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The boiling water reactor (BWR) is a type of 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). The BWR was developed by the Idaho National Laboratory and General Electric in the mid-1950s. In the present, General Electric specializes in the design and construction of this type of reactor.

Contents

Overview

BoilingWaterReactor.gif

The BWR uses demineralized water (light 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 PWR because of the high pressure maintained in its primary loop—approximately 158 atm (16 MPa, 2300 psi).

Description of major components and systems

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Feedwater

Steam exiting from 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 region and combines with water exiting the water separators. The feedwater subcools the saturated water from the steam separators. This water now flows down the downcomer 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 about 12 to 15% saturated steam (by mass), 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.

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. 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 water 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 region. In the downcomer 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 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 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. 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. Fine reactivity adjustment would be accomplished by modulating the recirculation flow of the reactor vessel.

Changing (increasing or decreasing) the flow of water through the core is the normal and convenient method for controlling 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. 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.

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.

Size

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 tons of 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

Like the pressurized water reactor, the BWR reactor core continues to produce heat from radioactive decay after the fission reactions have stopped, making a core damage incident possible in the event that all safety systems have failed and the core does not receive coolant. Also like the pressurized water reactor, a boiling water reactor has a negative void coefficient, that is, the neutron (and the thermal) output of the reactor decreases as the proportion of steam to liquid water increases inside the reactor.

However, unlike a pressurized water reactor which contains no steam in the reactor core, a sudden increase in BWR steam pressure (caused, for example, by the actuation of the main steam isolation valve (MSIV) from the reactor) will result in a sudden decrease in the proportion of steam to liquid water inside the reactor. The increased ratio of water to steam will lead to increased neutron moderation, which in turn will cause an increase in the power output of the reactor. This type of event is referred to as a "pressure transient".

The BWR is specifically designed to respond to pressure transients, having a "pressure suppression" type of design which vents overpressure using safety relief valves to below the surface of a pool of liquid water within the containment, known as the "wetwell" or "torus". There are 11 safety overpressure relief valves on BWR/1-BWR/6 models (7 of which are part of the ADS)[1] and 18 safety overpressure relief valves on ABWR models[2], only a few of which have to function to stop the pressure rise of a transient. In addition, the reactor will have already have rapidly shut down before the transient affects the RPV (as described in the Reactor Protection System section below).

Because of this effect in BWRs, operating components and safety systems are designed to ensure that no credible scenario can cause a pressure and power increase that exceeds the systems' capability to quickly shutdown the reactor before damage to the fuel or to components containing the reactor coolant can occur. In the limiting case of an ATWS derangement, high neutron power levels (~ 200%) can occur for less than a second, after which actuation of SRVs will cause the pressure to rapidly drop off. Neutronic power will fall to far below nominal power (the range of 30% with the cessation of circulation, and thus, void clearance) even before ARI or SLCS actuation occurs. Thermal power will be barely affected.

In the event of an contingency that disables all of the safety systems, each reactor is surrounded by a containment building consisting of 1.2–2.4 m (4–8 ft) of steel-reinforced, pre-stressed concrete designed to seal off the reactor from the environment.

Reactor Protection System (RPS)

The Reactor Protection System (RPS) is a system, computerized in later BWR models, that is designed to automatically, rapidly, and completely shut down and make safe the Nuclear Steam Supply System (NSSS - the reactor pressure vessel, pumps, and water/steam piping within the containment) if some event occurs that could result in the reactor entering an unsafe operating condition. In addition, the RPS can automatically spin up the Emergency Core Cooling System (ECCS) upon detection of several signals. It does not require human intervention to operate. However, the reactor operators can override parts of the RPS if necessary. If an operator recognizes a deteriorating condition, and knows an automatic safety system will activate, they are trained to pre-emptively activate the safety system.

If the reactor is at power or ascending to power (i.e. if the reactor is critical; the control rods are withdrawn to the point where the reactor generates more neutrons than it absorbs) there are safety-related contingencies that may arise that necessitate a rapid shutdown of the reactor, or, in Western nuclear parlance, a "SCRAM". The SCRAM is a manually-triggered or automatically-triggered rapid insertion of all control rods into the reactor, which will take the reactor to decay heat power levels within tens of seconds. Since ~ 0.6% of neutrons are emitted from fission products ("delayed" neutrons), which are born seconds/minutes after fission, all fission can not be terminated instantaneously, but the fuel soon returns to decay heat power levels. Manual SCRAMs may be initiated by the reactor operators; while automatic SCRAMs are initiated upon:

