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 water then flows to a steam generator where it transfers its thermal energy to a secondary system where steam is generated and flows to turbines which, in turn, spin an electric generator. In contrast to a boiling water reactor, pressure in the primary coolant loop prevents the water from boiling within the reactor.
All LWRs use ordinary water as both coolant and neutron moderator. 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.
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.
Rickover that a viable commercial plant would include none of the "crazy thermodynamic cycles that everyone else wants to build.
The pressurized water reactor has three new Generation III reactor evolutionary designs: 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. 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. Two things are characteristic for the pressurized water reactor PWR when compared with other reactor types: 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.
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. The water remains liquid despite the high temperature due to the high pressure in the primary coolant loop, usually around bar 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 bars 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. The cooled primary coolant is then returned to the reactor vessel to be heated again. Pressurized water reactors, like all thermal reactor designs, require the fast fission neutrons to be slowed down a process called moderation or thermal 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 thermalisation—thereby reducing the extent to which neutrons are slowed down 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.
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 down 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.
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. 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.
A typical PWR has fuel assemblies of to rods each, and a large reactor would have about — such assemblies with 80— tonnes of uranium in all. 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.
Boron and 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.
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:.
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.
From Wikipedia, the free encyclopedia. Nuclear Regulatory Commission image of pressurized water reactor vessel heads. An animation of a PWR power station with cooling towers. Setting the Nuclear Navy's Course". Oak Ridge National Laboratory , U. Archived from the original PDF on February 25, Union of Concerned Scientists.
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