Pressurized water reactor (PWRs) (also VVER if of Russian design) are generation II nuclear power reactors that use ordinary water under high pressure (superheated water) as coolant and neutron moderator. The primary coolant loop is kept under high pressure to prevent the water from reaching film boiling, hence the name. PWRs are the most common type of power producing reactor 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 at the Oak Ridge National Laboratory for use as a nuclear submarine power plant. Follow-on work was conducted by Bettis Atomic Power Laboratory.
A PWR works because the nuclear fuel in the reactor vessel is engaged in a chain reaction, which produces heat, heating the water in the primary coolant loop by thermal conduction through the fuel cladding. (The primary coolant loop is shown in the schematic as a red dashed line.) The hot water is pumped into a heat exchanger called steam generator, which allows the primary coolant to heat up and boil the secondary coolant (shown as the loop steam generator → turbine → condenser). The transfer of heat is accomplished without mixing the two fluids. This is desirable, since the primary coolant is necessarily radioactive. The steam formed in the steam generator is allowed to flow through a steam turbine, and the energy extracted by the turbine is used to drive an electric generator. Other uses for the steam from a PWR include:
In a nuclear power station, the steam is fed through a steam turbine which drives a generator connected to the electric grid for distribution, as shown above. After passing through the turbine the secondary coolant (water-steam mixture) is cooled down and condensed in a condenser before being fed into the steam generator again. This converts the steam to a liquid so that it can be pumped back into the high pressure steam generator.
Two things are characteristic for the pressurized water reactor (PWR) when compared with other reactor types:
Heat from small PWRs has been used for heating in polar regions in the Army Nuclear Power Program.
The coolant is pumped around the primary circuit by powerful pumps, which can consume up to 6 MWe each. After picking up heat as it passes through the reactor core, the primary coolant gives up 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 pumped back to the reactor to be heated again.
Although coolant flow rate in commercial PWRs is constant, it is not in nuclear reactors used on U.S. Navy ships.
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 high positive 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 lessor 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 accident.
CANDU reactors, (which use heavy water as a coolant and neutron moderator) also have a positive void coefficient, though it is not as large as that of an RBMK like Chernobyl; these reactors are designed with a number of safety systems not found in an RBMK, which are designed to handle or react to this as needed.
After enrichment the 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 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. The finished fuel rods are grouped in fuel assemblies, called fuel bundles, that are then used to build the core of the reactor. As a safety measure PWR designs do not contain enough fissile uranium to sustain a prompt critical chain reaction (i.e, substained only by prompt neutrons). Avoiding prompt criticality is important as a prompt critical chain reaction could very rapidly produce enough energy to damage or even melt 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 14 × 14 to 17 × 17. A PWR produces on the order of 900 to 1,500 MWe. PWR fuel bundles are about 4 meters in length.
Refuelings for most commercial PWRs is on an 18–24 month cycle. Approximately one third of the core is replaced each refueling.
Generally, 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 reactor to fission less and decrease in power. The operator could then add boric acid and/or insert control rods to decrease temperature to the desired point.
Reactivity adjustments to maintain 100% power as the fuel is burned up in most commercial PWRs is normally controlled by varying the concentration of boric acid dissolved in the primary reactor coolant. The 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 top directly into the fuel bundles, are moved for the following reasons:
1. To enable reactor start up.
2. To shut down the reactor.
3. To compensate for nuclear poison inventory.
4. To compensate for nuclear fuel depletion.
In contrast, BWRs have no boron in the reactor coolant and control the reactor power by adjusting the reactor coolant flow rate.
Due to design and fuel enrichment differences, naval nuclear reactors do not use boric acid.
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