Pressurized water reactor

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Rancho Seco PWR reactor hall and cooling tower (being decommissioned, 2004)

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 to remove heat generated by nuclear chain reaction from nuclear fuel, and as the moderator to thermalise the neutron flux so that it interacts with the nuclear fuel to maintain the chain reaction. 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 nuclear reactor, and are widely used in power stations such as Three Mile Island in the USA, ships and submarines all over the world. More than 230 of them are in use in nuclear power plants to generate electric power, and several hundred more for marine propulsion in aircraft carriers, submarines and ice breakers. They were originally designed at the Oak Ridge National Laboratory in the USA for use as a nuclear submarine power plant. Follow-on work was conducted by Bettis Atomic Power Laboratory.[1]


[edit] Overview

Nuclear fuel in the reactor 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 primary coolant loop is shown in the schematic as a red dashed line.) The hot primary coolant is pumped into a heat exchanger called steam generator, where heat is transferred to the lower pressure secondary coolant, which evaporates to pressurised steam (shown as the loop steam generatorturbinecondenser). The transfer of heat is accomplished without mixing the two fluids, which is desirable since the primary coolant might become radioactive.

In a nuclear power station, the pressurised steam is fed through a steam turbine which drives an electrical 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. 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 maximised.

Other uses for the steam from a PWR include:

  • 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 for things like a steam-powered aircraft catapult
  • District heating by the steam

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 demineralized/deionized 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 (e.g. sodium in its liquid state).
  • The pressure in the primary coolant loop is typically 15–16 megapascals (150–160 bar), which is notably higher than in other nuclear reactors, and nearly twice that of a Boiling water reactor (BWR). As an effect of this, only localized boiling occurs and will recondense promptly in the bulk fluid. By contrast, in a boiling water reactor the primary coolant is designed to boil.

Heat from small PWRs has been used for heating in polar regions in the Army Nuclear Power Program.

[edit] PWR reactor design

[edit] Coolant

Light water is used as the primary coolant in a PWR. It enters the bottom of the reactor core at about 275 °C (530 °F) and is heated as it flows upwards through the reactor core to a temperature of about 315 °C (600 °F). The water remains liquid despite the high temperature due to the high pressure in the primary coolant loop, usually around 155 bar (15 MPa 150 atm, 2,250 psig). Pressure in the primary circuit is maintained by a Pressuriser, 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 155 bar, the pressuriser temperature is maintained at 345 °C, which gives a subcooling margin (the difference between the pressuriser temperature and the highest temperature in the reactor core) of 30 °C. To achieve maximum heat transfer, the primary circuit temperature, pressure and flow rate are arranged such that subcooled nucleate boiling takes place as the coolant passes over the nuclear fuel rods.

The coolant is pumped around the primary circuit by powerful pumps, which can consume up to 6 MW 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.

[edit] Moderator

Pressurized water reactors, like thermal reactor designs, require the fast fission neutrons to be slowed down (a process called moderation or thermalization) 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 denser (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 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.

See main article: Passive nuclear safety

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.[2]

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.

[edit] Fuel

PWR fuel bundle This 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.

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.[3] 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.

[edit] Control

NRC image of PWR Reactor Vessel heads.
PWR reactor vessel

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. (See: Negative temperature coefficient.) 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 adjustment to maintain 100% power as the fuel is burned up in most commercial PWRs is normally achieved by varying the concentration of boric acid dissolved in the primary reactor coolant. 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:

  • To start up the reactor.
  • To shut down the reactor.
  • To accommodate short term transients such as changes to load on the turbine.

The control rods can also be used:

but 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.

Due to design and fuel enrichment differences, naval nuclear reactors do not use boric acid.

[edit] Advantages

  • PWR reactors are very stable due to their tendency to produce less power as temperatures increase; this makes the reactor easier to operate from a stability standpoint.
  • PWR reactors can be operated with a core containing less fissile material than is required for them to go prompt critical. This significantly reduces the chance that the reactor will run out of control and makes PWR designs relatively safe from criticality accidents.
  • PWR turbine cycle loop is separate from the primary loop, so the water in the secondary loop is not contaminated by radioactive materials.

[edit] Disadvantages

  • The coolant water must be highly pressurized to remain liquid at high temperatures. This requires high strength piping and a heavy pressure vessel and hence increases construction costs. The higher pressure can increase the consequences of a loss of coolant accident.
  • Most pressurized water reactors cannot be refueled while operating. This decreases the availability of the reactor- it has to go offline for comparably long periods of time (some weeks).
  • The high temperature water coolant with boric acid dissolved in it is corrosive to carbon steel (but not stainless steel), this can cause radioactive corrosion products to circulate in the primary coolant loop. This not only limits the lifetime of the reactor, but the systems that filter out the corrosion products and adjust the boric acid concentration add significantly to the overall cost of the reactor and radiation exposure. Occasionally, this has resulted in severe corrosion to control rod drive mechanisms when the boric acid solution leaked through the seal between the mechanism itself and the primary system.[4][5]
  • Natural uranium is only 0.7% Uranium-235, the isotope necessary for thermal reactors. This makes it necessary to enrich the uranium fuel, which increases the costs of fuel production. If heavy water is used it is possible to operate the reactor with natural uranium, but the production of heavy water requires large amounts of energy and is hence expensive.

[edit] Notes

[edit] See also

[edit] Next generation designs

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