  1. Turbine stop valve or turbine control valve closure.
    1. If turbine protection systems detect a significant anomaly, admission of steam is halted. Reactor rapid shutdown is in anticipation of a pressure transient that could increase reactivity.
    2. Generator load rejection will also cause closure of turbine valves and trip RPS.
  2. Loss of Offsite Power (LOOP)
    1. During normal operation, the reactor protection system (RPS) is powered by offsite power
      1. Loss of offsite power would open all relays in the RPS would open causing all rapid shutdown signals to come in redundantly.
      2. would also cause MSIV to close since RPS is fail safe; plant assumes a main steam break is coincident with loss of offsite power.
  3. Neutron Monitor Trips - the purpose of these trips are to ensure an even increase in neutron and thermal power during startup.
    1. Source Range Monitor (SRM) / Intermediate Range Monitor (IRM) Upscale:
      1. The SRM, used during instrument calibration, pre-critical, and early non-thermal criticality, and the IRM, used during ascension to power, middle/late non-thermal, and early/middle thermal stages, both have trips built in that prevent rapid decreases in reactor period when reactor is intensely reactive (e.g. when no voids exist, water is cold, and water is dense) without positive operator confirmation that such decreases in period are their intention. Prior to trips occurring, rod movement blocks will be activated to ensure operator vigilance if preset levels are marginally exceeded.
    2. Average Power Range Monitor (APRM) Upscale:
      1. Prevents reactor from exceeding preset neutron power level maxima during operation or relative maxima prior to positive operator confirmation of end of startup by transition of reactor state into "Run".
    3. Average Power Range Monitor/Coolant Flow Thermal Trip:
      1. Prevents reactor from exceeding variable power levels without sufficient coolant flow for that level being present.
  4. Low reactor water level indicative of:
    1. Loss of coolant contingency (LOCA)
    2. Loss of proper feedwater (LOFW)
    3. etc.
  5. High drywell (primary containment) pressure
    1. Indicative of potential loss of coolant contingency
  6. Main Steam Isolation Valve Closure (MSIV)
    1. Redundant backup for turbine trip
    2. Indicative of potential main steam line break
  7. High RPV pressure:
    1. Indicative of MSIV closure.
    2. Decreases reactivity to compensate for boiling void collapse due to high pressure.
    3. Prevents pressure relief valves from opening.
    4. Serves as a backup for several other trips, like Turbine Trip.

Emergency Core Cooling System (ECCS)

Diagram of a generic BWR Reactor Pressure Vessel

While the RPS is designed to prevent contingencies from happening, the ECCS is designed to respond to contingencies if they do happen. The ECCS is a set of interrelated safety systems that are designed to protect the fuel within the reactor pressure vessel, which is referred to as the "reactor core", from overheating. These systems accomplish this by maintaining reactor pressure vessel (RPV) cooling water level, or if that is impossible, by directly flooding the core with coolant.

These systems are of 3 major types:

  1. High pressure systems: These are designed to protect the core by injecting large quantities of water into it to prevent the fuel from being uncovered by a decreasing water level. Generally used in cases with stuck-open safety valves, small breaks of auxiliary pipes, and particularly violent transients caused by turbine trip and MSIV closure. If the water level cannot be maintained with high pressure systems alone (the water level still is falling below a preset point with the high-pressure systems working full-bore), the next set of systems responds.
  2. Depressurization systems: These systems are designed to reduce and maintain the level of pressure at a lower level. As pressure is reduced, steam condenses to liquid water, and this increases the core water level. If water level is still falling with partial depressurization and full high-pressure system functioning, full depressurization will occur to a much lower pressure level, where the next set of systems respond.
  3. Low pressure systems: These systems are designed to function after the depressurization systems function. They have extremely large capacities compared to the high pressure systems and are supplied by multiple, redundant power sources. They will maintain any maintainable water level, and, in the event of a large pipe break of the worst type below the core that leads to temporary fuel rod "uncovery", to rapidly mitigate that state prior to the fuel heating to the point where core damage could occur.

High Pressure Coolant Injection System (HPCI)

The High Pressure Coolant Injection System is the first line of defense in the Emergency Core Cooling System. HPCI is designed to inject substantial quantities of water into the reactor while it is at high pressure so as to prevent the activation of the ADS, CS, and LPCI systems. HPCI is powered by steam from the reactor, and takes approximately 10 seconds to spin up from an initiating signal, and can deliver approximately 19,000 L/min (5,000 US gal/min) to the core at any core pressure above 6.8 atm (690 kPa, 100 psi). This is usually enough to keep water levels sufficient to avoid automatic depressurization except in a major contingency, such as a large break in the makeup water line.

