 at a rate faster than it consumes its fuel. By repeated recycling of the fuel it should be realistically possible to exploit 50% of the fuel value of the uranium feed. This means that fast reactors could extend the energy output from the world's uranium fuel reserves 25 fold. About the term "Fast" A neutron released from the fission of Uranium-235 (or Plutonium-239) has a high energy. To increase the probability of the neutron causing the fission of another nucleus of the fuel material - and thereby continuing the chain reaction - either its energy must be reduced, or the concentration of fissionable target nuclei must be increased. The approach of reducing the neutron energy led to the development of the nuclear reactors which are now widely used for power production. In these reactors the energy of the neutron is reduced by "bouncing" it off the atoms in a so-called moderator material until the neutron is in thermal equilibrium with the atoms with which it is interacting. The neutron is then termed a "thermal neutron" and reactors using this principle are called "thermal reactors". The alternative approach of increasing the concentration of fissionable material and using fission caused by high energy or "fast" neutrons led to the development of "fast" reactors. About the term "Breeder" If a neutron is captured by a Uranium-238 nucleus the following reaction takes place:
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  The result isthat Uranium-238, which is very difficult to fission, is transformed into Plutonium- 239 which can be fissioned much more easily. This means that a useful reactor fuel can be made from an otherwise useless natural resource. The symbol +n represents the gain of a neutron by capture and β-  represents radioactive decay by beta emission with the half-life shown below the arrow. As mentioned above, for the chain reaction to continue, on average one neutron released during fission must go on to cause the fission of another nucleus. The average number of neutrons released by fission depends on several factors but is usually around 2.5. Some of these neutrons are inevitably lost as they are captured by the reactor structure or coolant but if, on average, more than one of the available 1.5 neutrons is captured by Uranium-238 then the total number of atoms in the reactor that can be fissioned will increase as the reactor operates. This is the principle of the "breeder" reactor. All reactors contain Uranium-238 in their core and so the reaction which produces Plutonium- 239 occurs to some extent in all reactors. However, only in a reactor using fast neutrons for fission is it possible to "breed" more new fuel than the reactor consumes; this is what a Fast Breeder Reactor does. The amount of breeding that takes place in a fast reactor depends on the size and design of the reactor core and the concentration of the fissile material that is in it. Some fast reactors may be deliberately designed not to breed at all. It is important to realise that not all fast reactors are breeder reactors. Breeding and burning In the early days of research in fast reactor technology it was considered very important to maximise the amount of plutonium that was bred in order to fuel an increasing number of reactors that would follow. Two important changes have taken place since then which have changed this ... Around 1960 it became clear that the economics of operating a fast reactor depend more on minimising the number of times that its fuel must be recycled than on simply maximising the breeding. Fuel assemblies cannot stay in the fast reactor core indefinitely for a number of reasons, principally because the fissile material is gradually consumed. Breeding does take place in the fuel but because it also takes place in the blanket there is a net loss of fuel in the core centre and a net gain around the edge where it cannot contribute to powering the reactor. In addition the steel fuel pins which hold the fuel are weakened by the gradual swelling of the fuel pellets, chemical attack by the fuel and damage by fast neutrons. So from the early 1960's until the end of the 1980's the emphasis was on maximising the amount of energy which could be extracted from a fuel assembly before it had to be removed for reprocessing - the energy extracted is usually referred to as the "burn-up" of the fuel. The results of the years of development carried out in this field were impressive; a twenty-fold increase in the burn-up was achieved by using oxide fuel in specially developed alloy fuel pins.
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  Since the early1990's the picture has again changed. The widespread use of thermal reactors means that there is now an ample stockpile of plutonium. Plutonium from dismantled nuclear weapons may also become available with the end of the Cold War. A fast reactor which is configured to consume plutonium is called a "burner" reactor. The fast neutrons in the core of a fast reactor have another potential use: for a long time the disposal of a certain type of nuclear waste produced in thermal reactors which belong to the category of elements called actinides has been a problem. By putting this nuclear waste into the fast reactor it can be broken down into materials which are more easily disposed of. The plutonium burning and actinide disposal project is led by France with the participation of the United Kingdom and Japan. The fast reactor is very versatile, a reactor can be designed so that by changing the core it can be used to create more plutonium than it destroys or destroy more plutonium than it creates according to what is required. The Fuel: As explained above, fuel cannot stay in the reactor indefinitely. To recover the useful materials in the fuel it must be sent to a reprocessing plant for chemical separation. The recovered unburnt and bred plutonium is then returned to the fuel fabrication plant to be put into new fuel rods. Fast Reactor Design The coolant The coolant which passes through the core of the fast reactor must not moderate (slow down) the neutrons emitted from the fission reaction. This limits the choice of coolant slightly, in particular water is unsuitable as a coolant because it is a very effective moderator. To avoid moderating the neutrons suitable coolants are liquid metals and inert gases (e.g. helium). Little research has been done on inert gases in fast reactor applications, liquid metals are the preferred option due to their excellent heat transfer properties. The liquid metal coolant would preferably remain in the liquid state below or as close as possible to ambient temperatures (to avoid expensive pre-heating of coolant filled vessels and pipes), have a boiling point which gives a margin of safety above the proposed reactor operating temperatures and be obtainable in sufficiently large quantities at a reasonable cost. Possible choices of liquid metal coolant would include mercury, lead, sodium and a sodium- potassium mixture; except for lead, all these have indeed been used as fast reactor coolants and designs with lead coolant have been made. However, mercury and lead have problems of chemical toxicity which make them unattractive and mercury is also very expensive. The sodium-potassium mixture - often referred to as NaK - has the advantage of remaining liquid at room temperature, obviating the need for pre-heating, but in other respects is more difficult to handle than sodium. Fast reactor engineers around the world have therefore arrived upon the same conclusion that the best coolant is sodium.
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  The reactor layout Oneof the most important choices facing the designer of a fast reactor is the layout to be adopted for the primary circuit which includes the reactor. There are two main concepts, the "pool" and the "loop". In a pool-type reactor one large vessel holds the core, the Intermediate Heat Exchanger (IHX) which passes heat to the secondary loops, and the pump which circulates the primary sodium. In a loop-type reactor, the core, IHX and pump are each in their own smaller vessels linked by pipes. The layouts are illustrated below, sodium flows upwards through the core (shown yellow) in both designs: The choice is not simple since each has its own advantages and disadvantages. The vessel of the pool is a very simple design with no branches to cause stress concentrations. It can be arranged so that hot coolant never comes into contact with the vessel wall. The disadvantages of the pool are that the vessel is so large that it must be fabricated on-site where quality assurance is more difficult. Once in operation its internal structures are difficult to inspect. The reactor vessel of the loop-type, being much smaller, can be built in a factory and transported to the site. The pipework of the loop reactor may be longer and more complicated but it is easier to inspect. Both types of reactor have been built but, to date, there is less experience with large scale loop-type reactors. MONJU which is a loop-type reactor will provide an interesting comparison with the European pool-type prototypes. Back to menu