Boiling water reactor

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Schematic diagram of a boiling water reactor (BWR):
Reactor pressure vessel
Nuclear fuel element
Control rods
Recirculation pumps
Control rod drives
High-pressure turbine
Low-pressure turbine
Feedwater pump
Cold-water pump
Concrete enclosure
Connection to electricity grid Boiling water reactor no text.svg
Schematic diagram of a boiling water reactor (BWR):
  1. Reactor pressure vessel
  2. Nuclear fuel element
  3. Control rods
  4. Recirculation pumps
  5. Control rod drives
  6. Steam
  7. Feedwater
  8. High-pressure turbine
  9. Low-pressure turbine
  10. Generator
  11. Exciter
  12. Condenser
  13. Coolant
  14. Pre-heater
  15. Feedwater pump
  16. Cold-water pump
  17. Concrete enclosure
  18. Connection to electricity grid

A boiling water reactor (BWR) is a type of light water nuclear reactor used for 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 a lower pressure water system, which turns to steam and 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.

Nuclear reactor device to initiate and control a sustained nuclear chain reaction

A nuclear reactor, formerly known as an atomic pile, is a device used to initiate and control a self-sustained nuclear chain reaction. Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nuclear fission is passed to a working fluid, which in turn runs through steam turbines. These either drive a ship's propellers or turn electrical generators' shafts. Nuclear generated steam in principle can be used for industrial process heat or for district heating. Some reactors are used to produce isotopes for medical and industrial use, or for production of weapons-grade plutonium. Research reactors are run only for research. As of early 2019, the IAEA reports there are 454 nuclear power reactors and 226 nuclear research reactors in operation around the world.

Pressurized water reactor nuclear power plant with a cooling system that operates under high pressure

Pressurized water reactors (PWRs) constitute the large majority of the world's nuclear power plants and are one of three types of light water reactor (LWR), the other types being boiling water reactors (BWRs) and supercritical water reactors (SCWRs). 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 water then flows to a steam generator where it transfers its thermal energy to a secondary system where steam is generated and flows to turbines which, in turn, spin an electric generator. In contrast to a boiling water reactor, pressure in the primary coolant loop prevents the water from boiling within the reactor. All LWRs use ordinary water as both coolant and neutron moderator.

Nuclear reactor core portion of a nuclear reactor containing the nuclear fuel

A nuclear reactor core is the portion of a nuclear reactor containing the nuclear fuel components where the nuclear reactions take place and the heat is generated. Typically, the fuel will be low-enriched uranium contained in thousands of individual fuel pins. The core also contains structural components, the means to both moderate the neutrons and control the reaction, and the means to transfer the heat from the fuel to where it is required, outside the core.



Animation of a BWR with cooling towers.

A boiling water reactor (BWR) uses demineralized water as a coolant and neutron moderator. Heat 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 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 104 and 107 (i.e., one core damage accident per every 10,000 to 10,000,000 reactor years). [1]

Neutron moderator medium that reduces the speed of fast neutrons, turning them into thermal neutrons that can sustain nuclear chain reactions; e.g. water, graphite, heavy water, beryllium

In nuclear engineering, a neutron moderator is a medium that reduces the speed of fast neutrons, thereby turning them into thermal neutrons capable of sustaining a nuclear chain reaction involving uranium-235 or a similar fissile nuclide.

Turbine rotary mechanical device that extracts energy from a fluid flow

A turbine is a rotary mechanical device that extracts energy from a fluid flow and converts it into useful work. The work produced by a turbine can be used for generating electrical power when combined with a generator. A turbine is a turbomachine with at least one moving part called a rotor assembly, which is a shaft or drum with blades attached. Moving fluid acts on the blades so that they move and impart rotational energy to the rotor. Early turbine examples are windmills and waterwheels.

Condenser (heat transfer) device used to condense a substance from its gaseous to its liquid state

In systems involving heat transfer, a condenser is a device or unit used to condense a substance from its gaseous to its liquid state, by cooling it. In so doing, the latent heat is given up by the substance and transferred to the surrounding environment. Condensers can be made according to numerous designs, and come in many sizes ranging from rather small (hand-held) to very large. For example, a refrigerator uses a condenser to get rid of heat extracted from the interior of the unit to the outside air. Condensers are used in air conditioning, industrial chemical processes such as distillation, steam power plants and other heat-exchange systems. Use of cooling water or surrounding air as the coolant is common in many condensers.


