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132-598: It consists of two Pressurized water reactors (PWRs) of 1011 and 1006 MWe. The refrigeration of the Almaraz nuclear power plant was the first reason for the construction of the Arrocampo Reservoir in 1976. The water is taken from the Tagus River and covers a U-shaped circuit of 25 km which allows the cooling of the heat generated by the two nuclear reactors of the plant. (See the illustration of

264-449: A LBLOCA (large-break loss-of-coolant accident – a massive pipe rupture leading to catastrophic loss of coolant pressure within the reactor, considered the most threatening "design basis accident" in probabilistic risk assessment and nuclear safety and security ), which is anticipated to lead to the temporary exposure of the core; this core drying-out event is termed core "uncovery", for the core loses its heat-removing cover of coolant, in

396-545: A boiling water reactor would be feasible for use in energy production. He found that it was, after subjecting his reactors to quite strenuous tests, proving the safety principles of the BWR. Following this series of tests, GE got involved and collaborated with Argonne National Laboratory to bring this technology to market. Larger-scale tests were conducted through the late 1950s/early/mid-1960s that only partially used directly generated (primary) nuclear boiler system steam to feed

528-446: A reactor core , holding up to approximately 140 short tons of low-enriched uranium . The number of fuel assemblies in a specific reactor is based on considerations of desired reactor power output, reactor core size and reactor power density. A modern reactor has many safety systems that are designed with a defence in depth philosophy, which is a design philosophy that is integrated throughout construction and commissioning . A BWR

660-442: 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

792-409: A BWR core is substantiated by a calculation that proves that 99.9% of fuel rods in a BWR core will not enter the transition to film boiling during normal operation or anticipated operational occurrences. Since the BWR is boiling water, and steam does not transfer heat as well as liquid water, MFLCPR typically occurs at the top of a fuel assembly, where steam volume is the highest. FLLHGR (FDLRX, MFLPD)

924-524: A BWR: MFLCPR, FLLHGR, and APLHGR must be kept less than 1.0 during normal operation; administrative controls are in place to assure some margin of error and margin of safety to these licensed limits. Typical computer simulations divide the reactor core into 24–25 axial planes ; relevant quantities (margins, burnup, power, void history) are tracked for each "node" in the reactor core (764 fuel assemblies x 25 nodes/assembly = 19100 nodal calculations/quantity). Specifically, MFLCPR represents how close

1056-479: 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

1188-465: 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

1320-405: 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

1452-475: 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

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1584-459: 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

1716-656: 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

1848-444: 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

1980-632: A high power output (1350 MWe per reactor), and a significantly lowered probability of core damage. Most significantly, the ABWR was a completely standardized design, that could be made for series production. The ABWR was approved by the United States Nuclear Regulatory Commission for production as a standardized design in the early 1990s. Subsequently, numerous ABWRs were built in Japan. One development spurred by

2112-497: A lower pressure system, which turns water into steam that drives the turbine. The BWR was developed by the Argonne National Laboratory and General Electric (GE) in the mid-1950s. The main present manufacturer is GE Hitachi Nuclear Energy , which specializes in the design and construction of this type of reactor. A boiling water reactor uses demineralized water as a coolant and neutron moderator . Heat

2244-601: A meeting between Portuguese and Spanish delegates in Madrid, which ended in deadlock and Portugal to complain to the EU that Spain ignored the potential cross-border impact with no studies being carried out, which is against European Union rules. Spanish secretary of State for the EU Jorge Toledo Albiñana has said work will start regardless of Portugal's complaints, and uranium bars that will remain radioactive for

2376-402: A model of the fuel assembly but power it with resistive heaters. These mock fuel assemblies are put into a test stand where data points are taken at specific powers, flows, pressures. Experimental data is conservatively applied to BWR fuel to ensure that the transition to film boiling does not occur during normal or transient operation. Typical SLMCPR/MCPRSL (Safety Limit MCPR) licensing limit for

2508-426: 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

2640-423: 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,

2772-562: A pressurized water reactor (although 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

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2904-542: 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

3036-469: 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,

3168-410: A safety-related contingency developed. For example, if the reactor got too hot, it would trigger a system that would release soluble neutron absorbers (generally a solution of borated materials, or a solution of borax ), or materials that greatly hamper a chain reaction by absorbing neutrons, into the reactor core. The tank containing the soluble neutron absorbers would be located above the reactor, and

