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103-449: There are three varieties of light-water reactors: the pressurized water reactor (PWR), the boiling water reactor (BWR), and (most designs of) the supercritical water reactor (SCWR). After the discoveries of fission , moderation and of the theoretical possibility of a nuclear chain reaction , early experimental results rapidly showed that natural uranium could only undergo a sustained chain reaction using graphite or heavy water as

206-732: A fast neutron reactor . The leaders in national experience with PWRs, offering reactors for export, are the United States (which offers the passively safe AP1000 , a Westinghouse design, as well as several smaller, modular, passively safe PWRs, such as the Babcock & Wilcox MPower , and the NuScale MASLWR), the Russian Federation (offering both the VVER-1000 and the VVER-1200 for export),

309-435: A heavy water reactor , which uses heavy water as a neutron moderator. While ordinary water has some heavy water molecules in it, it is not enough to be important in most applications. In pressurized water reactors 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

412-442: A (partially) closed nuclear fuel cycle . Water is a nontoxic, transparent, chemically unreactive (by comparison with e.g. NaK ) coolant that is liquid at room temperature which makes visual inspection and maintenance easier. It is also easy and cheap to obtain unlike heavy water or even nuclear graphite . Compared to reactors operating on natural uranium , PWRs can achieve a relatively high burnup . A typical PWR will exchange

515-479: A CANDU reactor or any other heavy water reactor when ordinary light water is supplied to the reactor as an emergency coolant. Depending on burnup , boric acid or another neutron poison will have to be added to emergency coolant to avoid a criticality accident . PWRs are designed to be maintained in an undermoderated state, meaning that there is room for increased water volume or density to further increase moderation, because if moderation were near saturation, then

618-465: A PWR cannot exceed a temperature of 647 K (374 °C; 705 °F) or a pressure of 22.064 MPa (3200 psi or 218 atm), because those are the critical point of water. Supercritical water reactors are (as of 2022) only a proposed concept in which the coolant would never leave the supercritical state. However, as this requires even higher pressures than a PWR and can cause issues of corrosion, so far no such reactor has been built. Pressure in

721-405: A PWR design. Nuclear fuel in the reactor pressure vessel is engaged in a controlled 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 several thousand small tubes. Heat is transferred through

824-470: A PWR is not suitable for most industrial applications as those require temperatures in excess of 400 °C (752 °F). Radiolysis and certain accident scenarios which involve interactions between hot steam and zircalloy cladding can produce hydrogen from the cooling water leading to hydrogen explosions as a potential accident scenario. During the Fukushima nuclear accident a hydrogen explosion damaging

927-455: A PWR. It can, however, be used in a CANDU with only minimal reprocessing in a process called "DUPIC" - Direct Use of spent PWR fuel in CANDU. Thermal efficiency , while better than for boiling water reactors , cannot achieve the values of reactors with higher operating temperatures such as those cooled with high temperature gases, liquid metals or molten salts. Similarly process heat drawn from

1030-438: A commercial pressurized water reactor assembly — and inserted into guide tubes within a fuel element. A control rod is removed from or inserted into the central core of a nuclear reactor in order to control the number of neutrons which will split further uranium atoms. This in turn affects the thermal power of the reactor, the amount of steam generated, and hence the electricity produced. The control rods are partially removed from

1133-475: A compromise of the reactor core's integrity, the resulting release of the light-water moderator will act to stop the nuclear reaction and shut the reactor down. This capability is known as a negative void coefficient of reactivity . Data from the International Atomic Energy Agency in 2009: The light-water reactor produces heat by controlled nuclear fission . The nuclear reactor core is

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1236-646: A given temperature set by the position of the control rods. In contrast, the Soviet 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. This means reactivity and heat generation increases when coolant and fuel temperatures increase, which makes the RBMK design less stable than pressurized water reactors at high operating temperature. In addition to its property of slowing down neutrons when serving as