Versioning note: The BWR/6 replaces HPCI with High Pressure Core Spray (HPCS); ABWRs and (E)SBWRs replace HPCI with High Pressure Core Flooder (HPCF), a mode of the RCIC system, as described below.

Reactor Core Isolation Cooling System (RCIC)

The Reactor Core Isolation Cooling System is not a safety-related system proper, but is included because it can help cool the reactor in the event of a contingency, and it has additional functionality in advanced versions of the BWR.

RCIC is designed to remove the residual heat of the fuel from the reactor once it has been shut down. It injects approximately 2,000 L/min (600 gpm) into the reactor core for this purpose, at high pressure. It also takes less time to start than the HPCI system, approximately 5 seconds from an initiating signal.

Versioning note: RCIC and HPCF are integrated in ABWRs and (E)SBWRs, with HPCF representing the high-capacity mode of RCIC. In the (E)SBWR series of reactors, there is an additional contingency residual heat removal capability for RCIC, the Isolation Condenser System (IC); in the (E)SBWR, there are several separate trains of heat exchangers located above the RPV in deep pools of water within the reactor building but outside and above the primary containment. In the event of a contingency, the decay heat of the reactor will boil water to steam within the RPV. The RPS will activate several valves connecting the RPV to the IC system; the steam from the RPV decay heat will flow into the heat exchangers (called Isolation Condensers) and be condensed and cooled back to liquid. The water will then return to the RPV through the force of gravity.

Automatic Depressurization System (ADS)

The Automatic Depressurization System is not a part of the cooling system proper, but is an essential adjunct to the ECCS. It is designed to activate in the event that the RPV is retaining pressure, but RPV water level cannot be maintained using high pressure cooling alone, and low pressure cooling must be initiated. When ADS fires, it rapidly releases pressure from the RPV in the form of steam through pipes that are piped to below the water level in the suppression pool (the torus/wetwell), which is designed to condense the steam released by ADS or other safety valve activation into water), bringing the reactor vessel below 32 atm (3200 kPa, 465 psi), allowing the low pressure cooling systems (LPCS/LPCI/LPCF/GDCS), with extremely large and robust comparative coolant injection capacities to be brought to bear on the reactor core.

Low Pressure Core Spray System (LPCS)

The Low Pressure Core Spray System is designed to suppress steam generated by a major contingency. As such, it prevents reactor vessel pressure from going above the point where LPCI and LPCS would be ineffective, which is above 32 atm (3200 kPa, 465 psi). It activates below that level, and delivers approximately 48,000 L/min (12,500 US gal/min) of water in a deluge from the top of the core.

Versioning note: In ABWRs and (E)SBWRs, there are additional water spray systems to cool the drywell and the suppression pool.

Low Pressure Coolant Injection System (LPCI)

The Low Pressure Coolant Injection System is the "heavy artillery" in the ECCS. Consisting of 4 pumps driven by diesel engines, it is capable of injecting a mammoth 150,000 L/min (40,000 US gal/min) of water into the core. It is capable of being brought to bear at reactor vessel pressures below 465 psi. Combined with the CS to keep steam pressure low, the LPCI can suppress all contingencies by rapidly and completely flooding the core with coolant.

Versioning note: ABWRs replace LPCI with Low Pressure Core Flooder (LPCF), which operates using similar principles. (E)SBWRs replace LPCI with the DPVS/PCCS/GDCS, as described below.

Depressurization Valve System (DPVS)/Passive Containment Cooling System (PCCS)/Gravity Driven Cooling System (GDCS)

The (E)SBWR has an additional ECCS capacity that is completely passive, quite unique, and significantly improves defense in depth. This system is activated when the water level within the RPV reaches Level 1. At this point, a countdown timer is started.

There are several large depressurization valves are located near the top of the reactor pressure vessel. These constitute the DPVS. This is a capability supplemental to the ADS, which is also included on the (E)SBWR. The DPVS consists of eight of these valves, four on main steamlines that vent to the drywell when actuated and four venting directly into the drywell.

If Level 1 is not resubmerged within 50 seconds of the timer starting, DPVS will fire and will rapidly vent any pressure contained within the reactor pressure vessel into the drywell. This will cause the water within the RPV to gain in volume (due to the drop in pressure) which will increase the water available to cool the core. In addition, the depressurization will cause a lower boiling point, and thus more steam bubbles will form, decreasing moderation; this, in turn, decreases decay heat production, while still maintaining adequate cooling. (In fact, both the ESBWR and the ABWR are designed so that even in the maximum feasible contingency, the core never loses its layer of water coolant.)