3300-567: 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 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

3432-402: A series of notched positions with fixed intervals between these positions. Due to the limitations of the manual control system, it is possible while starting-up that the core can be placed into a condition where movement of a single control rod can cause a large nonlinear reactivity change, which could heat fuel elements to the point they fail (melt, ignite, weaken, etc.). As a result, GE developed

3564-415: A set of rules in 1977 called BPWS (Banked Position Withdrawal Sequence) which help minimize the effect of any single control rod movement and prevent fuel damage in the case of a control rod drop accident. BPWS separates control rods into four groups, A1, A2, B1, and B2. Then, either all of the A control rods or B control rods are pulled full out in a defined sequence to create a " checkerboard " pattern. Next,

3696-420: 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

3828-514: A single core-damaging event during their 100-year lifetimes. Earlier designs of the BWR, the BWR/4, had core damage probabilities as high as 1×10 core-damage events per reactor-year. This extraordinarily low CDP for the ESBWR far exceeds the other large LWRs on the market. Reactor start up ( criticality ) is achieved by withdrawing control rods from the core to raise core reactivity to a level where it

3960-468: A small group of engineers accidentally increased the reactor power level on an experimental reactor to such an extent that the water quickly boiled. This shut down the reactor, indicating the useful self-moderating property in emergency circumstances. In particular, Samuel Untermyer II , a researcher at Argonne National Laboratory , proposed and oversaw a series of experiments: the BORAX experiments —to see if

4092-554: A tall shroud. The water then goes through either jet pumps or internal recirculation pumps that provide additional pumping power (hydraulic head). The water now makes a 180-degree turn and moves up through the lower core plate into the nuclear core, where the fuel elements heat the water. Water exiting the fuel channels at the top guide is saturated with a steam quality of about 15%. Typical core flow may be 45,000,000 kg/h (100,000,000 lb/h) with 6,500,000 kg/h (14,500,000 lb/h) steam flow. However, core-average void fraction

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4224-401: Is a limit on fuel rod power in the reactor core. For new fuel, this limit is typically around 13 kW/ft (43 kW/m) of fuel rod. This limit ensures that the centerline temperature of the fuel pellets in the rods will not exceed the melting point of the fuel material ( uranium / gadolinium oxides) in the event of the worst possible plant transient/scram anticipated to occur. To illustrate

4356-454: Is a significantly higher fraction (~40%). These sort of values may be found in each plant's publicly available Technical Specifications, Final Safety Analysis Report, or Core Operating Limits Report. The heating from the core creates a thermal head that assists the recirculation pumps in recirculating the water inside of the RPV. A BWR can be designed with no recirculation pumps and rely entirely on

4488-517: 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 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

4620-447: Is evident that the nuclear chain reaction is self-sustaining. This is known as "going critical". Control rod withdrawal is performed slowly, as to carefully monitor core conditions as the reactor approaches criticality. When the reactor is observed to become slightly super-critical, that is, reactor power is increasing on its own, the reactor is declared critical. Rod motion is performed using rod drive control systems. Newer BWRs such as

4752-402: 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

4884-569: Is in place to ensure that the highest powered fuel rod will not melt if its power was rapidly increased following a pressurization transient. Abiding by the LHGR limit precludes melting of fuel in a pressurization transient. APLHGR, being an average of the Linear Heat Generation Rate (LHGR), a measure of the decay heat present in the fuel bundles, is a margin of safety associated with the potential for fuel failure to occur during

5016-423: Is known as the advanced boiling water reactor (ABWR). The ABWR was developed in the late 1980s and early 1990s, and has been further improved to the present day. The ABWR incorporates advanced technologies in the design, including computer control, plant automation, control rod removal, motion, and insertion, in-core pumping, and nuclear safety to deliver improvements over the original series of production BWRs, with

5148-433: Is licensed to operate, the fuel vendor/licensee simulate events with computer models. Their approach is to simulate worst case events when the reactor is in its most vulnerable state. APLHGR is commonly pronounced as "Apple Hugger" in the industry. PCIOMR is a set of rules and limits to prevent cladding damage due to pellet-clad interaction. During the first nuclear heatup, nuclear fuel pellets can crack. The jagged edges of

