A pressurized water reactor (PWR) is a type of light-water nuclear reactor. PWRs are the most common type of nuclear power reactor, representing almost 70% of the world's commercial reactor fleet. Some countries, such as Canada and (historically) the United Kingdom, have relied heavily on other designs (CANDU pressurized heavy‑water reactors and gas‑cooled reactors respectively) while others, including Japan and India, operate a mix of PWRs and other reactor types (notably boiling water reactors and pressurized heavy‑water reactors).

In a PWR, water is used both as a neutron moderator and as coolant fluid for the reactor core. In the core, water is heated by the energy released by the fission of atoms contained in the fuel. Using high pressure (around 155 bar: 2250 psi) ensures that the water stays in a liquid state. The heated water then flows to a steam generator, where it transfers its thermal energy to the water of a secondary cycle kept at a lower pressure which allows it to vaporize. The resulting steam then drives steam turbines linked to an electric generator. A boiling water reactor (BWR) by contrast does not maintain such high pressure in the primary cycle and the water vaporizes inside the reactor pressure vessel before being sent to the turbine. Most PWR designs make use of two to six steam generators, each associated with a coolant loop.

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 are operated in the United States, France, Russia, China, South Korea and several other countries. The majority are Generation II reactors; newer Generation III designs such as the AP1000, Hualong One, APR-1400 and VVER-1000. New Generation III+ reactor designs: the AP1000, VVER-1200, CAP-1400, and EPR. The first AP1000 and EPR reactors were connected to the power grid in China in 2018.

History

Development of pressurised water reactors began in 1946 with the initiation of the US Naval Nuclear Propulsion Program at Oak Ridge National Laboratory in Tennessee to develop nuclear propulsion for submarines. The first fully operational submarine power plant was 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, becoming the first new nuclear reactor in the United States since 1996.

In 2020, NuScale Power became the first U.S. company to receive regulatory approval from the Nuclear Regulatory Commission for a small modular reactor with a modified PWR design. Also in 2020, the Energy Impact Center introduced the OPEN100 project, which published open-source blueprints for the construction of a 100 MWelectric nuclear power plant with a PWR design.

Design

Nuclear fuel in the reactor pressure vessel is engaged in a controlled 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 several thousand small tubes. Heat is transferred through the walls of these tubes to the lower pressure secondary coolant located on the shell side of the exchanger where the secondary coolant evaporates to pressurized steam. This 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.

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 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.

Two characteristics for pressurized water reactor when compared with other reactor types are: coolant loop separation from the steam system and pressure inside the primary coolant loop. In a PWR, there are two coolant loops (primary and secondary), which are both filled with demineralized/deionized water. A boiling water reactor, by contrast, has 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 , which is notably higher than in other nuclear reactors, and nearly twice that of a boiling water reactor. 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.

Reactor

Coolant

Light water is used as the primary coolant in a PWR. Water enters through the bottom of the reactor's core at about and is heated as it flows upwards through the reactor core to a temperature of about . 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). The water in a PWR cannot exceed a temperature of or a pressure of 22.064 MPa (3200 psi or 218 atm), because those are the critical point of water.International Association for the Properties of Water and Steam, 2007. Supercritical water reactors are (as of 2022) only a proposed concept in which the coolant would never leave the supercritical state. However, as this requires higher pressures than a PWR and can cause issues of corrosion, no such reactor has yet been built.

Pressurizer

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 , 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.

Pumps

The coolant is pumped around the primary circuit by powerful pumps. 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 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.

Moderator

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, producing less heat. This property, known as the negative temperature coefficient of reactivity, makes PWR reactors 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 Soviet RBMK reactor design used at Chernobyl, Ukraine, which uses graphite instead of water as the moderator and uses boiling water as the coolant, has a large positive thermal coefficient of reactivity. This means reactivity and heat generation increases when coolant and fuel temperatures increase, which makes the RBMK design less stable than pressurized water reactors at high operating temperature. In addition to its property of slowing 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, making it very hard to regulate when the reaction begins to run away. The RBMK reactors also have a flawed control rod design in which during rapid scrams, the graphite reaction enhancement tips of the rods would displace water at the bottom of the reactor and locally increase reactivity there. This is called the "positive scram effect" that is unique to the flawed RBMK control rods design. These design flaws, in addition to operator errors that pushed the reactor to its limits, are generally seen as the causes of the Chernobyl disaster.