A pressurized water reactor (PWR) is a type of light-water nuclear reactor. PWRs constitute the large majority of the world's nuclear power plants (with notable exceptions being the UK, Japan and Canada). In a PWR, the primary coolant (water) is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms. The heated, high pressure water then flows to a steam generator, where it transfers its thermal energy to lower pressure water of a secondary system where steam is generated. The steam then drives turbines, which spin an electric generator. In contrast to a boiling water reactor (BWR), pressure in the primary coolant loop prevents the water from boiling within the reactor. All light-water reactors use ordinary water as both coolant and neutron moderator. Most use anywhere from 2 to 4 vertically mounted steam generators; VVER reactors use horizontal steam generators.
PWRs were originally designed to serve as nuclear marine propulsion for nuclear submarines and were used in the original design of the second commercial power plant at Shippingport Atomic Power Station.
PWRs currently operating in the United States are considered Generation II reactors. Russia's VVER reactors are similar to U.S. PWRs, but the VVER-1200 is not considered Generation II (see below). France operates many PWRs to generate the bulk of its electricity.
Several hundred PWRs are used for marine propulsion in aircraft carriers, nuclear submarines and ice breakers. In the US, they were originally designed at the Oak Ridge National Laboratory for use as a nuclear submarine power plant with a fully operational submarine power plant located at the Idaho National Laboratory. Follow-on work was conducted by Westinghouse Bettis Atomic Power Laboratory.The first purely commercial nuclear power plant at Shippingport Atomic Power Station was originally designed as a pressurized water reactor (although the first power plant connected to the grid was at Obninsk, USSR), on insistence from Admiral Hyman G. Rickover that a viable commercial plant would include none of the "crazy thermodynamic cycles that everyone else wants to build."
The United States Army Nuclear Power Program operated pressurized water reactors from 1954 to 1974.
Three Mile Island Nuclear Generating Station initially operated two pressurized water reactor plants, TMI-1 and TMI-2.The partial meltdown of TMI-2 in 1979 essentially ended the growth in new construction of nuclear power plants in the United States for two decades.
Watts Bar unit 2 (a Westinghouse 4-loop PWR) came online in 2016.
The pressurized water reactor has several new Generation III reactor evolutionary designs: the AP1000, VVER-1200, ACPR1000+, APR1400, Hualong One,IPWR-900 and EPR.
Nuclear fuel in the reactor pressure vessel is engaged in a fission chain reaction, which produces heat, heating the water in the primary coolant loop by thermal conduction through the fuel cladding. The hot primary coolant is pumped into a heat exchanger called the steam generator, where it flows through hundreds or thousands of small tubes. Heat is transferred through the walls of these tubes to the lower pressure secondary coolant located on the sheet side of the exchanger where the coolant evaporates to pressurized steam. The transfer of heat is accomplished without mixing the two fluids to prevent the secondary coolant from becoming radioactive. Some common steam generator arrangements are u-tubes or single pass heat exchangers.[ citation needed ]
In a nuclear power station, the pressurized steam is fed through a steam turbine which drives an electrical generator connected to the electric grid for transmission. After passing through the turbine the secondary coolant (water-steam mixture) is cooled down and condensed in a condenser. The condenser converts the steam to a liquid so that it can be pumped back into the steam generator, and maintains a vacuum at the turbine outlet so that the pressure drop across the turbine, and hence the energy extracted from the steam, is maximized. Before being fed into the steam generator, the condensed steam (referred to as feedwater) is sometimes preheated in order to minimize thermal shock.
The steam generated has other uses besides power generation. In nuclear ships and submarines, the steam is fed through a steam turbine connected to a set of speed reduction gears to a shaft used for propulsion. Direct mechanical action by expansion of the steam can be used for a steam-powered aircraft catapult or similar applications. District heating by the steam is used in some countries and direct heating is applied to internal plant applications.[ citation needed ]
Two things are characteristic for the pressurized water reactor (PWR) when compared with other reactor types: coolant loop separation from the steam system and pressure inside the primary coolant loop. In a PWR, there are two separate coolant loops (primary and secondary), which are both filled with demineralized/deionized water. A boiling water reactor, by contrast, has only one coolant loop, while more exotic designs such as breeder reactors use substances other than water for coolant and moderator (e.g. sodium in its liquid state as coolant or graphite as a moderator). The pressure in the primary coolant loop is typically 15–16 megapascals (150–160 bar ), which is notably higher than in other nuclear reactors, and nearly twice that of a boiling water reactor (BWR). As an effect of this, only localized boiling occurs and steam will recondense promptly in the bulk fluid. By contrast, in a boiling water reactor the primary coolant is designed to boil.
