Pressurized water reactors ( PWRs ) are the majority of the world's nuclear power plants (notable exceptions are the UK, Japan and Canada) and are one of three types of water reactors lightweight (LWR), other types are boiling water reactors (BWRs) and supercritical water reactors (SCWRs). In PWR, the main cooler (water) is pumped under high pressure to the reactor core where it is heated by energy released by atomic fission. The heated water then flows to a steam generator where it transfers its heat energy to a secondary system where steam is generated and flows into a turbine which, in turn, turns the electric generator. Unlike the boiling water reactor, the pressure in the primary cooling ring prevents the boiling water inside the reactor. All LWRs use ordinary water as cooling and neutron moderators.
PWR was originally designed to serve as a nuclear marine drive for nuclear submarines and used in the original design of a second commercial power station in Shippingport Atomic Power Station.
The PWR currently operating in the United States is considered a Generation II reactor. The Russian VVER reactor is similar to the US PWRs. France operates many PWRs to generate most of its electricity.
Video Pressurized water reactor
History
Several hundred PWRs are used for marine propulsion on aircraft carriers, nuclear submarines and ice breakers. In the US, they were originally designed at Oak Ridge National Laboratory for use as an underwater nuclear power plant. Follow-up work is undertaken by Westinghouse Bettis Atomic Power Laboratory. The first commercially pure commercial nuclear power station at Shippingport Atomic Power Station was originally designed as a pressurized water reactor (though the first power plant connected to the network was in Calder Hall, England), at the insistence of Admiral Hyman G. Rickover that a commercially viable plant would not including "the crazy thermodynamic cycle that everyone wants to build."
The United States Army Nuclear Power Program operated a pressurized water reactor from 1954 to 1974.
Three Gener Island Nuclear Generating Station initially operates two pressurized water reactors, TMI-1 and TMI-2. The partial decay of TMI-2 in 1979 essentially ended the growth of new construction of nuclear power plants in the United States for two decades.
Pressurized water reactors have three new generation III reactor evolution designs: AP-1000, VVER-1200, ACPR1000
Maps Pressurized water reactor
Design
Nuclear fuel in the reactor pressure vessel is involved in the fission chain reaction, which generates heat, heats water in the primary cooling loop by thermal conduction through the fuel cladding. The primary heat cooler is pumped into a heat exchanger called a steam generator, where it flows through hundreds or thousands of small tubes. Heat is transferred through the wall of this tube to a low pressure secondary cooling located on the side of the exchanger sheet where the coolant evaporates into the pressurized vapor. The heat transfer is done without mixing the two liquids to prevent the secondary fluid from becoming radioactive. Some general steam generator setups are u-tube or single pass heat exchanger.
At a nuclear power plant, pressurized steam is introduced through a steam turbine that drives an electrical generator connected to the power grid for transmission. After passing through the turbine, the secondary coolant (water-vapor mixture) is cooled and condensed in the condenser. The condenser converts the vapor to a liquid so it can be pumped back to the steam generator, and maintains the vacuum at the turbine outlet so that the pressure drop in the turbine, and therefore the energy extracted from the vapor, is maximized. Prior to incorporation into the vapor generator, condensed vapor (referred to as feed water) is sometimes heated to minimize thermal shock.
The resulting vapor has other uses than a power plant. In nuclear and submarine vessels, steam is introduced through a steam turbine connected to a set of speed reducing gears to the shaft used for propulsion. Direct mechanical action with steam expansion can be used to catapel steam-powered aircraft or similar applications. District heating by steam is used in some countries and direct heating is applied to internal plant applications.
Two things are characteristic for pressurized water reactors (PWRs) when compared to other reactor types: separation of the cooling loop from the steam system and the pressure inside the primary cooling loop. In PWR, there are two separate cooling loops (primary and secondary), both of which are filled with demineralized/deionized water. A boiling water reactor, on the other hand, has only one cooling ring, while more exotic designs such as a breeder reactor use a substance other than water for coolants and moderators (eg sodium in liquid as coolant or graphite as a moderator). Pressure in primer cooling circles is usually 15-16 megapascals (150-160 bar), which is mainly higher than other nuclear reactors, and almost twice that of the boiling water reactor (BWR). As a result, only localized boiling occurs and the vapors will soon be reconditioned in bulk liquids. Instead, in a boiling water reactor the main cooler is designed to boil.