1339-524: A grinding process to achieve a uniform pellet size. 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. The pellets are stacked, according to each nuclear core's design specifications, into tubes of corrosion-resistant metal alloy. The tubes are sealed to contain the fuel pellets: these tubes are called fuel rods. The finished fuel rods are grouped in special fuel assemblies that are then used to build up

1442-444: A heavy pressure vessel and hence increases construction costs. The higher pressure can increase the consequences of a loss-of-coolant accident . The reactor pressure vessel is manufactured from ductile steel but, as the plant is operated, neutron flux from the reactor causes this steel to become less ductile. Eventually the ductility of the steel will reach limits determined by the applicable boiler and pressure vessel standards, and

1545-476: A lightly enriched uranium, criticality could be reached. This experiment was the first practical step toward the light-water reactor. After World War II and with the availability of enriched uranium, new reactor concepts became feasible. In 1946, Eugene Wigner and Alvin Weinberg proposed and developed the concept of a reactor using enriched uranium as a fuel, and light water as a moderator and coolant. This concept

1648-632: A material full of atoms with light nuclei which do not easily absorb neutrons. The neutrons strike the nuclei and bounce off. After sufficient impacts, the velocity of the neutron will be comparable to the thermal velocities of the nuclei; this neutron is then called a thermal neutron. The light-water reactor uses ordinary water , also called light water, as its neutron moderator. The light water absorbs too many neutrons to be used with unenriched natural uranium, and therefore uranium enrichment or nuclear reprocessing becomes necessary to operate such reactors, increasing overall costs. This differentiates it from

1751-426: A moderator). 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 steam will recondense promptly in the bulk fluid. By contrast, in a boiling water reactor the primary coolant is designed to boil. Light water

1854-423: 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 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,

1957-470: A moderator. By the end of the war , following an idea of Alvin Weinberg , natural uranium fuel elements were arranged in a lattice in ordinary water at the top of the X10 reactor to evaluate the neutron multiplication factor. The purpose of this experiment was to determine the feasibility of a nuclear reactor using light water as a moderator and coolant, and clad solid uranium as fuel. The results showed that, with

2060-484: A moderator. While the world's first reactors ( CP-1 , X10 etc.) were successfully reaching criticality , uranium enrichment began to develop from theoretical concept to practical applications in order to meet the goal of the Manhattan Project , to build a nuclear explosive . In May 1944, the first grams of enriched uranium ever produced reached criticality in the low power (LOPO) reactor at Los Alamos , which

2163-632: A publicly accessible national monument. Since 1987, the United States Environmental Protection Agency has classified the burial ground as Superfund site Operable Unit 6-01, one of two such sites (along with SL-1 ) at the Idaho National Laboratory . In 1995, the EPA ordered the primary remedy of the burial ground to be: "Containment by capping with an engineered barrier constructed primarily of native materials." The site

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2266-542: A quarter to a third of its fuel load every 18-24 months and have maintenance and inspection, that requires the reactor to be shut down, scheduled for this window. While more uranium ore is consumed per unit of electricity produced than in a natural uranium fueled reactor, the amount of spent fuel is less with the balance being depleted uranium whose radiological danger is lower than that of natural uranium. The coolant water must be highly pressurized to remain liquid at high temperatures. This requires high strength piping and

2369-432: A reactor moderator and coolant, but the vast majority of Russian nuclear-powered boats and ships use light-water reactors exclusively. The reason for near exclusive LWR use aboard nuclear naval vessels is the level of inherent safety built into these types of reactors. Since light water is used as both a coolant and a neutron moderator in these reactors, if one of these reactors suffers damage due to military action, leading to

2472-399: A reactor using direct boiling of water would be practical, rather than unstable, because of the bubble formation in the core. Subsequently, the reactor was used for power excursion tests which showed that rapid conversion of water to steam would safely control the reaction. The final, deliberately destructive test in 1954 produced an unexpectedly large power excursion that "instead of

2575-469: A reduction in density of the moderator/coolant could reduce neutron absorption significantly while reducing moderation only slightly, making the void coefficient positive. 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,

2678-420: 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: coolant loop separation from