Condensate and feedwater

Steam exiting 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 the top of the nuclear fuel assemblies (these nuclear fuel assemblies constitute the "core") but below the water level.

Feedwater heater

A feedwater heater is a power plant component used to pre-heat water delivered to a steam generating boiler. Preheating the feedwater reduces the irreversibilities involved in steam generation and therefore improves the thermodynamic efficiency of the system. This reduces plant operating costs and also helps to avoid thermal shock to the boiler metal when the feedwater is introduced back into the steam cycle.

Reactor pressure vessel

A reactor pressure vessel (RPV) in a nuclear power plant is the pressure vessel containing the nuclear reactor coolant, core shroud, and the reactor core.

Nuclear fuel material that can be used in nuclear fission or fusion to derive nuclear energy

Nuclear fuel is material used in nuclear power stations to produce heat to power turbines. Heat is created when nuclear fuel undergoes nuclear fission.

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 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 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.

Porosity or void fraction is a measure of the void spaces in a material, and is a fraction of the volume of voids over the total volume, between 0 and 1, or as a percentage between 0% and 100%. Strictly speaking, some tests measure the "accessible void", the total amount of void space accessible from the surface. There are many ways to test porosity in a substance or part, such as industrial CT scanning. The term porosity is used in multiple fields including pharmaceutics, ceramics, metallurgy, materials, manufacturing, hydrology, earth sciences, soil mechanics and engineering.

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 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 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 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 down and directed out into the downcomer or annulus region. The "dry" steam then exits the RPV through four main steam lines and goes to the turbine.

Control systems

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.

Control rod

Control rods are used in nuclear reactors to control the fission rate of uranium and plutonium. They are composed of chemical elements such as boron, silver, indium and cadmium that are capable of absorbing many neutrons without themselves fissioning. Because these elements have different capture cross sections for neutrons of varying energies, the composition of the control rods must be designed for the reactor's neutron spectrum. Boiling water reactors (BWR), pressurized water reactors (PWR) and heavy water reactors (HWR) operate with thermal neutrons, while breeder reactors operate with fast neutrons.

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 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 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 compensate the consumption of the fuel, while the power is controlled through the water flow (see below). [2] 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.

Boron carbide chemical compound

Boron carbide (chemical formula approximately B4C) is an extremely hard boron–carbon ceramic, and covalent material used in tank armor, bulletproof vests, engine sabotage powders, as well as numerous industrial applications. With a Vickers Hardness of >30 GPa, it is one of the hardest known materials, behind cubic boron nitride and diamond.

Changing (increasing or decreasing) 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 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 down 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 down to be absorbed by the fuel, and reactor power decreases. [3]

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 will automatically follow reactor power changes. When 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 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, 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 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 turbines

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, 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 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 for 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 reactor'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.

Reactor core

A modern BWR fuel assembly comprises 74 to 100 fuel rods, and there are up to approximately 800 assemblies in 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.

Safety systems

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 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.

Refueling systems

The reactor fuel rods are occasionally replaced by removing them from the top of the containment vessel. 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 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 reactor incident this became problematic because water was lost from one or more spent fuel pools and the earthquake could have altered the geometry. The fact that the fuel rods' cladding is a zirconium alloy was also problematic since this element can react with steam at extreme temperatures 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 the rod.


Early concepts

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 3 military services. The Navy, seeing the possibility of turning submarines into full-time underwater vehicles, and ships that could steam around 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 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 in such, in any event.

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, 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 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. [4]

Following this series of tests, GE got involved and collaborated with ANL [5] 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 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.

First series of production

Cross-section sketch of a typical BWR Mark I containment
Browns Ferry Unit 1 under construction.jpg
Browns Ferry Unit 1 drywell and wetwell under construction, a BWR/4 using the Mark I containment

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 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 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. [6]

Apart from the GE designs there were others by ABB, MITSU, Toshiba and KWU. See List of boiling water reactors.