5280-408: Is monitored with an empirical correlation that is formulated by vendors of BWR fuel (GE, Westinghouse, AREVA-NP). The vendors have test rigs where they simulate nuclear heat with resistive heating and determine experimentally what conditions of coolant flow, fuel assembly power, and reactor pressure will be in/out of the transition boiling region for a particular fuel design. In essence, the vendors make

5412-432: 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,

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5544-401: 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

5676-527: 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

5808-407: Is produced by nuclear fission in the reactor core, and this causes the cooling water to boil, producing steam. The steam is directly used to drive a turbine , after which it is cooled in a condenser and converted back to liquid water. This water is then returned to the reactor core, completing the loop. The cooling water is maintained at about 75 atm (7.6 MPa , 1000–1100 psi ) so that it boils in

5940-453: Is required that the decay heat stored in the fuel assemblies at any one time does not overwhelm the ECCS. As such, the measure of decay heat generation known as LHGR was developed by GE's engineers, and from this measure, APLHGR is derived. APLHGR is monitored to ensure that the reactor is not operated at an average power level that would defeat the primary containment systems. When a refueled core

6072-484: Is similar to a pressurized water reactor (PWR) in that the reactor will continue to produce heat even after the fission reactions have stopped, which could make a core damage incident possible. This heat is produced by the radioactive decay of fission products and materials that have been activated by neutron absorption . BWRs contain multiple safety systems for cooling the core after emergency shut down. The reactor fuel rods are occasionally replaced by moving them from

6204-492: Is terminated by the automatic insertion of the control rods. So, when the reactor is isolated from the turbine rapidly, pressure in the vessel rises rapidly, which collapses the water vapor, which causes a power excursion which is terminated by the Reactor Protection System. If a fuel pin was operating at 13.0 kW/ft prior to the transient, the void collapse would cause its power to rise. The FLLHGR limit

6336-465: 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

6468-554: The ABWR and ESBWR as well as all German and Swedish BWRs use the Fine Motion Control Rod Drive system, which allows multiple rods to be controlled with very smooth motions. This allows a reactor operator to evenly increase the core's reactivity until the reactor is critical. Older BWR designs use a manual control system, which is usually limited to controlling one or four control rods at a time, and only through

6600-565: The Emergency Core Cooling System . The ECCS is designed to rapidly flood the reactor pressure vessel, spray water on the core itself, and sufficiently cool the reactor fuel in this event. However, like any system, the ECCS has limits, in this case, to its cooling capacity, and there is a possibility that fuel could be designed that produces so much decay heat that the ECCS would be overwhelmed and could not cool it down successfully. So as to prevent this from happening, it

6732-452: 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,

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6864-420: The turbine flows into condensers located underneath the low-pressure turbines, where the steam is cooled and returned to the liquid state (condensate). The condensate is then pumped through feedwater heaters that raise its temperature using extraction steam from various turbine stages. Feedwater from the feedwater heaters enters the reactor pressure vessel (RPV) through nozzles high on the vessel, well above

6996-531: 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

7128-531: The US's first research effort in nuclear power being devoted to the PWR, which was highly suited for naval vessels (submarines, especially), as space was at a premium, and PWRs could be made compact and high-power enough to fit into such vessels. But other researchers wanted to investigate whether the supposed instability caused by boiling water in a reactor core would really cause instability. During early reactor development,

7260-565: The US, they were originally designed at 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

7392-499: The absorption solution, once the system was triggered, would flow into the core through force of gravity, and bring the reaction to a near-complete stop. Another example was the Isolation Condenser system , which relied on the principle of hot water/steam rising to bring hot coolant into large heat exchangers located above the reactor in very deep tanks of water, thus accomplishing residual heat removal. Yet another example

7524-436: The case of a BWR, light water. If the core is uncovered for too long, fuel failure can occur; for the purpose of design, fuel failure is assumed to occur when the temperature of the uncovered fuel reaches a critical temperature (1100 °C, 2200 °F). BWR designs incorporate failsafe protection systems to rapidly cool and make safe the uncovered fuel prior to it reaching this temperature; these failsafe systems are known as

7656-443: 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. Boiling water reactor A boiling water reactor ( BWR ) is a type of light water nuclear reactor used for