If Level 1 is not again not resubmerged within 100 seconds of DPVS actuation, then the GDCS valves fire. The GDCS is a series of very large water tanks located above and to the side of the Reactor Pressure Vessel within the drywell. When these valves fire, the GDCS is directly connected to the RPV. After ~50 more seconds of depressurization, the pressure within the GDCS will equalize with that of the RPV and drywell, and the water of the GDCS will begin flowing into the RPV.

The water within the RPV will boil into steam from the decay heat, and natural convection will cause it to travel upwards into the drywell, into piping assemblies in the ceiling that will take the steam to four large heat exchangers - the Passive Containment Cooling System (PCCS) - located above the drywell - in deep pools of water. The steam will be cooled, and will condense back into liquid water. The liquid water will drain from the heat exchanger back into the GDCS pool, where it can flow back into the RPV to make up for additional water boiled by decay heat. In addition, if the GDCS lines break, the shape of the RPV and the drywell will ensure that a "lake" of liquid water forms that submerges the bottom of the RPV (and the core within).

There is sufficient water to cool the heat exchangers of the PCCS for 72 hours. At this point, all that needs to happen is for the pools that cool the PCCS heat exchangers to be refilled, which is a comparatively trivial operation, doable with a portable fire pump and hoses.

GE has a computerized animation of how the ESBWR functions during a pipe break incident on their website.

Standby Liquid Control System (SLCS)

The Standby Liquid Control System is used in the event of major contingencies as a last measure to prevent core damage. It is not intended ever to be used, as the RPS and ECCS are designed to respond to all contingencies, even if a quite a few of their components fail, but if a complete ECCS failure occurs, during a limiting fault, it could be the only thing capable of preventing core damage. The SLCS consists of a tank containing a large quantity of soluble neutron absorbers (typically borated fluids, such as borax) protected by explosively-opened valves and redundant battery-operated pumps, allowing the injection of the borated fluids into the reactor against any pressure within; these borated fluids can and will shut down a reactor gone out of control. The SLCS also provides an additional layer of defense in depth against a ATWS derangement, but this is an extreme measure that can be avoided by numerous other channels (ARI and use of redundant hydraulics).

Versioning note: The SLCS is a system that is never meant to be activated unless all other measures have failed. In the BWR/1 - BWR/6, its activation could cause sufficient damage to the plant that it could make the older BWRs inoperable without a complete overhaul. With the arrival of the ABWR and (E)SBWR, operators do not have to be as reticent about activating the SLCS, as these reactors have a Reactor Water Cleanup System (RWCS) - once the reactor has stabilized, the borated water within the RPV can be filtered through this system to promptly remove the soluble neutron absorbers that it contains and thus avoid damage to the internals of the plant.

Containment system

The ultimate safety system inside and outside of every BWR are the numerous levels of physical shielding that both protect the reactor from the outside world and protect the outside world from the reactor.

There are five levels of shielding:

  1. The fuel rods inside the reactor pressure vessel are coated in thick Zircalloy shielding;
  2. The reactor pressure vessel itself is manufactured out of 6 inch thick steel, with extremely temperature, vibration, and corrosion resistant surgical stainless steel grade grade 316L plate on both the inside and outside;
  3. The primary containment structure is made of steel 1 inch thick;
  4. The secondary containment structure is made of steel-reinforced, pre-stressed concrete 1.2–2.4 meters (4–8 ft) thick.
  5. The reactor building (the shield wall/missile shield) is also made of steel-reinforced, pre-stressed concrete 0.3 m to 1 m (1–3 feet) thick.

If every possible measure standing between safe operation and core damage fails, the containment can be sealed indefinitely, and it will prevent any substantial release of radiation to the environment from occurring in nearly any circumstance.

Varieties of BWR containments

As illustrated by the descriptions of the systems above, BWRs are quite divergent in design from PWRs. Unlike the PWR, which has generally followed a very predictable external containment design (the stereotypical dome atop a cylinder), BWR containments are varied in external form but their internal distinctiveness is extremely striking in comparison to the PWR. There are five major varieties of BWR containments:

