Editing Boiling water reactor (section)

== Components ==
=== Condensate and feedwater ===
{{Unreferenced section|date=July 2011}}
[[File:Boiling water reactor no text.svg|thumb|upright=1.60|Schematic diagram of a ''boiling water reactor'' (BWR):
{{columns-list|colwidth=30em|{{ordered list
 | Reactor pressure vessel
 | Nuclear fuel element
 | Control rods
 | Recirculation pumps
 | Control rod drives
 | Steam
 | Feedwater
 | High-pressure turbine
 | Low-pressure turbine
 | Generator
 | Exciter
 | Condenser
 | Coolant
 | Pre-heater
 | Feedwater pump
 | Cold-water pump
 | Concrete enclosure
 | Connection to electricity grid
}}}}
]]
Steam exiting the [[turbine]] flows into [[condenser (heat transfer)|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 heater]]s 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 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&nbsp;kg/h (100,000,000&nbsp;lb/h) with 6,500,000&nbsp;kg/h (14,500,000&nbsp;lb/h) steam flow. However, core-average [[porosity|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.

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 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 rod]]s (control blades) and by changing the water flow through the [[Nuclear reactor core|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 to compensate for the consumption of the fuel, while the power is controlled through the water flow (see below).<ref name="bonin">{{cite book |last1=Bonin |first1=Bernhard |last2=Klein |first2=Etienne |date=2012 |title=Le nucléaire expliqué par des physiciens}}</ref> 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) 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 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 the fuel, and reactor power decreases.<ref>{{cite web |url=http://www.powermag.com/nuclear/Upgrade-your-BWR-recirc-pumps-with-adjustable-speed-drives_369.html |title=Upgrade your BWR recirc pumps with adjustable-speed drives |author=James W. Morgan, Exelon Nuclear |date=15 November 2007 |publisher=Power: Business and Technology for the Global Generation Industry |access-date=20 March 2011 |archive-date=2 October 2011 |archive-url=https://web.archive.org/web/20111002150952/http://www.powermag.com/nuclear/Upgrade-your-BWR-recirc-pumps-with-adjustable-speed-drives_369.html |url-status=dead }}</ref> 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 and bypass valves will modulate to maintain reactor pressure at a setpoint. Under this control mode, the turbine output 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 [[radionuclide]]s due to neutron capture from 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 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. Some of the High-pressure turbine exhaust is sent to the Moisture Separator Reheaters (MSR) and some is sent to the feedwater heaters. This moisture separator reheater employs a group of chevrons to remove entrained moisture from the steam. The steam is then superheated to over 400 degrees F (204.4 degrees Celsius) for the low-pressure turbines to use. The exhaust of the low-pressure turbines is sent to the main condenser. The moisture separator reheaters take some of the high pressure turbine's steam, and 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 and reactor, the net result is that the reheaters improve the thermodynamic efficiency of the plant.

=== Reactor core ===
{{main|Nuclear reactor core|Fuel rod}}
A modern BWR fuel assembly comprises 74 to 100 [[fuel rod]]s, and there are up to approximately 800 assemblies in a [[Nuclear reactor core|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 ===
{{main|Boiling water reactor safety systems}}
In order to be considered safe, nuclear reactors must at all time comply with the fundamental nuclear safety functions:

# Control of reactivity
# Removal of heat from the fuel
# Confinement of radioactive materials, shielding against radiation, and control of planned radioactive releases, as well as limitation of accidental releases

During normal operation of the reactor, good design, construction and operation practices ensure the compliance with the safety functions. In addition, dedicated [[Boiling water reactor safety systems|safety systems]] must be installed to ensure that compliance with the safety functions is maintained in case of deviations from normal conditions. The safety systems of a modern reactor are designed to according to a [[Defence in depth (non-military)|defence in depth]] philosophy, which is a design philosophy that is integrated throughout construction and [[Building commissioning|commissioning]] of the plant. In order to minimize the risks of inability to fulfil a safety function, a BWR uses redundant safety systems, relying on different action mechanisms and spatially separated.

Following emergency shutdown of the fission reaction, heat continues to be produced inside nuclear reactors due to the [[radioactive decay]] of fission products and materials that have been activated by [[neutron absorption]]. The decay heat of a reactor core is of the order of a few tens of megawatt (depending on the core's operating power and operation duration) which is sufficient to lead to significant damage and even melting. To prevent such cases, safety systems grouped under the umbrellas of [[Emergency Core Cooling System|emergency core cooling systems]] (ECCS) are provided in order to remove heat through the injection of cooling water.

=== Refueling systems ===
The reactor fuel rods are occasionally replaced by moving them from the reactor pressure vessel to the spent fuel pool. A typical fuel cycle lasts 12–24 months, with about one third to one fifth 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 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 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 {{convert|1500|K|C}} to produce hydrogen,<ref>{{Cite book
| last = Kuan
| first = P.
|author2=Hanson, D. J. |author3=Odar, F.
| title = Managing water addition to a degraded core
| year = 1991
| osti = 5642843
}}</ref><ref name="HaskinStages">{{cite book
|title=Perspectives on Reactor Safety (NUREG/CR-6042) (Reactor Safety Course R-800), 1st Edition
|year=1994
|publisher=U.S. Nuclear Regulatory Commission
|url=https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6042/
|author1=Haskin, F.E. |author2=Camp, A.L. |access-date=23 November 2010
|location=Beltsville, MD
|page=3.1–5
}}</ref>
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.