Light water is used as the primary coolant in a PWR. Water enters through the bottom of the reactor's core at about 548 K (275 °C; 527 °F) and is heated as it flows upwards through the reactor core to a temperature of about 588 K (315 °C; 599 °F). The water remains liquid despite the high temperature due to the high pressure in the primary coolant loop, usually around 155 bar (15.5 MPa 153 atm, 2,250 psi). In water, the critical point occurs at around 647 K (374 °C; 705 °F) and 22.064 MPa (3200 psi or 218 atm).
Pressure in the primary circuit is maintained by a pressurizer, a separate vessel that is connected to the primary circuit and partially filled with water which is heated to the saturation temperature (boiling point) for the desired pressure by submerged electrical heaters. To achieve a pressure of 155 bars (15.5 MPa), the pressurizer temperature is maintained at 345 °C (653 °F), which gives a subcooling margin (the difference between the pressurizer temperature and the highest temperature in the reactor core) of 30 °C (54 °F). As 345 °C is the boiling point of water at 155 bar, the liquid water is at the edge of a phase change. Thermal transients in the reactor coolant system result in large swings in pressurizer liquid/steam volume, and total pressurizer volume is designed around absorbing these transients without uncovering the heaters or emptying the pressurizer. Pressure transients in the primary coolant system manifest as temperature transients in the pressurizer and are controlled through the use of automatic heaters and water spray, which raise and lower pressurizer temperature, respectively.
The coolant is pumped around the primary circuit by powerful pumps. MPa (60 atm, 900 psia), 275 °C (530 °F) — for use in the steam turbine. The cooled primary coolant is then returned to the reactor vessel to be heated again.These pumps have a rate of ~100,000 gallons of coolant per minute. After picking up heat as it passes through the reactor core, the primary coolant transfers heat in a steam generator to water in a lower pressure secondary circuit, evaporating the secondary coolant to saturated steam — in most designs 6.2
Pressurized water reactors, like all thermal reactor designs, require the fast fission neutrons to be slowed (a process called moderation or thermalizing) in order to interact with the nuclear fuel and sustain the chain reaction. In PWRs the coolant water is used as a moderator by letting the neutrons undergo multiple collisions with light hydrogen atoms in the water, losing speed in the process. This "moderating" of neutrons will happen more often when the water is more dense (more collisions will occur). The use of water as a moderator is an important safety feature of PWRs, as an increase in temperature may cause the water to expand, giving greater 'gaps' between the water molecules and reducing the probability of thermalization — thereby reducing the extent to which neutrons are slowed and hence reducing the reactivity in the reactor. Therefore, if reactivity increases beyond normal, the reduced moderation of neutrons will cause the chain reaction to slow down, producing less heat. This property, known as the negative temperature coefficient of reactivity, makes PWR reactors very stable. This process is referred to as 'Self-Regulating', i.e. the hotter the coolant becomes, the less reactive the plant becomes, shutting itself down slightly to compensate and vice versa. Thus the plant controls itself around a given temperature set by the position of the control rods.
In contrast, the RBMK reactor design used at Chernobyl, which uses graphite instead of water as the moderator and uses boiling water as the coolant, has a large positive thermal coefficient of reactivity that increases heat generation when coolant water temperatures increase. This makes the RBMK design less stable than pressurized water reactors. In addition to its property of slowing down neutrons when serving as a moderator, water also has a property of absorbing neutrons, albeit to a lesser degree. When the coolant water temperature increases, the boiling increases, which creates voids. Thus there is less water to absorb thermal neutrons that have already been slowed by the graphite moderator, causing an increase in reactivity. This property is called the void coefficient of reactivity, and in an RBMK reactor like Chernobyl, the void coefficient is positive, and fairly large, causing rapid transients. This design characteristic of the RBMK reactor is generally seen as one of several causes of the Chernobyl disaster.
Heavy water has very low neutron absorption, so heavy water reactors tend to have a positive void coefficient, though the CANDU reactor design mitigates this issue by using unenriched, natural uranium; these reactors are also designed with a number of passive safety systems not found in the original RBMK design.