  • The "premodern" containment (Generation I); spherical in shape, and featuring a steam drum separator, or an out-of-RPV steam separator, and a heat exchanger for low pressure steam, this containment is now obsolete, and is not used by any operative reactor (??except perhaps the early BWR of Tarapur, India that is still in operation??).
  • the Mark I containment, consisting of a rectangular steel-reinforced concrete building, along with an additional layer of steel-reinforced concrete surrounding the steel-lined cylindrical drywell and the steel-lined pressure suppression torus below. The Mark I was the earliest type of containment in wide use, and many reactors with Mark Is are still in service today. There have been numerous safety upgrades made over the years to this type of containment, especially to provide for orderly reduction of containment load caused by pressure in a compounded limiting fault. The reactor building of the Mark I generally is in the form of a large rectangular cube of reinforced concrete.
  • the Mark II containment, similar to the Mark I, but omitting a distinct pressure suppression torus in favor of a cylindrical wetwell below the non-reactor cavity section of the drywell. Both the wetwell and the drywell have a primary containment structure of steel as in the Mark I, as well as the Mark I's layers of steel-reinforced concrete composing the secondary containment between the outer primary containment structure and the outer wall of the reactor building proper. The reactor building of the Mark II generally is in the form of a flat-topped cylinder.
  • the Mark III containment, generally similar in external shape to the stereotypical PWR, and with some similarities on the inside, at least on a superficial level. For example, rather than having a slab of concrete that staff could walk upon while the reactor was not being refueled covering the top of the primary containment and the RPV directly underneath, the Mark III takes the BWR in a more PWRish direction by placing a water pool over this slab. Additional changes include abstracting the wetwell into a pressure-suppression pool with a weir wall separating it from the drywell.
  • Advanced containments; the present models of BWR containments for the ABWR and the ESBWR are harkbacks to the classical Mark I/II style of being quite distinct from the PWR on the outside as well as the inside, though both reactors incorporate the Mark III-ish style of having non-safety-related buildings surrounding or attached to the reactor building, rather than being overtly distinct from it. These containments are also designed to take far more than previous containments were, providing advanced safety. In particular, GE regards these containments as being able to withstand a direct hit by a tornado of Old Fujitsa Scale 6 with winds of 330+ miles per hour. Such a tornado has never occurred. They are also designed to withstand seismic accelerations of .2 G, or nearly 2 meters per second in any direction.

The safety systems in action: the Design Basis Accident

The Design Basis Accident (DBA) for a nuclear power plant is the most severe possible single accident that the designers of the plant and the regulatory authorities could imagine. It is, also, by definition, the accident the safety systems of the reactor are designed to respond to successfully, even if it occurs when the reactor is in its most vulnerable state. The DBA for the BWR consists of the total rupture of a large coolant pipe in the location that is considered to place the reactor in the most danger of harm—specifically, for older BWRs (BWR/1-BWR/6), the DBA consists of a "guillotine break" in the coolant loop of one of the recirculation jet pumps, which is substantially below the core waterline (LBLOCA, large break loss of coolant accident) combined with loss of feedwater to make up for the water boiled in the reactor (LOFW, loss of proper feedwater), combined with a simultaneous collapse of the regional power grid, resulting in a loss of power to certain reactor emergency systems (LOOP, loss of offsite power). The BWR is designed to shrug this accident off without core damage.

The description of this accident is applicable for the BWR/4, which is the oldest model of BWR in common service.

The immediate result of such a break (call it time T+0) would be a pressurized stream of water well above the boiling point shooting out of the broken pipe into the drywell, which is at atmospheric pressure. As this water stream flashes into steam, due to the decrease in pressure and that it is above the water boiling point at normal atmospheric pressure, the pressure sensors within the drywell will report a pressure increase anomaly within it to the Reactor Protection System at latest T+0.3. The RPS will interpret this pressure increase signal, correctly, as the sign of a break in a pipe within the drywell. As a result, the RPS immediately initiates a full SCRAM, closes the Main Steam Isolation Valve (isolating the containment building), trips the turbines, attempts to begin the spinup of RCIC and HPCI, using residual steam, and starts the diesel pumps for LPCI and CS.

Now let us assume that the power outage hits at T+0.5. The RPS is on a float uninterruptable power supply, so it continues to function; its sensors, however, are not, and thus the RPS assumes that they are all detecting emergency conditions. Within less than a second from power outage, auxiliary batteries and compressed air supplies are starting the Emergency Diesel Generators. Power will be restored by T+25 seconds.

Let us return to the reactor core. Due to the closure of the MSIV (complete by T+2), a wave of backpressure will hit the rapidly depressurizing RPV but this is immaterial, as the depressurization due to the recirculation line break is so rapid and complete that no steam voids will probably flash to water. HPCI and RCIC will fail due to loss of steam pressure in the general depressurization, but this is again immaterial, as the 2,000 L/min (600 US gal/min) flow rate of RCIC available after T+5 is insufficient to maintain the water level; nor would the 19,000 L/min (5,000 US gal/min) flow of HPCI, available at T+10, be enough to maintain the water level, if it could work without steam. At T+10, the temperature of the reactor core, at approximately 285 °C (550 °F) at and before this point, begins to rise as enough coolant has been lost from the core that voids begin to form in the coolant between the fuel rods and they begin to heat rapidly. By T+12 seconds from the accident start, fuel rod uncovery begins. At approximately T+18 areas in the rods have reached 540 °C (1000 °F). Some relief comes at T+20 or so, as the negative temperature coefficient and the negative void coefficient slows the rate of temperature increase. T+25 sees power restored; however, LPCI and CS will not be online until T+40.

