Pressurized water reactor

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A pressurised water reactor (PWR) is a type of nuclear power reactor that uses ordinary light water for both coolant and for neutron moderation.

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Contents

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Description

In a PWR, the primary coolant loop is pressurized and this water does not achieve bulk-boiling; heat exchangers called steam generators are used to transfer heat to a secondary coolant which is allowed to boil to produce steam either for warship propulsion or for electricity generation. In having this secondary loop, the PWR differs from the boiling water reactor (BWR), in which the primary coolant is allowed to boil in the reactor core and drive a turbine directly. Heat from small PWRs has also been used for heating in polar regions, see Army Nuclear Power Program.

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This is one of the most common types of nuclear power reactor and is widely used all over the world. More than 230 are in use to generate electric power, and several hundred more for naval propulsion. It was originally designed by the Bettis Atomic Power Laboratory as a nuclear submarine power plant.

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Nuclear reactor design

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Coolant

Neutrons striking nuclear fuel (early in the cycle, mainly U-235) in fuel rods lead to fissioning of the fissile atoms, releasing more neutrons and heat. The heat transfers from the fuel ceramic pellets to the surrounding metal fuel "cladding" which in turn heats the water flowing by the fuel rods. The fuel rods are arranged in a matrix (a fuel bundle). Water flows in between the fuel rods from the bottom to the top of the reactor -- the bundles are 12 to 14 feet long depending on the vintage of the reactor. That water flows to a steam generator. There, the heat (~315 °C ~(600 °F) and 2200 psig / 150 atm) passes to water in a secondary circuit that becomes saturated steam (in most designs 900 psia / 60 atm, 275 °C (530 °F)) for use in the steam plant.

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Moderator

Nuclear fission produces neutrons that are too energetic to trigger significant further fission within the reactor fuel. Their energy must first come down to so-called "thermal" levels in rough equilibrium with the temperature of the surrounding medium, which might be 450 °C (800 °F). In the PWR, these neutrons initially lose heat when they collide with molecules of coolant water. After several collisions (8 to 10 on average), a neutron reaches the temperature of its surroundings and is likely to be absorbed by a uranium-235 atom. Such absorption leads quickly to fission of the uranium atom.

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Nuclear fuel

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Uranium dioxide (UO2) powder is fired in a high-temperature, sintering furnace to create hard, ceramic pellets of enriched uranium dioxide. The cylindrical pellets are then put into tubes of a corrosion-resistant zirconium metal alloy (Zircoloy). The tubes are then backfilled with helium; the tubes are now fuel rods. The finished fuel rods are grouped in fuel assemblies 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 tonnes of uranium in all. Generally, the fuel bundles consist of fuel rods bundled 14x14 to 17x17. A PWR produces on the order of 900 to 1500 MWe. PWR fuel bundles are about 4 meters in length. Control rods are inserted through the top directly into the fuel bundle. The fuel bundles usually are enriched several percent in 235U. The uranium oxide is dried before inserting into the tubes to try to eliminate moisture in the ceramic fuel that can lead to corrosion and hydrogen embrittlement.

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Control

A key mechanism that controls any nuclear reactor is the rate at which fission events release neutrons. On average, each fission releases just over two neutrons with a large amount of heat. When a neutron strikes a uranium atom a further fission event can occur, and this can lead to a chain reaction. If all neutrons were released instantaneously, their number would grow very fast, resulting in the destruction of the fuel ceramic and a melt-down of the reactor. However, a small fraction of these neutrons are released over an extended period (perhaps one minute). This small, but crucial, delayed release permits the other control mechanisms (negative temperature co-efficient, human or computer manipulation of neutron-absorbing control rods, etc.) to have an effect.