PWRs are designed to be maintained in an undermoderated state, meaning that there is room for increased water volume or density to further increase moderation, because if moderation were near saturation, then a reduction in density of the moderator/coolant could reduce neutron absorption significantly while reducing moderation only slightly, making the void coefficient positive. Also, light water is actually a somewhat stronger moderator of neutrons than heavy water, though heavy water's neutron absorption is much lower. Because of these two facts, light water reactors have a relatively small moderator volume and therefore have compact cores. One next generation design, the supercritical water reactor, is even less moderated. A less moderated neutron energy spectrum does worsen the capture/fission ratio for 235U and especially 239Pu, meaning that more fissile nuclei fail to fission on neutron absorption and instead capture the neutron to become a heavier nonfissile isotope, wasting one or more neutrons and increasing accumulation of heavy transuranic actinides, some of which have long half-lives.
After enrichment, the uranium dioxide (UO
2) powder is fired in a high-temperature, sintering furnace to create hard, ceramic pellets of enriched uranium dioxide. The cylindrical pellets are then clad in a corrosion-resistant zirconium metal alloy Zircaloy which are backfilled with helium to aid heat conduction and detect leakages. Zircaloy is chosen because of its mechanical properties and its low absorption cross section. The finished fuel rods are grouped in fuel assemblies, called fuel bundles, that are then used to build the core of the reactor. A typical PWR has fuel assemblies of 200 to 300 rods each, and a large reactor would have about 150–250 such assemblies with 80–100 tons of uranium in all. Generally, the fuel bundles consist of fuel rods bundled 14 × 14 to 17 × 17. A PWR produces on the order of 900 to 1,600 MWe. PWR fuel bundles are about 4 meters in length.
Refuelings for most commercial PWRs is on an 18–24 month cycle. Approximately one third of the core is replaced each refueling, though some more modern refueling schemes may reduce refuel time to a few days and allow refueling to occur on a shorter periodicity.
In PWRs reactor power can be viewed as following steam (turbine) demand due to the reactivity feedback of the temperature change caused by increased or decreased steam flow. (See: Negative temperature coefficient.) Boron and cadmium control rods are used to maintain primary system temperature at the desired point. In order to decrease power, the operator throttles shut turbine inlet valves. This would result in less steam being drawn from the steam generators. This results in the primary loop increasing in temperature. The higher temperature causes the density of the primary reactor coolant water to decrease, allowing higher neutron speeds, thus less fission and decreased power output. This decrease of power will eventually result in primary system temperature returning to its previous steady-state value. The operator can control the steady state operating temperature by addition of boric acid and/or movement of control rods.
Reactivity adjustment to maintain 100% power as the fuel is burned up in most commercial PWRs is normally achieved by varying the concentration of boric acid dissolved in the primary reactor coolant. Boron readily absorbs neutrons and increasing or decreasing its concentration in the reactor coolant will therefore affect the neutron activity correspondingly. An entire control system involving high pressure pumps (usually called the charging and letdown system) is required to remove water from the high pressure primary loop and re-inject the water back in with differing concentrations of boric acid. The reactor control rods, inserted through the reactor vessel head directly into the fuel bundles, are moved for the following reasons: to start up the reactor, to shut down the primary nuclear reactions in the reactor, to accommodate short term transients, such as changes to load on the turbine,
The control rods can also be used to compensate for nuclear poison inventory and to compensate for nuclear fuel depletion. However, these effects are more usually accommodated by altering the primary coolant boric acid concentration.
In contrast, BWRs have no boron in the reactor coolant and control the reactor power by adjusting the reactor coolant flow rate.
PWR reactors are very stable due to their tendency to produce less power as temperatures increase; this makes the reactor easier to operate from a stability standpoint.
PWR turbine cycle loop is separate from the primary loop, so the water in the secondary loop is not contaminated by radioactive materials.
PWRs can passively scram the reactor in the event that offsite power is lost to immediately stop the primary nuclear reaction. The control rods are held by electromagnets and fall by gravity when current is lost; full insertion safely shuts down the primary nuclear reaction.
PWR technology is favoured by nations seeking to develop a nuclear navy; the compact reactors fit well in nuclear submarines and other nuclear ships.
The coolant water must be highly pressurized to remain liquid at high temperatures. This requires high strength piping and a heavy pressure vessel and hence increases construction costs. The higher pressure can increase the consequences of a loss-of-coolant accident.The reactor pressure vessel is manufactured from ductile steel but, as the plant is operated, neutron flux from the reactor causes this steel to become less ductile. Eventually the ductility of the steel will reach limits determined by the applicable boiler and pressure vessel standards, and the pressure vessel must be repaired or replaced. This might not be practical or economic, and so determines the life of the plant.