7788-419: The core at about 285 °C (550 °F). In comparison, there is no significant boiling allowed in a pressurized water reactor (PWR) because of the high pressure maintained in its primary loop—approximately 158 atm (16 MPa, 2300 psi). The core damage frequency of the reactor was estimated to be between 10 and 10 (i.e., one core damage accident per every 10,000 to 10,000,000 reactor years). Steam exiting

7920-430: The core enters the riser area, which is the upper region contained inside of the shroud. The height of this region may be increased to increase the thermal natural recirculation pumping head. At the top of the riser area is the moisture separator. By swirling the two-phase flow in cyclone separators, the steam is separated and rises upwards towards the steam dryer while the water remains behind and flows horizontally out into

8052-466: The core is increased, steam bubbles ("voids") are more quickly removed from the core, the amount of liquid water in the core increases, neutron moderation increases, more neutrons are slowed to be absorbed by the fuel, and reactor power increases. As flow of water through the core is decreased, steam voids remain longer in the core, the amount of liquid water in the core decreases, neutron moderation decreases, fewer neutrons are slowed enough to be absorbed by

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8184-426: 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

8316-536: The core water level. If all feedwater is lost, the reactor will scram and the Emergency Core Cooling System is used to restore reactor water level. Steam produced in the reactor core passes through steam separators and dryer plates above the core and then directly to the turbine , which is part of the reactor circuit. Because the water around the core of a reactor is always contaminated with traces of radionuclides due to neutron capture from

8448-544: 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

8580-432: The downcomer or annulus region. In the downcomer or annulus region, it combines with the feedwater flow and the cycle repeats. The saturated steam that rises above the separator is dried by a chevron dryer structure. The "wet" steam goes through a tortuous path where the water droplets are slowed and directed out into the downcomer or annulus region. The "dry" steam then exits the RPV through four main steam lines and goes to

8712-605: The environment in Spain and in neighboring Portugal. Greenpeace has labelled the plant as an "extreme case" in its study on the application of minimum safety standards introduced in Europe after the Fukushima accident. On the 21 September 2016, defective parts were used on unit 1's second and third steam generators and on unit 2's third steam generator, as well as the rim of the reactor lid in unit 2. It had already been reported that

8844-532: The event of a major safety contingency for at least 48 hours following the safety contingency; thence, it would only require periodic refilling of cooling water tanks located completely outside of the reactor, isolated from the cooling system, and designed to remove reactor waste heat through evaporation. The simplified boiling water reactor was submitted to the United States Nuclear Regulatory Commission , however, it

8976-428: The event of a transient requiring the quenching of steam), as well as the drywell, the elimination of the heat exchanger, the steam dryer, the distinctive general layout of the reactor building, and the standardization of reactor control and safety systems. The first, General Electric ( GE ), series of production BWRs evolved through 6 iterative design phases, each termed BWR/1 through BWR/6. (BWR/4s, BWR/5s, and BWR/6s are

9108-399: 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

9240-418: The feed water control system can rapidly anticipate water level deviations and respond to maintain water level within a few inches of set point. If one of the two feedwater pumps fails during operation, the feedwater system will command the recirculation system to rapidly reduce core flow, effectively reducing reactor power from 100% to 50% in a few seconds. At this power level a single feedwater pump can maintain

9372-559: 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

9504-411: The flow of water through the core is the normal and convenient method for controlling power from approximately 30% to 100% reactor power. When operating on the so-called "100% rod line", power may be varied from approximately 30% to 100% of rated power by changing the reactor recirculation system flow by varying the speed of the recirculation pumps or modulating flow control valves. As flow of water through

9636-442: The fuel, and reactor power decreases. Thus the BWR has a negative void coefficient . Reactor pressure in a BWR is controlled by the main turbine or main steam bypass valves. Unlike a PWR, where the turbine steam demand is set manually by the operators, in a BWR, the turbine valves will modulate to maintain reactor pressure at a setpoint. Under this control mode, the turbine output will automatically follow reactor power changes. When

9768-414: The fuel, so reactor power increases. As control rods are inserted, neutron absorption increases in the control material and decreases in the fuel, so reactor power decreases. Differently from the PWR, in a BWR the control rods ( boron carbide plates) are inserted from below to give a more homogeneous distribution of the power: in the upper side the density of the water is lower due to vapour formation, making