2781-502: Is a major risk factor in LWR safety record. Pressurized water reactor A pressurized water reactor ( PWR ) is a type of light-water nuclear reactor . PWRs constitute the large majority of the world's nuclear power plants (with notable exceptions being the UK, Japan and Canada). In a PWR, the primary coolant ( water ) is pumped under high pressure to the reactor core where it is heated by

2884-446: Is boiled directly by the reactor core, for example the boiling-water reactor. Many other reactors are also light-water cooled, notably the RBMK and some military plutonium -production reactors. These are not regarded as LWRs, as they are moderated by graphite , and as a result their nuclear characteristics are very different. Although the coolant flow rate in commercial PWRs is constant, it

2987-400: Is circulated past the reactor core to absorb the heat that it generates. The heat is carried away from the reactor and is then used to generate steam. Most reactor systems employ a cooling system that is physically separate from the water that will be boiled to produce pressurized steam for the turbines , like the pressurized-water reactor. But in some reactors the water for the steam turbines

3090-418: Is consumed in the reactor. Light-water reactors are generally refueled every 12 to 18 months, at which time, about 25 percent of the fuel is replaced. The enriched UF 6 is converted into uranium dioxide powder that is then processed into pellet form. The pellets are then fired in a high-temperature, sintering furnace to create hard, ceramic pellets of enriched uranium . The cylindrical pellets then undergo

3193-405: Is expected to produce no more than a 2 in 10,000 increase in cancer risk for long-term residential use after 320 years, with no significant increase after that time. This risk calculation ignores the shielding provided by the soil cover, which at the time of the EPA decision had reduced exposure to little more than background level, and makes very pessimistic modeling assumptions that greatly increase

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3296-691: Is formed into pellets and inserted into zirconium alloy tubes that are bundled together. The zirconium alloy tubes are about 1 cm in diameter, and the fuel cladding gap is filled with helium gas to improve the conduction of heat from the fuel to the cladding. There are about 179-264 fuel rods per fuel bundle and about 121 to 193 fuel bundles are loaded into a reactor core . Generally, the fuel bundles consist of fuel rods bundled 14x14 to 17x17. PWR fuel bundles are about 4 meters in length. The zirconium alloy tubes are pressurized with helium to try to minimize pellet cladding interaction which can lead to fuel rod failure over long periods. In boiling water reactors,

3399-402: Is generated per unit of uranium ore even though a higher burnup can be achieved. Nuclear reprocessing can "stretch" the fuel supply of both natural uranium and enriched uranium reactors but is virtually only practiced for light water reactors operating with lightly enriched fuel as spent fuel from e.g. CANDU reactors is very low in fissile material. Because water acts as a neutron moderator, it

3502-432: 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 thermalization — thereby reducing the extent to which neutrons are slowed and hence reducing the reactivity in the reactor. Therefore, if reactivity increases beyond normal,

3605-469: Is not in nuclear reactors used on U.S. Navy ships. The use of ordinary water makes it necessary to do a certain amount of enrichment of the uranium fuel before the necessary criticality of the reactor can be maintained. The light-water reactor uses uranium 235 as a fuel, enriched to approximately 3 percent. Although this is its major fuel, the uranium 238 atoms also contribute to the fission process by converting to plutonium 239 ; about one-half of which

3708-401: Is not possible to build a fast-neutron reactor with a PWR design. A reduced moderation water reactor may however achieve a breeding ratio greater than unity, though this reactor design has disadvantages of its own. Spent fuel from a PWR usually has a higher content of fissile material than natural uranium. Without nuclear reprocessing , this fissile material cannot be used as fuel in

3811-527: 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. (See: Negative temperature coefficient .) Boron and cadmium control rods are used to maintain primary system temperature at

3914-465: Is used as the primary coolant in a PWR. Water enters through the bottom of the reactor's core at about 548  K (275 °C; 527 °F) and is heated as it flows upwards through the reactor core to a temperature of about 588 K (315 °C; 599 °F). The water remains liquid despite the high temperature due to the high pressure in the primary coolant loop, usually around 155 bar (15.5  MPa 153  atm , 2,250  psi ). The water in