Additional high pressure components such as reactor coolant pumps, pressurizer, steam generators, etc. are also needed. This also increases the capital cost and complexity of a PWR power plant.
The high temperature water coolant with boric acid dissolved in it is corrosive to carbon steel (but not stainless steel); this can cause radioactive corrosion products to circulate in the primary coolant loop. This not only limits the lifetime of the reactor, but the systems that filter out the corrosion products and adjust the boric acid concentration add significantly to the overall cost of the reactor and to radiation exposure. In one instance, this has resulted in severe corrosion to control rod drive mechanisms when the boric acid solution leaked through the seal between the mechanism itself and the primary system.
Due to the requirement to load a pressurized water reactor's primary coolant loop with boron, undesirable radioactive secondary tritium production in the water is over 25 times greater than in boiling water reactors of similar power, owing to the latter's absence of the neutron moderating element in its coolant loop. The tritium is created by the absorption of a fast neutron in the nucleus of a boron-10 atom which subsequently splits into a lithium-7 and tritium atom. Pressurized water reactors annually emit several hundred curies of tritium to the environment as part of normal operation. https://www.nrc.gov/reactors/operating/ops-experience/tritium/faqs.html
Natural uranium is only 0.7% uranium-235, the isotope necessary for thermal reactors. This makes it necessary to enrich the uranium fuel, which significantly increases the costs of fuel production.
Because water acts as a neutron moderator, it is not possible to build a fast-neutron reactor with a PWR design. A reduced moderation water reactor may however achieve a breeding ratio greater than unity, though this reactor design has disadvantages of its own.
In 1954 the world's first nuclear powered electricity generator began operation in the then closed city of Obninsk at the Institute of Physics and Power Engineering (FEI or IPPE).
A nuclear reactor, formerly known as an atomic pile, is a device used to initiate and control a fission nuclear chain reaction or nuclear fusion reactions. Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nuclear fission is passed to a working fluid, which in turn runs through steam turbines. These either drive a ship's propellers or turn electrical generators' shafts. Nuclear generated steam in principle can be used for industrial process heat or for district heating. Some reactors are used to produce isotopes for medical and industrial use, or for production of weapons-grade plutonium. As of early 2019, the IAEA reports there are 454 nuclear power reactors and 226 nuclear research reactors in operation around the world.
A boiling water reactor (BWR) is a type of light water nuclear reactor used for the generation of electrical power. It is the second most common type of electricity-generating nuclear reactor after the pressurized water reactor (PWR), which is also a type of light water nuclear reactor. The main difference between a BWR and PWR is that in a BWR, the reactor core heats water, which turns to steam and then drives a steam turbine. In a PWR, the reactor core heats water, which does not boil. This hot water then exchanges heat with a lower pressure water system, which turns to steam and drives the turbine. The BWR was developed by the Argonne National Laboratory and General Electric (GE) in the mid-1950s. The main present manufacturer is GE Hitachi Nuclear Energy, which specializes in the design and construction of this type of reactor.
A nuclear meltdown is a severe nuclear reactor accident that results in core damage from overheating. The term nuclear meltdown is not officially defined by the International Atomic Energy Agency or by the United States Nuclear Regulatory Commission. It has been defined to mean the accidental melting of the core of a nuclear reactor, however, and is in common usage a reference to the core's either complete or partial collapse.
The RBMK is a class of graphite-moderated nuclear power reactor designed and built by the Soviet Union. The name refers to its unusual design where, instead of a large steel pressure vessel surrounding the entire core, the core is surrounded by a cylindrical annular steel tank inside a concrete vault and each fuel assembly is enclosed in an individual 8 cm diameter pipe surrounded in graphite which allows the flow of cooling water around the fuel.
The A2W reactor is a naval nuclear reactor used by the United States Navy to provide electricity generation and propulsion on warships. The A2W designation stands for:
A loss-of-coolant accident (LOCA) is a mode of failure for a nuclear reactor; if not managed effectively, the results of a LOCA could result in reactor core damage. Each nuclear plant's emergency core cooling system (ECCS) exists specifically to deal with a LOCA.