Advanced boiling water reactor

Cross section of UK ABWR design Reinforced Concrete Containment Vessel UK ABWR cross section.png
Cross section of UK ABWR design Reinforced Concrete Containment Vessel

A newer design of BWR 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 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.[ citation needed ]

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 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.

Simplified boiling water reactor

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 was notable for its incorporation—for the first time ever in a light water reactor[ citation needed ]—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 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 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 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 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 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 was withdrawn prior to approval; still, the concept remained intriguing to General Electric's designers, and served as the basis of future developments.

Economic simplified boiling water reactor

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 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. [7]

Reportedly, this design has been advertised as having a core damage probability of only 3×10−8 core damage events per reactor-year.[ citation needed ] That is, there would need to be 3 million ESBWRs operating before one would expect 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 1×10−5 core-damage events per reactor-year. [8] This extraordinarily low CDP for the ESBWR far exceeds the other large LWRs on the market.



Technical and background information

Start-up ("going critical")

Reactor start up (criticality) is achieved by withdrawing control rods from the core to raise core reactivity to a level where it 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 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 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 a single control rod can cause a large uneven reactivity change which can potentially challenge the fuel's thermal design margins. As a result, GE developed a set of rules in 1977 called BPWS (Banked Position Withdrawal Sequence) which help minimize the worth of any single control rod 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 "checkboard" pattern. Next the opposing group (B or A) is pulled in a defined sequence to positions 02, then 04, 08, 16, and finally full out (48), until the reactor enters the power operation range where thermal limits are no longer bounding. 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. [10]

Thermal margins

Several calculated/measured quantities are tracked while operating 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).

Maximum fraction limiting critical power ratio (MFLCPR)

Specifically, MFLCPR represents how close 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 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).

MFLCPR 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 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. Nuclear fuel could be damaged by film boiling; this would cause the fuel cladding to overheat and fail. 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 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. [11] 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.

Fraction limiting linear heat generation rate (FLLHGR)

FLLHGR (FDLRX, MFLPD) 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 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 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 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.

Average planar linear heat generation rate (APLHGR)

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 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 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 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 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 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.

Pre-Conditioning Interim Operating Management Recommendation (PCIOMR)

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 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 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 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.

List of BWRs

For a list of operational and decommissioned BWRs, see List of BWRs.

Experimental and other types

Experimental and other non-commercial BWRs include:

Next-generation designs

See also

References and notes

  1. Susan Dingman; Jeff LaChance; Allen Canip; Mary Drouin. "Core damage frequency perspectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results". Retrieved 2013-08-02.
  2. 1 2 3 Bonin, Bernhard; Klein, Etienne (2012). Le nucléaire expliqué par des physiciens.
  3. James W. Morgan, Exelon Nuclear (15 November 2007). "Upgrade your BWR recirc pumps with adjustable-speed drives". Power: Business and Technology for the Global Generation Industry. Retrieved 20 March 2011.
  4. Boiling Water Reactor Simulator with Passive Safety Systems - IAEA (PDF (11 MB)), IAEA, October 2009, p. 14, retrieved 8 June 2012
  6. Sandia National Laboratories (July 2006), Containment Integrity Research at Sandia National Laboratories – An Overview (PDF), U.S. Nuclear Regulatory Commission, NUREG/CR-6906, SAND2006-2274P, retrieved 13 March 2011
  8. Hinds, David; Maslak, Chris (January 2006). "Next-generation nuclear energy: The ESBWR" (PDF). Nuclear News. La Grange Park, Illinois, United States of America: American Nuclear Society. 49 (1): 35–40. ISSN   0029-5574 . Retrieved 2009-04-04.
  9. KEIJI TAKEUCHI COMMENTARY: Crucial vents were not installed until 1990s
  10. NEDO-21231, "Banked Position Withdrawal Sequence," January 1977. General Electric Corporation
  11. NUREG-0800, (67:234) Chpt 4, Section 4.4, Rev. 1, Thermal and Hydraulic Design, of the Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants. LWR Edition. (10 page(s), 7/31/1981)

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