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94-468: The Gentilly reactors were constructed in stages between 1966 and 1983 and were originally part of a plan for 30-35 nuclear reactors in Quebec. A third reactor, Gentilly-3, was scheduled to be built on the same site but was cancelled because of a drop in demand growth in the late 1970s. In October 2012, it was decided for economic reasons not to proceed with the refurbishment of Gentilly-2, and to decommission

188-449: 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

282-545: 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

376-416: A decommissioning process will proceed over a period of 50 years and is expected to cost $ 1.8 billion. The permanent shut down and decommissioning of the power plant followed an election pledge from Quebec premier Pauline Marois . The Gentilly site also houses a 411MW gas turbine generation plant. The Bécancour generating station was commissioned in 1992-1993. Gentilly-3 was a proposed nuclear reactor at

470-446: 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

564-409: 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)

658-524: 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

752-565: A North American grid-scale SMR. The Advanced Boiling Water Reactor (ABWR) is the world's first operational Generation III Class advanced light water reactor design. The NRC has registered GEH's petition for renewal of ABWR certification. The Economic Simplified Boiling Water Reactor (ESBWR), the Generation III+ Class design reactor, received a positive final safety evaluation report and final design approval in March 2011, and

846-556: A good service record since start-up in 1982, with a cumulative operating factor of 76.4%. In an August 19, 2008 announcement, Québec planned to spend $ 1.9B to overhaul Gentilly-2 in order to extend its lifespan to 2040. Refurbishment of the reactor was eventually cancelled when on 3 October 2012, Hydro-Quebec's CEO, Thierry Vandal, announced the decommissioning of the Gentilly-2 generating station for economic reasons, scheduled to occur on 28 December 2012 at 10:30 p.m. At that time,

940-632: 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

1034-536: A list of operational and decommissioned BWRs, see List of BWRs . Experimental and other non-commercial BWRs include: GE Hitachi Nuclear Energy GE Hitachi Nuclear Energy ( GEH ) is a provider of advanced reactors and nuclear services. It is headquartered in Wilmington, North Carolina , United States. Established in June 2007, GEH is a nuclear alliance created by General Electric and Hitachi . In Japan,

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1128-497: 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

1222-402: 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

1316-459: A positive Safety Evaluation Report and Final Design Approval on March 9, 2011. On June 7, 2011, the NRC completed its public comment period. Final rule was issued on September 16, 2014, after two outstanding problems with GE-Hitachi's modeling of loads on the steam dryer were solved. In 2013, following its purchase of Horizon Nuclear Power , Hitachi began the process of generic design assessment of

1410-410: 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

1504-402: 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

1598-415: 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,

1692-514: 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

1786-468: 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

1880-554: 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

1974-401: 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

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2068-454: 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

2162-447: 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

2256-778: Is expected to receive a license from the NRC by September 2011. GEH's Power Reactor Innovative Small Modular (PRISM) is a Generation IV reactor that uses liquid sodium as a coolant. In 2020 GEH partnered with TerraPower to develop a Natrium reactor. In 2018, GEH agreed to collaborate with Holtec International on the commercialization of the Holtec SMR-160, a 160 MWe pressurized water reactor (PWR) small modular reactor . As nuclear plants get older and worldwide demand for energy increases, GEH offers services for adapting plant performance and power output as well as maintenance for extending plant life. GEH also offers services in many other areas, including: Here are

2350-569: 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

2444-564: Is in pre-licensing in Poland. GEH has memoranda of understanding with companies in Canada, Poland, UK, US, and Sweden, among others, and has begun the licensing process in the UK. In 2023, the company signed a contract with Ontario Power Generation (OPG), SNC-Lavalin , and Aecon to deploy a BWRX-300 small modular reactor (SMR) at OPG's Darlington New Nuclear Project site, the first contract for

2538-423: 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

2632-495: Is leading in small modular reactor (SMR) development, offering clean, reliable, and decarbonized power. Nuclear energy, like the BWRX-300, operates at higher capacity factors than renewables, providing consistent, safe, and efficient power, essential for energy security and achieving net-zero emissions. GEH Fuel Manufacturing Sites (Worldwide) GEH’s fuel cycle business supplies fuel products and services to customers around

2726-433: 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

2820-408: 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

2914-407: 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

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3008-453: 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

3102-484: 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

3196-492: 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

3290-419: 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

3384-554: 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

3478-565: 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

3572-420: 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

3666-529: The 1960s, it got involved in constructing and building the Boiling water reactor (BWR). The research into the project continued in the next 50 years resulting in production of 6 different BWR generations. In 1997, the GE-Hitachi U.S. Advanced boiling water reactor (ABWR) design was certified as a final design in final form by the U.S. Nuclear Regulatory Commission . GE and Hitachi officially established

3760-811: The GE Hitachi Nuclear Energy (GEH) global alliance in 2007 by combining parts of their respective power businesses. Based in Wilmington, North Carolina is creating and supplying BWRs and giving assistance with boiling water and pressurized water reactors. In Canada, the organization was known as GE Hitachi Nuclear Energy Canada and its purpose is to provide fuel and service nuclear power plants that operate on heavy water reactors made by Atomic Energy Canada. In 2016, GE and Hitachi sold GE Hitachi Nuclear Energy Canada to BWXT Canada Ltd. and renamed BWXT Nuclear Energy Canada In 2005, GE Hitachi filed design certification by NRC for their Economic Simplified Boiling Water Reactor (ESBWR). The ESBWR received