9900-458: The generation of electrical power. It is the second most common type of electricity-generating nuclear reactor after the pressurized water reactor (PWR), which is also a type of light water nuclear reactor. The main difference between a BWR and PWR is that in a BWR, the reactor core heats water, which turns to steam and then drives a steam turbine. In a PWR, the reactor core heats water, which does not boil. This hot water then exchanges heat with

10032-410: The geometry. The fact that the fuel rods' cladding is a zirconium alloy was also problematic since this element can react with steam at temperatures above 1,500 K (1,230 °C) to produce hydrogen, which can ignite with oxygen in the air. Normally the fuel rods are kept sufficiently cool in the reactor and spent fuel pools that this is not a concern, and the cladding remains intact for the life of

10164-578: 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

10296-405: The heated surface to increase drastically to once again reach equilibrium heat transfer with the cooling fluid. In other words, steam semi-insulates the heated surface and surface temperature rises to allow heat to get to the cooling fluid (through convection and radiative heat transfer). Nuclear fuel could be damaged by film boiling; this would cause the fuel cladding to overheat and fail. MFLCPR

10428-460: 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

10560-405: The inner walls of the fuel cladding which are resistant to perforation due to pellet-clad interactions, and the second is a set of rules created under PCIOMR. The PCIOMR rules require initial "conditioning" of new fuel. This means, for the first nuclear heatup of each fuel element, that local bundle power must be ramped very slowly to prevent cracking of the fuel pellets and limit the differences in

10692-400: The leading fuel bundle is to "dry-out" (or "departure from nucleate boiling" for a PWR). Transition boiling is the unstable transient region where nucleate boiling tends toward film boiling . A water drop dancing on a hot frying pan is an example of film boiling. During film boiling a volume of insulating vapor separates the heated surface from the cooling fluid; this causes the temperature of

10824-508: The low-pressure turbines to use. The exhaust of the low-pressure turbines is sent to the main condenser. The steam reheaters take some of the turbine's steam and use it as a heating source to reheat what comes out of the high-pressure turbine exhaust. While the reheaters take steam away from the turbine, the net result is that the reheaters improve the thermodynamic efficiency of the plant. A modern BWR fuel assembly comprises 74 to 100 fuel rods , and there are up to approximately 800 assemblies in

10956-571: The most common types in service today.) The vast majority of BWRs in service throughout the world belong to one of these design phases. Containment variants were constructed using either concrete or steel for the Primary Containment, Drywell and Wetwell in various combinations. Apart from the GE designs there were others by ABB (Asea-Atom), MITSU, Toshiba and KWU (Kraftwerk Union). See List of boiling water reactors . A newer design of BWR

11088-518: 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

11220-430: 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

11352-591: The neutron moderation less efficient and the fission probability lower. In normal operation, the control rods are only used to keep a homogeneous power distribution in the reactor and to compensate for the consumption of the fuel, while the power is controlled through the water flow (see below). Some early BWRs and the proposed ESBWR (Economic Simplified BWR made by General Electric Hitachi) designs use only natural circulation with control rod positioning to control power from zero to 100% because they do not have reactor recirculation systems. Changing (increasing or decreasing)

11484-695: The next 300 years will be stored on site. In May 2017 the Portuguese Parliament approved the Ecologist Party "The Greens" motion to request the closure of Spain's Almaraz nuclear plant during the next Iberian summit. Stating that after 2020 the plant should be shut down and the Greens asked the government to take a "resolute position", for the facility, located 100 kilometres from the Portuguese border. Environmentalists have warned that

11616-480: 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

11748-469: The opposing group (B or A) is pulled in a defined sequence to positions 02, then 04, 08, 16, and finally full out (48). By following a BPWS compliant start-up sequence, the manual control system can be used to evenly and safely raise the entire core to critical, and prevent any fuel rods from exceeding 280 cal/gm energy release during any postulated event which could potentially damage the fuel. Several calculated/measured quantities are tracked while operating

11880-544: The original design of the second commercial power plant at Shippingport Atomic Power Station . PWRs currently operating in 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

12012-404: The pellet can rub and interact with the inner cladding wall. During power increases in the fuel pellet, the ceramic fuel material expands faster than the fuel cladding, and the jagged edges of the fuel pellet begin to press into the cladding, potentially causing a perforation. To prevent this from occurring, two corrective actions were taken. The first is the inclusion of a thin barrier layer against

