| Pressurized Water Reactor |
Website Links For Water |
Information AboutPressurized Water Reactor |
| CATEGORIES ABOUT PRESSURIZED WATER REACTOR | |
| nuclear power reactor types | |
| pressurized water reactors | |
|
Heat from small PWRs has been used for heating in polar regions, see Army Nuclear Power Program . The Three Mile Island Accident occurred in a PWR manufactured by Babcock & Wilcox . OVERVIEW 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:
Note on Boiling in PWRs Both PWRs and BWR s are designed such that the water in the core will never reach Film Boiling during normal operation. In PWRs, however, the water is not required to boil, it just occurs as a consequence of the conditions at which the core is operated. This localized boiling that recondenses in the bulk fluid does help improve the heat transfer rate from the fuel and its cladding. This improves safety margins and efficiency. PWR REACTOR DESIGN Coolant PWR reactor hall and cooling tower (being decommissioned, 2004)]] Water is used as primary coolant in a PWR and flows through the reactor at a temperature of roughly 315 °C (600 °F ). The water remains liquid despite the high temperature due to the high pressure in the primary coolant loop (usually around 2200 Psig [15 MPa , 150 Atm ]). The primary coolant loop is used to heat water in a secondary circuit that becomes saturated steam (in most designs 900 Psia MPa, 60 atm , 275 °C °F ) for use in the steam turbine. Although coolant flow rate in commercial PWRs is constant, it is not in nuclear reactors used on U.S. Navy ships. Moderator Pressurized water reactors, like Thermal Reactor designs, require the fast fission neutrons in the reactor to be slowed down (a process called moderation) in order to sustain its 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 any increase in temperature causes the water to expand and become less dense; thereby reducing the extent to which neutrons are slowed down and hence reducing the reactivity in the reactor. Therefore, if reactor activity 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. In contrast, the RBMK reactor design used at Chernobyl (using graphite instead of water as the moderator) greatly increases heat generation when coolant water temperatures increase, making them very unstable. This flaw in the RBMK reactor design is generally seen as one of several causes of the Chernobyl Accident . Fuel See Also: Nuclear fuel . Designed and built by the Babcock and Wilcox Company.]] The uranium used in PWR fuel is usually enriched several percent in 235U. 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 put into tubes of a corrosion-resistant zirconium metal alloy (Zircaloy) which are backfilled with helium to aid heat conduction and detect leakages. 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 (as is suspected to have occurred during the accident at the Chernobyl plant). 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. Refuelings for most commercial PWRs is on an 18-24 month cycle. Approximately one third of the core is replaced each refueling. Control 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 PWR's 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 normally only used for power changes. 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. ADVANTAGES
DISADVANTAGES
SEE ALSO NEXT GENERATION DESIGNS
|
|
|