3854-621: The Gentilly site. It was cancelled by Quebec Premier René Lévesque . A white book study published by the Parti Québécois (PQ) before ascending to power found that Gentilly-3 was not needed for Quebec's future energy needs and that it could be fulfilled with hydroelectricity . After the election of the PQ government, a moratorium on construction of nuclear plants was put into place. The reactor had been scheduled to be completed before 1990, and

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3948-669: The Hitachi-GE ABWR with the UK Office for Nuclear Regulation . The process was completed in December 2017. In January 2020, the company started the regulatory licensing process for the BWRX-300 with the U.S. Nuclear Regulatory Commission . Tennessee Valley Authority (TVA) is undertaking preliminary licensing in collaboration with OPG. SaskPower is considering a deployment, and ORLEN Synthos Green Energy (OSGE) and partners

4042-531: 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,

4136-499: 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

4230-700: The alliance is Hitachi-GE Nuclear Energy . In November 2015, Jay Wileman was appointed CEO. In 1955, the Atomic Power Equipment Department was established by GE. Two years later, in 1957, GE's first privately financed nuclear power reactor provided electricity for commercial use in Vallecitos, California . Additionally, in 1960, GE made and contributed to the Dresden Nuclear Power Station in Chicago. In

4324-465: The area consisted of rural farmlands, but has developed into an industrial community, with the power plant becoming important to the economy in Yokosuka. GE Hitachi Nuclear Energy (GEH) announced plans to expand its Wilmington operations, aiming to create roughly 500 new jobs over the next five years to support the deployment of its modular reactor, the BWRX-300. These reactors are planned to go around

4418-436: 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

4512-512: The challenge of spent nuclear fuel by recycling uranium and transuranics, generating additional electricity while reducing long-term radiotoxicity. This innovative approach could provide a cleaner, more flexible energy solution for the future. BWRX-300 small module reactor GE Hitachi (GEH) has over 60 years of experience in designing, deploying, and servicing nuclear reactors, with 67 reactors licensed in 10 countries. They have produced over 165,000 BWR fuel bundles and hold over 6,660 patents. GEH

4606-419: 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

4700-430: 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

4794-466: 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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4888-536: 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

4982-505: The cost and complexity of the unit, again to make it an attractive export unit. However, the design was not successful, and over 7 years recorded only 180 on-power days. Gentilly-1 is no longer in operation. Gentilly-2 was a standard CANDU 6 reactor, similar to the Point Lepreau Nuclear Generating Station . The plant had a net output of 675MW(e). Unlike the adjacent Gentilly-1 reactor, Gentilly-2 had

5076-432: 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

5170-532: 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

5264-428: 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

5358-418: 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

5452-411: 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

5546-442: 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

5640-414: 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

5734-410: 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

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5828-405: 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

5922-405: 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

6016-400: 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

6110-638: The list of GEH’s Power Plant offerings Boiling Water Reactors GE and Hitachi have developed the world’s safest Boiling Water Reactors (BWRs) over 60 years, with 40 reactors operating in 5 countries. BWRs and Pressurized Water Reactors (PWRs) both use light water as coolant and steam source, but BWRs generate steam directly in the reactor core, while PWRs use a secondary loop to produce steam. Sodium Fast Reactors GEH has over 70 years of experience in developing Sodium Fast Reactors (SFRs), offering greater fuel efficiency (4x) and higher electricity output (100x) than light water reactors (LWRs). SFRs help address

6204-508: 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

6298-571: 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

6392-591: 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)

6486-469: 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

6580-404: 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

6674-530: The power plant instead. The process will take approximately 50 years to complete. In December of that same year, the remaining reactor was shut down and the decommissioning process started. In August 2023, Hydro-Québec reported it was assessing the state of the plant to determine whether or not the Gentilly-2 CANDU reactor could be recommissioned. This came as the province of Quebec looked towards options to increase its production of clean electricity. It

6768-479: 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

6862-510: 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

6956-423: 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

7050-479: 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

7144-706: 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

7238-427: 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

7332-485: 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

7426-440: 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

7520-429: 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

7614-406: 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

7708-461: 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

7802-403: 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

7896-450: 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

7990-400: 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

8084-437: 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,

8178-423: 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

8272-546: 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

8366-749: The world. GE Hitachi Nuclear Energy owns the Morris Operation —the only de facto high-level radioactive waste storage site in the United States. Wilmington ( New Hanover County , North Carolina, United States of America) It was dedicated in 1969, and is known as the home to the global headquarters. Takes up more than 1,500 acres. Produces zirconium-calloy components, uranium dioxide powder and pellets, and fuel assemblies for boiling water reactor market. Kurihama ( Yokosuka City, Kanagawa Prefecture, Japan) Established in 1970, it has helped Japanese consumers provide nuclear energy. Originally

8460-511: Was decided to not proceed with recommissioning Gentilly-2 due to social acceptability issues. Gentilly-1 was a prototype CANDU- BWR reactor, based on the SGHWR design. It was designed for a net output of 250MW(e). The reactor had several features unique amongst CANDU reactors, including vertically oriented pressure tubes (allowing for the use of a single fuelling machine below the core), and light-water coolant. These features were intended to reduce

8554-407: 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

8648-541: Was the last reactor firmly committed to by Hydro-Québec and the Province of Quebec, though Quebec had committed to buy enough heavy water for four Candu style reactors, processed by the La Prade heavy water plant (near Trois-Rivières ), scheduled for 1982 opening. Boiling water reactor A boiling water reactor ( BWR ) is a type of light water nuclear reactor used for the generation of electrical power. It

8742-439: 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

8836-399: 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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