4017-471: The Energy Impact Center announced publication of an open-sourced engineering design of a pressurized water reactor capable of producing 300 MWth/100 MWe of energy called OPEN100 . The family of nuclear reactors known as light-water reactors (LWR), cooled and moderated using ordinary water, tend to be simpler and cheaper to build than other types of nuclear reactors; due to these factors, they make up

4120-519: The Idaho National Laboratory ) in a series of tests called the BORAX experiments . PIUS, standing for Process Inherent Ultimate Safety , was a Swedish design designed by ASEA-ATOM. It is a concept for a light-water reactor system. Along with the SECURE reactor, it relied on passive measures, not requiring operator actions or external energy supplies, to provide safe operation. No units were ever built. In 2020,

4223-472: The Oak Ridge National Laboratory for use as a nuclear submarine power plant with a fully operational submarine power plant located at the Idaho National Laboratory . Follow-on work was conducted by Westinghouse Bettis Atomic Power Laboratory . The first purely commercial nuclear power plant at Shippingport Atomic Power Station was originally designed as a pressurized water reactor (although

Light-water reactor - Misplaced Pages Continue

4326-406: The boiling water reactor , the heat generated by fission turns the water into steam, which directly drives the power-generating turbines. But in the pressurized water reactor , the heat generated by fission is transferred to a secondary loop via a heat exchanger. Steam is produced in the secondary loop, and the secondary loop drives the power-generating turbines. In either case, after flowing through

4429-452: The supercritical water reactor , is even less moderated. A less moderated neutron energy spectrum does worsen the capture/fission ratio for U and especially Pu, meaning that more fissile nuclei fail to fission on neutron absorption and instead capture the neutron to become a heavier nonfissile isotope, wasting one or more neutrons and increasing accumulation of heavy transuranic actinides, some of which have long half-lives. After enrichment,

4532-531: 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. The finished fuel rods are grouped in fuel assemblies, called fuel bundles, that are then used to build

4635-535: The 1950s and 1960s at the National Reactor Testing Station in eastern Idaho . They were performed using the five BORAX reactors that were designed and built by Argonne. BORAX-III was the first nuclear reactor to supply electrical power to the grid in the United States in 1955. This series of tests began in 1952 with the construction of the BORAX-I nuclear reactor . BORAX-I experiment proved that

4738-694: The BORAX test facility (500 kW), and partially powered the National Reactor Testing Station (after 2004, the Idaho National Laboratory ) (1,000 kW). Thus, Arco became the first community solely powered by nuclear energy. The reactor continued to be used for tests until 1956. BORAX-IV, built in 1956, explored the thorium fuel cycle and uranium-233 fuel with a power of 20 MW thermal. This experiment used fuel plates that were purposely full of defects to explore long-term plant operation with damaged fuel plates. Radioactive gases were released into

4841-572: The BORAX-I experiments helped scientists understand the fatal meltdown at SL-1 in 1961. The BORAX-II reactor was built in 1954, with a design output of 6 MW(t). In March 1955, BORAX-II was intentionally destroyed by taking the reactor "prompt critical". BORAX-III added a turbine to the BORAX-II design, proving that turbine contamination would not be a problem. It was linked to the local power grid for about an hour on July 17, 1955. BORAX-III provided 2,000 kW to power nearby Arco, Idaho (500 kW),

4944-825: The Koreans currently designing and constructing their second generation of indigenous designs. The leaders in national experience with BWRs, offering reactors for export, are the United States and Japan, with the alliance of General Electric (of the US) and Hitachi (of Japan), offering both the Advanced Boiling Water Reactor (ABWR) and the Economic Simplified Boiling Water Reactor (ESBWR) for construction and export; in addition, Toshiba offers an ABWR variant for construction in Japan, as well. West Germany