At T+40, core temperature is at 650 °C (1200 °F) and rising steadily; CS and LPCI kick in and begins deluging the steam above the core, and then the core itself. First, a large amount of steam still trapped above and within the core has to be knocked down first, or the water will be flashed to steam prior to it hitting the rods. This happens after a few seconds, as the approximately 200,000 L/min (3,300 L/s, 52,500 US gal/min, 875 US gal/s) of water these systems release begin to cool first the top of the core, with LPCI deluging the fuel rods, and CS suppressing the generated steam until at approximately T+100 seconds, all of the fuel is now subject to deluge and the last remaining hot-spots at the bottom of the core are now being cooled. The peak temperature that was attained was 900 °C (1650 °F) (well below the maximum of 1200 °C (2200 °F) established by the NRC) at the bottom of the core, which was the last hot spot to be affected by the water deluge.

The core is cooled rapidly and completely, and following cooling to a reasonable temperature, below that consistent with the generation of steam, CS is shut down and LPCI is decreased in volume to a level consistent with maintenance of a steady-state temperature among the fuel rods, which will drop over a period of days due to the decrease in fission-product decay heat within the core.

After a few days of LPCI, decay heat will have sufficiently abated to the point that defueling of the reactor is able to commence with a degree of caution. Following defueling, LPCI can be shut down. A long period of physical repairs will be necessary to repair the broken recirculation loop; overhaul the ECCS; diesel pumps; and diesel generators; drain the drywell; fully inspect all reactor systems, bring non-conformal systems up to spec, replace old and worn parts, etc. At the same time, different personnel from the licensee working hand in hand with the NRC will evaluate what the immediate cause of the break was; search for what event led to the immediate cause of the break (the root causes of the accident); and then to analyze the root causes and take corrective actions based on the root causes and immediate causes discovered. This is followed by a period to generally reflect and post-mortem the accident, discuss what procedures worked, what procedures didn't, and if it all happened again, what could have been done better, and what could be done to ensure it doesn't happen again; and to record lessons learned to propagate them to other BWR licensees. When this is accomplished, the reactor can be refueled, resume operations, and begin producing power once more.

The ABWR and ESBWR, the most recent models of the BWR, are not vulnerable to anything like this incident in the first place, as they have no liquid penetrations (pipes) lower than several feet above the waterline of the core, and thus, the reactor pressure vessel holds in water much like a deep swimming pool in the event of a feedwater line break or a steam line break. The BWR 5s and 6s have additional tolerance, deeper water levels, and much faster emergency system reaction times. Fuel rod uncovery will briefly take place, but maximum temperature will only reach 600 °C (1,100 °F), far below the NRC safety limit.

It must be noted that no incident even approaching the DBA or even a LBLOCA in severity has ever occurred with a BWR. There have been minor incidents involving the ECCS, but in these circumstances it has performed at or beyond expectations. The most severe incident that ever occurred with a BWR was in 1975 due to a fire caused by extremely flammable urethane foam installed in the place of fireproofing materials at the Browns Ferry Nuclear Power Plant; for a short time, the control room's monitoring equipment was cut off from the reactor, but the reactor shut down successfully, and, as of 2009, is still producing power for the Tennessee Valley Authority, having sustained no damage to systems within the containment. The fire had nothing to do with the design of the BWR - it could have occurred in any power plant, and the lessons learned from that incident resulted in the creation of a separate backup control station, compartmentalization of the power plant into fire zones and clearly documented sets of equipment which would be available to shut down the reactor plant and maintain it in a safe condition in the event of a worst case fire in any one fire zone. These changes were retrofitted into every existing US and most Western nuclear power plants and built in to new plants from that point forth.

Evolution of the BWR

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 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. In particular, Samuel Untermyer II, a researcher at INL, 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.

Following this series of tests, GE got involved and collaborated with INL to bring this technology to market. Larger scale tests were conducted through the late 1950s/early/mid-1960s 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 BWRs (BWR/1–BWR/6)

The first generation of production boiling water reactors saw the incremental development of the unique and distinctive features of the BWR, such as 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, and the distinctive general layout of the reactor building, as well as the standardization of reactor control and safety systems. The first 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.