In nuclear engineering, the void coefficient is a number that can be used to estimate how much the reactivity of a nuclear reactor changes as voids form in the reactor moderator or coolant. Net reactivity in a reactor is the sum total of all these contributions, of which the void coefficient is but one. Reactors in which either the moderator or the coolant is a liquid typically will have a void coefficient value that is either negative or positive. Reactors in which neither the moderator nor the coolant is a liquid will have a void coefficient value equal to zero. It is unclear how the definition of 'void' coefficient applies to reactors in which the moderator/coolant is neither liquid nor gas.
The light-water reactor (LWR) is a type of thermal-neutron reactor that uses normal water, as opposed to heavy water, as both its coolant and neutron moderator – furthermore a solid form of fissile elements is used as fuel. Thermal-neutron reactors are the most common type of nuclear reactor, and light-water reactors are the most common type of thermal-neutron reactor.
Passive nuclear safety is a design approach for safety features, implemented in a nuclear reactor, that does not require any active intervention on the part of the operator or electrical/electronic feedback in order to bring the reactor to a safe shutdown state, in the event of a particular type of emergency. Such design features tend to rely on the engineering of components such that their predicted behaviour would slow down, rather than accelerate the deterioration of the reactor state; they typically take advantage of natural forces or phenomena such as gravity, buoyancy, pressure differences, conduction or natural heat convection to accomplish safety functions without requiring an active power source. Many older common reactor designs use passive safety systems to a limited extent, rather, relying on active safety systems such as diesel powered motors. Some newer reactor designs feature more passive systems; the motivation being that they are highly reliable and reduce the cost associated with the installation and maintenance of systems that would otherwise require multiple trains of equipment and redundant safety class power supplies in order the achieve the same level of reliability. However, weak driving forces that power many passive safety features can pose significant challenges to effectiveness of a passive system, particularly in the short term following an accident.
The supercritical water reactor (SCWR) is a concept Generation IV reactor, mostly designed as light water reactor (LWR) that operates at supercritical pressure. The term critical in this context refers to the critical point of water, and must not be confused with the concept of criticality of the nuclear reactor.
Steam generators are heat exchangers used to convert water into steam from heat produced in a nuclear reactor core. They are used in pressurized water reactors (PWR) between the primary and secondary coolant loops.
The water-water energetic reactor (WWER), or VVER is a series of pressurized water reactor designs originally developed in the Soviet Union, and now Russia, by OKB Gidropress. The idea of such a reactor was proposed at the Kurchatov Institute by Savely Moiseevich Feinberg. VVER were originally developed before the 1970s, and have been continually updated. As a result, the name VVER is associated with a wide variety of reactor designs spanning from generation I reactors to modern generation III+ reactor designs. Power output ranges from 70 to 1300 MWe, with designs of up to 1700 MWe in development. The first prototype VVER-210 was built at the Novovoronezh Nuclear Power Plant.
Carolinas–Virginia Tube Reactor (CVTR), also known as Parr Nuclear Station, was an experimental pressurized tube heavy water nuclear power reactor at Parr, South Carolina in Fairfield County. It was built and operated by the Carolinas Virginia Nuclear Power Associates. CVTR was a small test reactor, capable of generating 17 megawatts of electricity. It was officially commissioned in December 1963 and left service in January 1967.
International Reactor Innovative and Secure (IRIS) is a Generation IV reactor design made by an international team of companies, laboratories, and universities and coordinated by Westinghouse. IRIS is hoped to open up new markets for nuclear power and make a bridge from Generation III reactor to Generation IV reactor technology. The design is not yet specific to reactor power output. Notably, a 335 MW output has been proposed, but it could be tweaked to be as low as a 100 MW unit.
The three primary objectives of nuclear reactor safety systems as defined by the U.S. Nuclear Regulatory Commission are to shut down the reactor, maintain it in a shutdown condition and prevent the release of radioactive material.
A nuclear reactor coolant is a coolant in a nuclear reactor used to remove heat from the nuclear reactor core and transfer it to electrical generators and the environment. Frequently, a chain of two coolant loops are used because the primary coolant loop takes on short-term radioactivity from the reactor.
Boiling water reactor safety systems are nuclear safety systems constructed within boiling water reactors in order to prevent or mitigate environmental and health hazards in the event of accident or natural disaster.
General Electric's BWR product line of Boiling Water Reactors represents the designs of a relatively large percentage of the commercial fission reactors around the world.
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An organic nuclear reactor, or organic cooled reactor (OCR), is a type of nuclear reactor that uses some form of organic fluid, typically a hydrocarbon substance like polychlorinated biphenyl (PCB), for cooling and sometimes as a neutron moderator as well.
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