5047-538: The Low Intensity Test Reactor (LITR), reached criticality on February 4, 1950 and was the world's first light-water reactor. Immediately after the end of World War II the United States Navy started a program under the direction of Captain (later Admiral) Hyman Rickover , with the goal of nuclear propulsion for ships. It developed the first pressurized water reactors in the early 1950s, and led to

5150-745: The Republic of France (offering the AREVA EPR for export), and Japan (offering the Mitsubishi Advanced Pressurized Water Reactor for export); in addition, both the People's Republic of China and the Republic of Korea are both noted to be rapidly ascending into the front rank of PWR-constructing nations as well, with the Chinese being engaged in a massive program of nuclear power expansion, and

5253-431: The Russian Federation and former Soviet states. Though electricity generation capabilities are comparable between all these types of reactor, due to the aforementioned features, and the extensive experience with operations of the LWR, it is favored in the vast majority of new nuclear power plants. In addition, light-water reactors make up the vast majority of reactors that power naval nuclear-powered vessels . Four out of

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5356-498: The United States are considered Generation II reactors . Russia's VVER reactors are similar to US PWRs, but the VVER-1200 is not considered Generation II (see below). 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

5459-442: The atmosphere. BORAX-V continued the work on boiling water reactor designs, including the use of a superheater . It operated from 1962 to 1964. Test synopsis: The (test was) carried out by withdrawing four of the five control rods far enough to make the reactor critical at a very low power level. The fifth rod was then fired from the core by means of a spring. In this test, the rod was ejected in approximately 0.2 seconds. After

5562-469: The chain reaction to slow down, producing less heat. This property, known as the negative temperature coefficient of reactivity, makes PWRs very stable. In event of a loss-of-coolant accident , the moderator is also lost and the active fission reaction will stop. Heat is still produced after the chain reaction stops from the radioactive byproducts of fission, at about 5% of rated power. This "decay heat" will continue for 1 to 3 years after shut down, whereupon

5665-493: The containment building was a major concern, though the reactors at the plant were BWRs , which owing to the steam at the top of the pressure vessel by design carry a greater risk of this happening. Some reactors contain catalytic recombiners which let the hydrogen react with ambient oxygen in a non-explosive fashion. BORAX experiments The BORAX Experiments were a series of safety experiments on boiling water nuclear reactors conducted by Argonne National Laboratory in

5768-401: The control rod was ejected, an explosion took place in the reactor which carried away the control mechanism and blew out the core. At half a mile, the radiation level rose to 25 mr/hr. Personnel were evacuated for about 30 minutes. The destruction of BORAX-I caused the "aerial distribution of contaminants resulting from the final experiment of the BORAX-I reactor" and the likely contamination of

5871-415: The control rods are lowered into the core, they absorb neutrons, which thus cannot take part in the chain reaction . On the converse, when the control rods are lifted out of the way, more neutrons strike the fissile uranium-235 or plutonium-239 nuclei in nearby fuel rods, and the chain reaction intensifies. All of this is enclosed in a water-filled steel pressure vessel , called the reactor vessel . In

5974-426: 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 tons 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 1,600 MW e . PWR fuel bundles are about 4 meters in length. Refuelings for most commercial PWRs

6077-529: The core to allow a chain reaction to occur. The number of control rods inserted and the distance by which they are inserted can be varied to control the reactivity of the reactor. Usually there are also other means of controlling reactivity. In the PWR design a soluble neutron absorber, usually boric acid , is added to the reactor coolant allowing the complete extraction of the control rods during stationary power operation ensuring an even power and flux distribution over

6180-544: 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. The operator can control

6283-413: The energy released by the fission of atoms. The heated, high pressure water then flows to a steam generator , where it transfers its thermal energy to lower pressure water of a secondary system where steam is generated. The steam then drives turbines, which spin an electric generator. In contrast to a boiling water reactor (BWR), pressure in the primary coolant loop prevents the water from boiling within