The advanced boiling water reactor (ABWR)

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 mass-production (insofar as any extraordinarily complex system can be mass-produced.)

The ABWR was approved by the U.S. Nuclear Regulatory Commission for production as a standardized design in the early 1990s. Subsequently, numerous ABWRs were built in Japan; these have reportedly performed with distinction, both safely and economically, in Japanese service. One development spurred by the success of the ABWR in Japan is that GE's nuclear energy division merged with Hitachi Corporation's nuclear energy division, forming GE Hitachi, who is now the major worldwide developer of the BWR design.

The simplified boiling water reactor (SBWR)

GE also developed a different concept for a new BWR at the same time as the ABWR, known as the simplified boiling water reactor (SBWR). This smaller (600 MWe per reactor) 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 if a safety-related contingency developed solely through operation of natural forces.

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 NRC, however, it was withdrawn prior to approval; still, the concept remained intriguing to GE's designers, and served as the basis of future developments.

The economic simplified boiling water reactor (ESBWR)

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 1600 MWe (4500 MWth). This design has been submitted to the U.S. Nuclear Regulatory Commission for approval, and the subsequent Final Design Review is near completion.

Reportedly, this design has a best-in-class core damage probability of 3×10−8 core damage events per reactor-year. (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 long as 1×10−5 core damage events per reactor-year.)[3] This extraordinarily low CDP for the ESBWR far exceeds the other large LWRs on the market.

Several reactors of this design are now on order within the United States.

Advantages and disadvantages

Advantages

  • The reactor vessel and associated components operate at a substantially lower pressure (about 75 times atmospheric pressure) compared to a PWR (about 158 times atmospheric pressure).
  • 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.
  • Fewer components due to no steam generators and no pressurizer vessel. (Older BWRs have external recirculation loops, but even this piping is eliminated in modern BWRs, such as the ABWR.)
  • 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.
  • NRC assessments of limiting fault potentials indicate if such a fault occurred, the average BWR would be less likely to sustain core damage than the average PWR due to the robustness and redundancy of the ECCS.
    • Unlike PWRs, BWRs have at least a few steam-turbine driven ECCS systems that can be directly operated by steam produced after a reactor shutdown, and require no electrical power. This results in less dependence on emergency diesel generators—of which there are four—in any event.
  • 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. (The new ESBWR design uses natural circulation.)
  • BWRs do not use boric acid to control fission burn-up, 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.)
  • 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.
  • Due to their single major vendor (GE/Hitachi), the current fleet of BWRs have predictable, uniform designs that, while not completely standardized, generally are very similar to one another. The ABWR/ESBWR designs are completely standardized. Lack of standardization remains a problem with PWRs, as, at least in the United States, there are three design families represented among the current PWR fleet (Combustion Engineering, Westinghouse, and Babcock & Wilcox), with two families having earned distinction in service, while one family has most of the "black sheep", such as they are. Even within these families, there are quite divergent designs.
    • Additional families of PWRs are being introduced. For example, Mitsubishi's APWR, Areva's US-EPR, and Westinghouse's AP1000/AP600 will add diversity and complexity to an already diverse crowd, and possibly cause customers seeking stability and predictability to seek other designs, such as the BWR.
  • BWRs are overrepresented in imports, if the importing nation doesn't 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 not exported, they cause national skill to be developed in PWR design, construction, and operation), or special national aspirations (special national aspirations lead to a marked preference for the CANDU reactor type due to special features of that type). This may be due to the fact that BWRs are ideally suited for peaceful uses like power generation, process/industrial/district heating, and desalinization, due to low cost, simplicity, and safety focus, which come at the expense of larger size and slightly lower thermal efficiency.
    • Sweden is standardized mainly on BWRs.
    • Germany has a large number of BWRs, which are overrepresented in their national fleet compared to the US.
    • Mexico's only two reactors are BWRs.
    • Japan experimented with both PWRs and BWRs, but most builds as of late have been of BWRs, specifically ABWRs (though indigenous PWRs and a rumored indigenously-designed Japanese BWR may give GE a run for its money).
    • In the CEGB open competition in the early 1960s for a standard design for UK 2nd-generation power reactors, the PWR didn't even make it to the final round, which was a showdown between the BWR (preferred for its easily understood design as well as for being predictable and "boring") and the AGCR, a uniquely British design; the indigenous design won, possibly on technical merits, possibly due to the proximity of a general election.