Reactor
Cooler
Light water is used as the main coolant in PWR. Water enters through the bottom of the reactor core at about 548 ° C (275 ° C; 527 ° F) and is heated as it flows up through the reactor core to a temperature of about 588 ° C (315 ° C; 599 ° ° F). Water remains liquid despite high temperatures due to high pressure in the primary cooling loop, typically about 155 bar (15.5 MPa 153Ã, atm, 2,250Ã, psi). In water, a critical point occurs at about 647 ° C (374 ° C; 705 ° F) and 22,064 MPa (3200 psi or 218 atm).
Pressurizer
The pressure in the primary circuit is maintained by a pressurizer, a separate vessel connected to the primary circuit and partially filled with water heated to a saturation temperature (boiling point) for the pressure desired by a submerged electric heater. To achieve a pressure of 155 bar (15.5 MPa), the pressurizer temperature is maintained at 345 ° C (653 ° F), which gives the sub cooling margin (the difference between the pressurizer temperature and the highest temperature in the reactor core). ) from 30 Â ° C (54 Â ° F). Since 345 ° C is the boiling point of water at 155 bar, liquid water is at the edges of phase change. The thermal transients in the reactor coolant system result in major changes in the volume of the liquid/vapor pressurizer, and the total pressure volume is designed around absorbing this transient without opening the heater or emptying the pressurizer. The transient pressure in the primary cooling system manifests as a transient temperature in the pressurizer and is controlled through the use of automatic heaters and water sprays, which increase and decrease the pressurizer temperature, respectively.
Pump
Coolant is pumped around the main circuit by a powerful pump. After extracting heat as it passes through the reactor core, the main coolant transfers heat in the steam generator to water in a low-pressure secondary circuit, vaporizing the secondary coolant to saturated vapor - in most 6.2 MPa (60Ã, atm, 900Ã, Â °) designs. psia), 275 Â ° C (530 Â ° F) - for use in steam turbines. The cooled primary cooler is then returned to the reactor vessel for heating again.
Moderator
Pressurized water reactors, like all thermal reactor designs, require fast fission neutrons to be slowed (a process called moderation or heat) to interact with nuclear fuel and maintain chain reactions. In PWR cooling water is used as a moderator by letting the neutrons experience multiple collisions with light hydrogen atoms in the water, losing speed in the process. This "moderation" neutrons will occur more often when water is denser (more collisions will occur). The use of water as a moderator is an important safety feature of PWR, since increased temperatures can cause water to expand, providing a larger 'gap' between water molecules and reducing the possibility of thermalization - thereby reducing the extent to which neutrons are slowed and hence reducing reactivity in the reactor. Therefore, if the reactivity increases beyond normal, a reduction in neutron moderation will cause the chain reaction to slow down, resulting in less heat. This property, known as the negative temperature reactivity coefficient, makes the PWR reactor extremely stable. This process is referred to as 'Self-Regulating', ie the more heat the coolant becomes, the less reactive the plant becomes, closing itself slightly to compensate and vice versa. Thus, the plant controls itself around a certain temperature determined by the position of the control rod.
In contrast, the design of the RBMK reactor used at Chernobyl, which uses graphite instead of water as a moderator and uses boiling water as a coolant, has a large positive thermal reactivity coefficient, which increases heat generation when the cooling water temperature increases. This makes the RBMK design less stable than the pressurized water reactor. In addition to its slow neutrons when serving as a moderator, water also has properties absorbing neutrons, albeit to a lesser extent. As the cooling water temperature increases, boiling increases, which creates the cavity. So there is less water to absorb the thermal neutrons that have been slowed by graphite moderators, leading to increased reactivity. This property is called the reactivity vacuum coefficient, and in RBMK reactors such as Chernobyl, the vacancy coefficient is positive, and large enough, causing rapid transients. This characteristic of the RBMK reactor design is generally seen as one of several causes of the Chernobyl disaster.