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6386-497: The entire core. Operators of the BWR design use the coolant flow through the core to control reactivity by varying the speed of the reactor recirculation pumps. An increase in the coolant flow through the core improves the removal of steam bubbles, thus increasing the density of the coolant/moderator with the result of decreasing power. The light-water reactor also uses ordinary water to keep the reactor cooled. The cooling source, light water,

6489-399: The fast fission neutrons to be slowed (a process called moderation or thermalizing) 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

6592-524: The first power plant connected to the grid was at Obninsk , USSR), on insistence from Admiral Hyman G. Rickover that a viable commercial plant would include none of the "crazy thermodynamic cycles that everyone else wants to build". The United States Army Nuclear Power Program operated pressurized water reactors from 1954 to 1974. Three Mile Island Nuclear Generating Station initially operated two pressurized water reactor plants, TMI-1 and TMI-2. The partial meltdown of TMI-2 in 1979 essentially ended

6695-714: The five great powers with nuclear naval propulsion capacity use light-water reactors exclusively: the British Royal Navy , the Chinese People's Liberation Army Navy , the French Marine nationale , and the United States Navy . Only the Russian Federation's Navy has used a relative handful of liquid-metal cooled reactors in production vessels, specifically the Alfa class submarine , which used lead-bismuth eutectic as

6798-548: The flawed RBMK control rods design. These design flaws, in addition to operator errors that pushed the reactor to its limits, are generally seen as the causes of the Chernobyl disaster . The Canadian CANDU heavy water reactor design have a slight positive void coefficient, these reactors mitigate this issues with a number of built-in advanced passive safety systems not found in the Soviet RBMK design. No criticality could occur in

6901-423: The fuel is similar to PWR fuel except that the bundles are "canned"; that is, there is a thin tube surrounding each bundle. This is primarily done to prevent local density variations from affecting neutronics and thermal hydraulics of the nuclear core on a global scale. In modern BWR fuel bundles, there are either 91, 92, or 96 fuel rods per assembly depending on the manufacturer. A range between 368 assemblies for

7004-511: The growth in new construction of nuclear power plants in the United States for two decades. Watts Bar unit 2 (a Westinghouse 4-loop PWR) came online in 2016, becoming the first new nuclear reactor in the United States since 1996. The pressurized water reactor has several new Generation III reactor evolutionary designs: the AP1000 , VVER-1200, ACPR1000+, APR1400, Hualong One , IPWR-900 and EPR . The first AP1000 and EPR reactors were connected to

7107-460: The heaters or emptying the pressurizer. 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. These pumps have a rate of ~100,000 gallons of coolant per minute. After picking up heat as it passes through

7210-525: The melting of a few fuel plates, the test melted a major fraction of the entire core." Data from this core meltdown and release of nuclear fuel and nuclear fission products helped improve mathematical models. The tests proved key safety principles of the design of modern nuclear power reactors. Design power of BORAX-I was 1.4 megawatts thermal. The BORAX-I design was a precursor to the SL-1 plant, which began operations nearby in 1958. The principles discovered in

7313-510: The most deployed type of reactor globally, allowing for a wide range of suppliers of new plants and parts for existing plants. Due to long experience with their operation they are the closest thing to mature technology that exists in nuclear energy. PWRs - depending on type - can be fueled with MOX-fuel and/or the Russian Remix Fuel (which has a lower Pu and a higher U content than "regular" U/Pu MOX-fuel) allowing for

7416-430: 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

7519-564: The nuclear fuel core of a power reactor. The metal used for the tubes depends on the design of the reactor – stainless steel was used in the past, but most reactors now use a zirconium alloy . For the most common types of reactors the tubes are assembled into bundles with the tubes spaced precise distances apart. These bundles are then given a unique identification number, which enables them to be tracked from manufacture through use and into disposal. Pressurized water reactor fuel consists of cylindrical rods put into bundles. A uranium oxide ceramic

7622-480: The nucleus of a boron-10 atom which subsequently splits into a lithium-7 and tritium atom. Pressurized water reactors annually emit several hundred curies of tritium to the environment as part of normal operation. Natural uranium is only 0.7% uranium-235, the isotope necessary for thermal reactors. This makes it necessary to enrich the uranium fuel, which significantly increases the costs of fuel production. Compared to reactors operating on natural uranium, less energy