12144-517: The plan to build a nuclear waste warehouse site next to the power plant almost certainly indicates that Spain plans to extend the life of the Almaraz power plant beyond the year 2020. On the 28 January 2016, the Spanish Nuclear Safety Council inspectors found serious failings in the water pump engines at the plant, which have potential operational issues of the cooling system and could pose a serious risk to local people and

12276-667: 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

12408-432: 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

12540-518: 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

12672-419: 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

12804-414: 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

12936-441: 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

13068-423: 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

13200-406: 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

13332-479: The pumps could be repaired during the next refueling outage. Instead, the designers of the simplified boiling water reactor used thermal analysis to design the reactor core such that natural circulation (cold water falls, hot water rises) would bring water to the center of the core to be boiled. The ultimate result of the passive safety features of the SBWR would be a reactor that would not require human intervention in

13464-510: The rates of thermal expansion of the fuel. PCIOMR rules also limit the maximum local power change (in kW/ft*hr), prevent pulling control rods below the tips of adjacent control rods, and require control rod sequences to be analyzed against core modelling software to prevent pellet-clad interactions. PCIOMR analysis look at local power peaks and xenon transients which could be caused by control rod position changes or rapid power changes to ensure that local power rates never exceed maximum ratings. For

13596-432: 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

13728-455: 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

13860-403: 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

13992-423: The reactor pressure vessel to the spent fuel pool. A typical fuel cycle lasts 18–24 months, with about one third of fuel assemblies being replaced during a refueling outage. The remaining fuel assemblies are shuffled to new core locations to maximize the efficiency and power produced in the next fuel cycle. Because they are hot both radioactively and thermally, this is done via cranes and under water. For this reason

14124-428: 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

14256-431: 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

14388-479: The response of LHGR in transient imagine the rapid closure of the valves that admit steam to the turbines at full power. This causes the immediate cessation of steam flow and an immediate rise in BWR pressure. This rise in pressure effectively subcools the reactor coolant instantaneously; the voids (vapor) collapse into solid water. When the voids collapse in the reactor, the fission reaction is encouraged (more thermal neutrons); power increases drastically (120%) until it

14520-755: The resulting design to a larger size of 1,600  MWe (4,500 MWth). This Economic Simplified Boiling Water Reactor (ESBWR) design was submitted to the US Nuclear Regulatory Commission for approval in April 2005, and design certification was granted by the NRC in September 2014. Reportedly, this design has been advertised as having a core damage probability of only 3×10 core damage events per reactor-year. That is, there would need to be 3 million ESBWRs operating before one would expect

14652-427: The rod. The BWR concept was developed slightly later than the PWR concept. Development of the BWR started in the early 1950s, and was a collaboration between General Electric (GE) and several US national laboratories. Research into nuclear power in the US was led by the three military services. The Navy, seeing the possibility of turning submarines into full-time underwater vehicles, and ships that could steam around

14784-485: The spent fuel storage pools are above the reactor in typical installations. They are shielded by water several times their height, and stored in rigid arrays in which their geometry is controlled to avoid criticality. In the Fukushima Daiichi nuclear disaster this became problematic because water was lost (as it was heated by the spent fuel) from one or more spent fuel pools and the earthquake could have altered

14916-426: 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

15048-427: 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

15180-440: The success of the ABWR in Japan is that General Electric's nuclear energy division merged with Hitachi Corporation's nuclear energy division, forming GE Hitachi Nuclear Energy , which is now the major worldwide developer of the BWR design. Parallel to the development of the ABWR, General Electric also developed a different concept, known as the simplified boiling water reactor (SBWR). This smaller 600 megawatt electrical reactor

15312-429: The thermal head to recirculate the water inside of the RPV. The forced recirculation head from the recirculation pumps is very useful in controlling power, however, and allows achieving higher power levels that would not otherwise be possible. The thermal power level is easily varied by simply increasing or decreasing the forced recirculation flow through the recirculation pumps. The two-phase fluid (water and steam) above

15444-406: The top of the nuclear fuel assemblies (these nuclear fuel assemblies constitute the "core") but below the water level. The feedwater enters into the downcomer or annulus region and combines with water exiting the moisture separators. The feedwater subcools the saturated water from the moisture separators. This water now flows down the downcomer or annulus region, which is separated from the core by

