Pressurized water reactor
Pressurised water reactors (
PWRs) are
nuclear power reactors that use ordinary
water for three tasks: as the primary coolant, the secondary coolant, and for
neutron moderation. They belong to the family of
light water reactors.
PWRs are one of the most common types of reactors and are widely used all over the world. More than 230 of them are in use to generate electric power, and several hundred more for naval propulsion. They were originally designed by the
Bettis Atomic Power Laboratory as a
nuclear submarine power plant.
Heat from small PWRs has also been used for heating in polar regions, see
Army Nuclear Power Program.
|
Rancho Seco PWR reactor hall and cooling tower (being decommissioned, 2004) |
How it works
A PWR works because the
nuclear fuel in the
reactor vessel is engaged in a
chain reaction, which produces heat as the main goal of the entire setup. This heats the water in the primary coolant loop (shown in the schematic as a red dashed line) that is constantly being fed into the reactor vessel. The hot water is pumped into a certain type of
heat exchanger called
steam generator, which allows the primary coolant to heat up the secondary coolant (shown as the loop
steam generator →
turbine →
condenser). The transfer of heat is accomplished without mixing the two fluids since the primary coolant is necessarily radioactive, but it is desirable to avoid this for the secondary coolant. Once the secondary coolant is hot enough to turn into steam, it can be used for warship propulsion or for generating electricity via a
turbine, as shown above. The secondary coolant afterwards is cooled down in a
condenser before being fed into the steam generator again.
In comparison
Two things are characteristic for the pressurized water reactor (PWR) when compared with other reactor types:
* In a PWR, there are two separate coolant loops (primary and secondary), which are both filled with ordinary
water (also called
light 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 the task.
* The pressure in the primary coolant loop is at typically 16
Megapascal, notably higher than in other
nuclear reactors. As an effect of this, the
gas laws guarantee that the primary coolant loop's water will never boil during the normal operation of the reactor. By contrast, the coolant does boil in
boiling water reactors.
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.
Moderator
Nuclear fission produces neutrons that are too hot 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.
Nuclear fuel
|
PWR fuel bundle The fuel bundle is from a pressurized water reactor of the nuclear passenger and cargo ship NS Savannah. Designed and built by the Babcock and Wilcox Company. |
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 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 MW
e. 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.
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 lot 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.
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
CANDU reactors also 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.
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.
(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.)
*
Arkansas Nuclear One,
Arkansas*
Beaver Valley,
Pennsylvania*
Bellefonte,
Alabama (Unfinished)
*
Braidwood,
Illinois*
Byron,
Illinois*
Callaway,
Missouri*
Calvert Cliffs,
Maryland*
Catawba,
South Carolina*
Comanche Peak,
Texas*
Connecticut Yankee,
Connecticut (Decommissioned)
*
Crystal River 3,
Florida*
Donald C. Cook,
Michigan*
Davis-Besse,
Ohio*
Diablo Canyon,
California*
Farley,
Alabama*
Fort Calhoun,
Nebraska*
Ginna,
New York*
Shearon Harris,
North Carolina*
Indian Point,
New York*
Kewaunee,
Wisconsin*
McGuire,
North Carolina*
Millstone,
Connecticut*
North Anna,
Virginia*
Oconee,
South Carolina*
Palisades,
Michigan*
Palo Verde,
Arizona*
Point Beach,
Wisconsin*
Prairie Island,
Minnesota*
Rancho Seco,
California (Decommissioned)
*
Robinson,
South Carolina*
St. Lucie,
Florida*
Salem,
New Jersey*
San Onofre,
California*
Seabrook,
New Hampshire*
Sequoyah,
Tennessee *
Shippingport,
Pennsylvania (Decommissioned)
*
South Texas,
Texas*
Summer,
South Carolina*
Surry,
Virginia*
Three Mile Island,
Pennsylvania*
Trojan,
Oregon (Decommissioned)
*
Turkey Point,
Florida *
Vogtle,
Georgia *
Waterford,
Louisiana*
Watts Bar,
Tennessee *
Wolf Creek,
Kansas*
Zion,
Illinois (Decommissioned)
*
Sizewell B,
Suffolk, UK
| Note that this list is very incomplete. The PWR is the most popular reactor type worldwide. |
*
Nuclear Power 2010 ProgramNext generation designs
*
European Pressurized Reactor (EPR)
*
Westinghouse Advanced Passive 1000 (AP1000)