Disadvantages

  • Complex calculations for managing consumption of nuclear fuel during operation due to "two phase (water and steam) fluid flow" in the upper part of the core. This requires more instrumentation in the reactor core. The innovation of computers, however, makes this less of an issue.
  • Much larger pressure vessel than for a PWR of similar power, with correspondingly higher cost. (However, the overall cost is reduced because a modern BWR has no main steam generators and associated piping.)
  • Contamination of the turbine by short-lived activation products. This means that shielding and access control around the steam turbine are required during normal operations due to the radiation levels arising from the steam entering directly from the reactor core. This is a moderately minor concern, as most of the radiation flux is due to Nitrogen - 16, which has a half-life measured in seconds, allowing the turbine chamber to be entered into within minutes of shutdown.
  • Though the present fleet of BWRs are less likely to suffer core damage from the 1 in 100,000 reactor-year limiting fault than the present fleet of PWRs are (due to increased ECCS robustness and redundancy) there have been concerns raised about the pressure containment ability of the as-built, unmodified Mark I containment - that such may be insufficient to contain pressures generated by a limiting fault combined with complete ECCS failure that results in extremely severe core damage. In this double worst-case, 1 in 100,000,000 reactor-year scenario, an unmodified Mark I containment is speculated to allow some degree of radioactive release to occur. However, this is mitigated by the modification of the Mark I containment; namely, the addition of an outgas stack system that, if containment pressure exceeds critical setpoints, will allow the orderly discharge of pressurizing gasses after the gasses pass through activated carbon filters designed to trap radionuclides.
  • Control rods are inserted from below for current BWR designs. There are two available hydraulic power sources that can drive the control rods into the core for a BWR under emergency conditions. There is a dedicated high pressure hydraulic accumulator and also the pressure inside of the reactor pressure vessel available to each control rod. Either the dedicated accumulator (one per rod) or reactor pressure is capable of fully inserting each rod. Most other reactor types use top entry control rods that are held up in the withdrawn position by electromagnets, causing them to fall into the reactor by gravity if power is lost.

Technical and background information

Start-up ("going critical")

Prior to the introduction of the Fine Motion Control Rod Drive with the ABWR and ESBWR, control rod motion could not be controlled in a boiling water reactor with smooth motion, but instead, control rods moved through a series of notched positions with fixed intervals between these positions, though control rods could be controlled individually or in banks.(All or one) So as to assure a smooth start-up, GE developed a set of rules in the 1970s called BPWS (Banked Position Withdrawal Sequence) that help minimize notch worths and aided in starting the reactor using asymmetric control rod withdrawal patterns. (For instance, rather than withdrawing all control rods from 50% to 49%, which was impossible without fine-grained control, 20% of the control rods would be lowered from 50% to 45%, giving an effective level of 49% for the reactor.)

Thermal margins

Several calculated/measured quantities are tracked while operating a BWR:

  • Maximum Fraction Limiting Critical Power Ratio, or MFLCPR;
  • Fraction Limiting Linear Heat Generation Rate, or FLLHGR;
  • Average Planar Linear Heat Generation Rate, or APLHGR;
  • Pre-Conditioning Interim Operating Management Recommendation, or PCIOMR;

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." 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. It is obvious that 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 99.4% of fuel rods in a BWR core will not enter the transition to film boiling in the event of the worst possible plant transient/SCRAM anticipated to occur. 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), 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 represents the quantitative margin necessary in fuel manufacture to prevent pellet-cladding interaction from occurring during BWR startup - the nuclear fuel pellets within the fuel rods swell more than the fuel rod cladding during reactor startup.

List of BWRs

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

Experimental and other BWRs

Experimental and other non-commercial BWRs include:

  • BORAX Experiments - BORAX I, BORAX II, BORAX III
  • SL-1 (permanently shut down following accident in 1961)

Next-generation designs

See also

References and notes

  1. ^ Staff, USNRC Technical Training Center (2002-09-27). GE Technology Manual (R-304B). 3rd (of 8 files) (Revision 0197 ed.). Chattanooga, Tennessee, United States of America: Office for Analysis and Evaluation of Operational Data, U.S. Nuclear Regulatory Commission. p. 2.5.2. http://adamswebsearch.nrc.gov/idmws/ViewDocByAccession.asp?AccessionNumber=ML023020088. Retrieved 2009-11-15.  
  2. ^ Various GE promotional slideshows & ABWR Tier 2 Design Control Document, USNRC
  3. ^ Hinds, David; Maslak, Chris (January 2006). "Next-generation nuclear energy: The ESBWR". Nuclear News (La Grange Park, Illinois, United States of America: American Nuclear Society) 49 (1): 35–40. ISSN 0029-5574. http://www.ans.org/pubs/magazines/nn/docs/2006-1-3.pdf. Retrieved 2009-04-04.  

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