Learn how and when to remove this template message 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. 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.
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Learn how and when to remove this template message 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.
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 degree turn and moves up through the lower core plate into the nuclear core, where the fuel elements heat the water.
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[ edit ] 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 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.
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. 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.
Reactor water level is controlled by the main feedwater system. From about 0. 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.
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[ edit ] 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, 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 degrees F for the low-pressure turbines to use.
The exhaust of the low-pressure turbines is sent to the main condenser. While the reheaters take steam away from the turbine, the net result is that the reheaters improve the thermodynamic efficiency of the plant.
Main articles: Nuclear reactor core and Fuel rod A modern BWR fuel assembly comprises 74 to fuel rods , and there are up to approximately assemblies in a reactor core , holding up to approximately 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.
Main article: Boiling water reactor 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[ edit ] 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.
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. 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.
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. The literature does not indicate why this was the case, but it was eliminated on production models of the BWR.
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. See List of boiling water reactors. The ABWR was developed in the late s and early s, and has been further improved to the present day. Most significantly, the ABWR was a completely standardized design, that could be made for series production.
This smaller 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.
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.
Pressure vessel is subject to significantly less irradiation compared to a PWR, and so does not become as brittle with age. Operates at a lower nuclear fuel temperature, largely due to heat transfer by the latent heat of vaporization, as opposed to sensible heat in PWRs. Fewer components due to a lack of steam generators and a pressurizer vessel, as well as the associated primary circuit pumps.
This also makes BWRs simpler to operate. Lower risk probability of a rupture causing loss of coolant compared to a PWR, and lower risk of core damage should such a rupture occur. This is due to fewer pipes, fewer large diameter pipes, fewer welds and no steam generator tubes.
Measuring the water level in the pressure vessel is the same for both normal and emergency operations, which results in easy and intuitive assessment of emergency conditions. Can operate at lower core power density levels using natural circulation without forced flow. A BWR may be designed to operate using only natural circulation so that recirculation pumps are eliminated entirely. BWRs do not use boric acid to control fission burn-up to avoid the production of tritium contamination of the turbines ,  leading to less possibility of corrosion within the reactor vessel and piping.
Corrosion from boric acid must be carefully monitored in PWRs; it has been demonstrated that reactor vessel head corrosion can occur if the reactor vessel head is not properly maintained.
See Davis-Besse. Since BWRs do not utilize boric acid, these contingencies are eliminated. The power control by reduction of the moderator density vapour bubbles in the water instead of by addition of neutron absorbers boric acid in PWR leads to breeding of U by fast neutrons, producing fissile Pu BWRs generally have N-2 redundancy on their major safety-related systems, which normally consist of four "trains" of components.
This generally means that up to two of the four components of a safety system can fail and the system will still perform if called upon. Still, some countries could reach a high level of standardisation with PWRs, like France. Additional families of PWRs are being introduced. BWRs are overrepresented in imports, when the importing nation does not have a nuclear navy PWRs are favored by nuclear naval states due to their compact, high-power design used on nuclear-powered vessels; since naval reactors are generally not exported, they cause national skill to be developed in PWR design, construction, and operation.
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It also uses natural circulation to drive coolant flow within the reactor pressure vessel RPV ; this results in fewer systems to maintain, and precludes significant BWR casualties such as recirculation line breaks. There are no circulation pumps or associated piping, power supplies, heat exchangers, instrumentation, or controls needed for these systems. These systems utilize natural circulation based on simple laws of physics to transfer the decay heat outside containment while maintaining water levels inside the reactor, keeping the nuclear fuel submerged in water and adequately cooled. In events where the reactor coolant pressure boundary remains intact, the Isolation Condenser System ICS is used to remove decay heat from the reactor and transfer it outside containment. The ICS system is a closed loop system that connects the reactor pressure vessel to a heat exchanger located in the upper elevation of the reactor building. Steam leaves the reactor through the ICS piping and travels to the ICS heat exchangers which are submerged in a large pool. The steam is condensed in the heat exchangers and the denser condensate then flows back down to the reactor to complete the cooling loop.
Boiling water reactor