7725-513: The portion of a nuclear reactor where the nuclear reactions take place. It mainly consists of nuclear fuel and control elements . The pencil-thin nuclear fuel rods, each about 12 feet (3.7 m) long, are grouped by the hundreds in bundles called fuel assemblies. Inside each fuel rod, pellets of uranium , or more commonly uranium oxide , are stacked end to end. The control elements, called control rods, are filled with pellets of substances like hafnium or cadmium that readily capture neutrons. When

7828-575: The power grid in China in 2018. In 2020, NuScale Power became the first U.S. company to receive regulatory approval from the Nuclear Regulatory Commission for a small modular reactor with a modified PWR design. Also in 2020, the Energy Impact Center introduced the OPEN100 project, which published open-source blueprints for the construction of a 100 MW electric nuclear power plant with

7931-432: 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

8034-518: The pressure vessel must be repaired or replaced. This might not be practical or economic, and so determines the life of the plant. Additional high pressure components such as reactor coolant pumps, pressurizer, and steam generators are also needed. This also increases the capital cost and complexity of a PWR power plant. 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

8137-419: 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

8240-414: The pressurizer temperature and the highest temperature in the reactor core) of 30 °C (54 °F). As 345 °C is the boiling point of water at 155 bar, the liquid water is at the edge of a phase change. Thermal transients in the reactor coolant system result in large swings in pressurizer liquid/steam volume, and total pressurizer volume is designed around absorbing these transients without uncovering

8343-441: 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 155 bars (15.5 MPa), the pressurizer temperature is maintained at 345 °C (653 °F), which gives a subcooling margin (the difference between

8446-423: 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 to radiation exposure. In one instance, 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

8549-406: The primary system. Due to the requirement to load a pressurized water reactor's primary coolant loop with boron, undesirable radioactive secondary tritium production in the water is over 25 times greater than in boiling water reactors of similar power, owing to the latter's absence of the neutron moderating element in its coolant loop. The tritium is created by the absorption of a fast neutron in

8652-432: The reactor coolant and control the reactor power by adjusting the reactor coolant flow rate. 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 turbine cycle loop is separate from the primary loop, so the water in the secondary loop is not contaminated by radioactive materials. PWRs can passively scram

8755-455: The reactor core, the primary coolant transfers 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 returned to the reactor vessel to be heated again. Pressurized water reactors, like all thermal reactor designs, require

8858-434: The reactor finally reaches "full cold shutdown". Decay heat, while dangerous and strong enough to melt the core, is not nearly as intense as an active fission reaction. During the post shutdown period the reactor requires cooling water to be pumped or the reactor will overheat. If the temperature exceeds 2200 °C, cooling water will break down into hydrogen and oxygen, which can form a (chemically) explosive mixture. Decay heat

8961-403: The reactor in case offsite power is lost to immediately stop the primary nuclear reaction. The control rods are held by electromagnets and fall by gravity when current is lost; full insertion safely shuts down the primary nuclear reaction. PWR technology is favoured by nations seeking to develop a nuclear navy; the compact reactors fit well in nuclear submarines and other nuclear ships. PWRs are

9064-428: The reactor, to shut down the primary nuclear reactions in the reactor, to accommodate short term transients, such as changes to load on the turbine, The control rods can also be used to compensate for nuclear poison inventory and to compensate for nuclear fuel depletion. However, these effects are more usually accommodated by altering the primary coolant boric acid concentration. In contrast, BWRs have no boron in

9167-459: The reactor. All light-water reactors use ordinary water as both coolant and neutron moderator . Most use anywhere from two to four vertically mounted steam generators; VVER reactors use horizontal steam generators. 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 . PWRs currently operating in

9270-431: 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.e. the hotter the coolant becomes, the less reactive the plant becomes, shutting itself down slightly to compensate and vice versa. Thus the plant controls itself around