Reactor power in most commercial and military PWR's is controlled during normal power operations by varying the concentration of boron (in the form of boric acid) in the primary reactor coolant. Reactor coolant flow rate in commercial PWRs is constant. Although in nuclear reactors used on U.S. Navy ships, reactor coolant flow rate is not constant, and is not used to control reactor power. Power in most naval nuclear reactors is regulated by the height of the control rods. Boron is a strong neutron absorber. 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 are only used for startup and shut down operations. In contrast, BWR's have no boron in the reactor coolant and control reactor power by adjusting the reactor coolant flow rate. This is an advantage for the BWR design because boric acid is very corrosive and the complex charging and letdown system is not required. However, as a backup to control-blade insertion, most commercial BWRs do have an emergency shutdown system which involves injecting a highly concentrated boric acid solution into the primary coolant circuit. CANDU reactors also inject boron as a backup means to shut down the nuclear chain reaction.

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Advantages

Water in a PWR reactor core reaches about 325°C (617°F), only remaining liquid under about 150 times atmospheric pressure to prevent bulk-boiling. Pressure is maintained by steam in a pressuriser, a separate tank connected directly to the reactor primary coolant circuit with electric heaters for increasing pressure in the overall system and cooling sprays for reducing pressure in the overall system. In the reactor core, the primary coolant (water) is also the neutron moderator. Moderating or slowing the fast neutrons ejected from the fuel nuclei is a requirement for the fission process to occur. As the fission process releases more heat and the reactor coolant temperature increases, the density of the coolant decreases (hot water is less dense than cold water) and its ability to serve as a neutron moderator decreases so the fission reaction slows down and the coolant temperature decreases. This negative feedback effect is called a negative temperature coefficient and is one of the safety features of the PWR. (The RBMK reactors at Chernobyl had a positive temperature coefficient which was one of the contributing factors to the accident). A disadvantage is that the reactor is susceptible to produce power at rates that result in damage to fuel in the event of introduction of cold water into the reactor or in the event the secondary system experiences a steam line rupture. A steam line rupture is a cooling event to the reactor coolant.

Another advantage of using coolant water as a moderator in a pressurized water reactor is that the moderating effect also decreases as a function of boiling which creates voids of steam in the coolant. Again, the formation of voids reduces the density of the coolant and thereby reduces the moderating efficiency of the coolant which slows the fission process. This is called a negative void coefficient. This negative void coefficient also acts as a negative feedback loop, ensuring that reactor power is self limiting. (The RBMK reactors at Chernobyl had a positive void coefficient which was one of the contributing factors to the accident. Also, the Candu reactors have a positive void coefficient, but it is very small.)

The secondary circuit is under less pressure than the primary. The secondary water boils in heat exchangers which generate steam (i.e., steam generators). The steam drives the turbine to produce electricity or turn a drive shaft of a ship. This steam then condenses into water and returns as feedwater to the steam generators.

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Disadvantages

One disadvantage to fission reactors (both PWR and BWR) is that radioactive decay continues to generate significant heat even after the fission reaction stops (up to 7% full power in the first instances after control rod insertion), possibly leading to nuclear meltdown if the reactor loses numerous primary and emergency means of circulating reactor (water) coolant. Reactor plants typically have extensive safety and backup systems to prevent this. However, the complexity of these systems has been criticized on the grounds that in an emergency, they may be prone to unexpected interactions and operator error. Therefore, each reactor is surrounded by a containment building designed as a final barrier to radioactive release.

A Babcock and Wilcox pressurized water reactor was involved in the accident at Three Mile Island. That design (B&W) uses much smaller steam generators and creates super-heated steam. Thus, the operators have a relatively short time to restore feedwater to the steam generators in a B&W design as compared to other designs such as Westinghouse. Much of the research in civilian nuclear reactors has been targeted to improve their resilience even after extensive equipment failure.

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U.S. commercial pressurized water reactor (PWR) nuclear power plants

(a complete list of nuclear reactors can be found at list of nuclear reactors. To see those in the United States, see the United States section of the same list.)

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Other commercial pressurized water reactor (PWR) nuclear power plants

Note that this list is very incomplete. The PWR is the most popular reactor type worldwide.
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See also

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Next generation designs

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