15576-461: The turbine and incorporated heat exchangers for the generation of secondary steam to drive separate parts of the turbines. The literature does not indicate why this was the case, but it was eliminated on production models of the BWR. The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR: the torus (used to quench steam in

15708-403: The turbine hall can be entered soon after the reactor is shut down. BWR steam turbines employ a high-pressure turbine designed to handle saturated steam, and multiple low-pressure turbines. The high-pressure turbine receives steam directly from the reactor. The high-pressure turbine exhaust passes through a steam reheater which superheats the steam to over 400 degrees F (204.4 degrees celcius) for

15840-450: The turbine is offline or trips, the main steam bypass/dump valves will open to direct steam directly to the condenser. These bypass valves will automatically or manually modulate as necessary to maintain reactor pressure and control the reactor's heatup and cooldown rates while steaming is still in progress. Reactor water level is controlled by the main feedwater system. From about 0.5% power to 100% power, feedwater will automatically control

15972-400: The turbine. Reactor power is controlled via two methods: by inserting or withdrawing control rods (control blades) and by changing the water flow through the reactor core . Positioning (withdrawing or inserting) control rods is the normal method for controlling power when starting up a BWR. As control rods are withdrawn, neutron absorption decreases in the control material and increases in

16104-409: 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

16236-414: 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,

16368-642: The water circulation in Arrocampo) The walls of thermic separation ( pantallas de separación térmica in Spanish) (PST) are 11 km long and 8 m high. The tops of these walls are used by great cormorants and great egret as standing, resting and sleeping areas. In 1975 Luis E. Echávarri was made project manager of the plant. In 1985 he became technical director of the Spanish Nuclear Safety Council (CSN), and in 1987 he

16500-437: The water level in the reactor. At low power conditions, the feedwater controller acts as a simple PID control by watching reactor water level. At high power conditions, the controller is switched to a "Three-Element" control mode, where the controller looks at the current water level in the reactor, as well as the amount of water going in and the amount of steam leaving the reactor. By using the water injection and steam flow rates,

16632-399: The water pump engines had been stopped twice, the plant's cooling system was reported as not 100% reliable. On the 10 April 2017, the Spanish Nuclear Safety Council issued a statement stating there was an unscheduled stoppage of the main number two pump at 9:57   am Spanish time (8:57   am Lisbon time). Pressurized water reactor A pressurized water reactor ( PWR )

16764-423: The water, the turbine must be shielded during normal operation, and radiological protection must be provided during maintenance. The increased cost related to operation and maintenance of a BWR tends to balance the savings due to the simpler design and greater thermal efficiency of a BWR when compared with a PWR. Most of the radioactivity in the water is very short-lived (mostly N-16, with a 7-second half-life ), so

16896-546: The world without refueling, sent their man in engineering, Captain Hyman Rickover to run their nuclear power program. Rickover decided on the PWR route for the Navy, as the early researchers in the field of nuclear power feared that the direct production of steam within a reactor would cause instability, while they knew that the use of pressurized water would definitively work as a means of heat transfer. This concern led to

17028-530: Was named Commissioner of the CSN. The first reactor began operating in 1981 and the second in 1983. It occupies an area of 1683 hectares. As of 2017 Spain had approved a nuclear waste warehouse at Almaraz without carrying out any consultations or impact studies. Portugal has taken the matter to the EU , protests planned on 12 January at Spanish consulates were organised by Movimiento Ibérico Antinuclear, which coincided with

17160-407: Was notable for its incorporation—for the first time ever in a light water reactor —of " passive safety " design principles. The concept of passive safety means that the reactor, rather than requiring the intervention of active systems, such as emergency injection pumps, to keep the reactor within safety margins, was instead designed to return to a safe state solely through operation of natural forces if

17292-439: Was the omission of recirculation pumps within the core; these pumps were used in other BWR designs to keep cooling water moving; they were expensive, hard to reach to repair, and could occasionally fail; so as to improve reliability, the ABWR incorporated no less than 10 of these recirculation pumps, so that even if several failed, a sufficient number would remain serviceable so that an unscheduled shutdown would not be necessary, and

17424-399: Was withdrawn prior to approval; still, the concept remained intriguing to General Electric's designers, and served as the basis of future developments. During a period beginning in the late 1990s, GE engineers proposed to combine the features of the advanced boiling water reactor design with the distinctive safety features of the simplified boiling water reactor design, along with scaling up

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