9373-401: The smallest and 800 assemblies for the largest U.S. BWR forms the reactor core. Each BWR fuel rod is back filled with helium to a pressure of about three atmospheres (300 kPa). A neutron moderator is a medium which reduces the velocity of fast neutrons , thereby turning them into thermal neutrons capable of sustaining a nuclear chain reaction involving uranium-235. A good neutron moderator is

9476-426: The steady state operating temperature by addition of boric acid and/or movement of control rods. 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

9579-427: The steam system and pressure inside the primary coolant loop. 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 coolant and moderator (e.g. sodium in its liquid state as coolant or graphite as

9682-623: The successful deployment of the first nuclear submarine, the USS ; Nautilus  (SSN-571) . The Soviet Union independently developed a version of the PWR in the late 1950s, under the name of VVER . While functionally very similar to the American effort, it also has certain design distinctions from Western PWRs. Researcher Samuel Untermyer II led the effort to develop the BWR at the US National Reactor Testing Station (now

9785-482: The topmost 1 foot of soil over about 2 acres in the vicinity. The site required cleanup before it could be used for subsequent experiments. The 84,000-square foot (7,800 m ) area was covered with 6 inches of gravel in 1954, but grass, sagebrush, and other plants reseeded the area since then. Debris from BORAX-I is buried about 2,730 feet (830 m) northwest of the Experimental Breeder Reactor-1 ,

9888-551: The turbines, the steam turns back into water in the condenser. The water required to cool the condenser is taken from a nearby river or ocean. It is then pumped back into the river or ocean, in warmed condition. The heat can also be dissipated via a cooling tower into the atmosphere. The United States uses LWR reactors for electric power production, in comparison to the heavy water reactors used in Canada. Control rods are usually combined into control rod assemblies — typically 20 rods for

9991-477: The vast majority of civil nuclear reactors and naval propulsion reactors in service throughout the world as of 2009. LWRs can be subdivided into three categories – pressurized water reactors (PWRs), boiling water reactors (BWRs), and supercritical water reactors ( SCWRs ). The SCWR remains hypothetical as of 2009; it is a Generation IV design that is still a light-water reactor, but it is only partially moderated by light water and exhibits certain characteristics of

10094-409: The void coefficient is positive, and fairly large, making it very hard to regulate when the reaction begins to run away. The RBMK reactors also have a flawed control rods design in which during rapid scrams, the graphite reaction enhancement tips of the rods would displace water at the bottom of the reactor and locally increase reactivity there. This is called the "positive scram effect" that is unique to

10197-414: The walls of these tubes to the lower pressure secondary coolant located on the shell side of the exchanger where the secondary coolant evaporates to pressurized steam. This 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,

10300-419: The water is denser, because 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

10403-534: Was also once a major player with BWRs. The other types of nuclear reactor in use for power generation are the heavy water moderated reactor , built by Canada ( CANDU ) and the Republic of India (AHWR), the advanced gas cooled reactor (AGCR), built by the United Kingdom, the liquid metal cooled reactor (LMFBR), built by the Russian Federation, the Republic of France, and Japan, and the graphite-moderated, water-cooled reactor (RBMK or LWGR), found exclusively within

10506-627: Was proposed for a reactor whose purpose was to test the behavior of materials under neutron flux . This reactor, the Material Testing Reactor (MTR) , was built in Idaho at INL and reached criticality on March 31, 1952. For the design of this reactor, experiments were necessary, so a mock-up of the MTR was built at ORNL , to assess the hydraulic performances of the primary circuit and then to test its neutronic characteristics. This MTR mock-up, later called

10609-419: Was used to estimate the critical mass of U235 to produce the atomic bomb. LOPO cannot be considered as the first light-water reactor because its fuel was not a solid uranium compound cladded with corrosion-resistant material, but was composed of uranyl sulfate salt dissolved in water. It is however the first aqueous homogeneous reactor and the first reactor using enriched uranium as fuel and ordinary water as

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