Heavy water has very low absorption of neutrons, so heavy water reactors tend to have positive void coefficients, although CANDU reactor designs mitigate this problem by using unreachable natural uranium; The reactor is also designed with a number of passive security systems not found in the original RBMK design.
PWR is designed to be maintained in an unmoderated state, meaning that there is room for increased water volume or density to further increase moderation, because if moderation approaches saturation, moderate/coolant density reductions can significantly reduce the absorption of neutrons while reducing moderation, positive void coefficient. Also, light water is actually a moderately strong neutron moderator rather than heavy water, although the absorption of heavy water neutrons is much lower. Because of these two facts, the light water reactor has a relatively small volume of moderators and therefore has a compact nucleus. One next generation design, a supercritical water reactor, is even less moderated. The less moderate neutron energy spectrum exacerbates the catch/fission ratio for 235 U and especially 239 Pu, which means that more fission nuclei fail in fission on neutron absorption and vice versa capture neutrons to heavier nonfissile isotopes, removing one or more neutrons and increasing the accumulation of heavy transuranic actinides, some of which have a long half-life.
Fuel
After enrichment, uranium dioxide ( UO
2 ) powder fired in high temperature, sintering furnace to make hard, enriched uranium dioxide ceramic pellets. The cylinder pellet is then bandaged with Zircaloy corrosion-resistant zirconium alloys that are hoarded with helium to aid heat conduction and detect leakage. Zircaloy was chosen because of its mechanical properties and its low cross-section of absorption. The finished fuel rods are grouped in a fuel assembly, called a fuel bundle, which is then used to build the reactor core. A typical PWR has a fuel assembly of 200 to 300 bars each, and a large reactor will have about 150-250 such assemblies with 80-100 tons of uranium in all. Generally, the fuel bundle consists of fuel rods bundled 14 ÃÆ'â € "Ã, 14 to 17Ã, ÃÆ'â €" 17. A PWR produces on the order of 900 to 1,600 mW e . PWR fuel package has a length of about 4 meters.
Refueling for most commercial PWRs is on the 18-24 month cycle. About a third of the core is replaced by each refueling, although some more modern refueling schemes can reduce refueling times by several days and allow fueling to occur in shorter periods.
Control
In the PWR reactor power can be seen as follows steam (turbine) demand due to feedback reactivity of the temperature changes caused by the increase or decrease of steam flow. (See: Negative temperature coefficient.) Boron and control rods are used to maintain the temperature of the primary system at the desired point. To reduce power, the operator will close the turbine valve. This will produce less steam drawn from the steam generator. This results in an increase in the primary loop in temperature. Higher temperatures cause the main reactor coolant water densities to decrease, allowing higher neutron speeds, resulting in less fission and decreased power output. This decrease in power will eventually result in the temperature of the primary system back to the previous steady-state value. The operator can control the steady state operating temperature by addition of boric acid and/or control rod movements.
Reactivity adjustments to maintain 100% power as fuel burned in most commercial PWRs are usually achieved by varying the concentration of boric acid dissolved in the primary reactor coolant. Boron readily absorbs neutrons and increases or decreases its concentration in the reactor coolant because it affects neutron activity simultaneously. All control systems involving high pressure pumps (usually called filling and letdown systems) are needed to remove water from high pressure main loops and inject water back with different concentrations of boric acid. The reactor control rod, inserted through the head of the reactor vessel directly into the fuel bundle, is removed for the following reasons:
- To start the reactor.
- To turn off the main nuclear reaction in the reactor.
- To accommodate short-term transients, such as changes to load on turbines.
Control bars can also be used:
- To compensate for the inventory of nuclear toxins.
- To compensate for the depletion of nuclear fuel.
However, this effect is more usually accommodated by altering the concentration of boric acid primary coolant.
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Source of the article : Wikipedia
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