Boiling water reactor safety systems

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Boiling water reactor safety systems are nuclear safety systems constructed within boiling water reactors in order to prevent or mitigate environmental and health hazards in the event of accident or natural disaster.

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".

Safety systems

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", "torus" or "suppression pool". All BWRs utilize a number of safety/relief valves for overpressure, up to 7 of these are a part of the Automatic Depressurization System (ADS) and 18 safety overpressure relief valves on ABWR models, only a few of which have to function to stop the pressure rise of a transient. In addition, the reactor will 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 with the intention that no credible scenario can cause a pressure and power increase that exceeds the systems' capability to quickly shut down the reactor before damage to the fuel or to components containing the reactor coolant can occur. In the limiting case of an ATWS (Anticipated Transient Without Scram) 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 a contingency that disables all of the safety systems, each reactor is surrounded by a containment building consisting of 1.2–2.4 m (3.9–7.9 ft) of steel-reinforced, pre-stressed concrete designed to seal off the reactor from the environment.

However, the containment building does not protect the fuel during the whole fuel cycle. Most importantly, the spent fuel resides long periods of time outside the primary containment. A typical spent fuel storage pool can hold roughly five times the fuel in the core. Since reloads typically discharge one third of a core, much of the spent fuel stored in the pool will have had considerable decay time. But if the pool were to be drained of water, the discharged fuel from the previous two refuelings would still be "fresh" enough to melt under decay heat. However, the zircaloy cladding of this fuel could be ignited during the heatup. The resulting fire would probably spread to most or all of the fuel in the pool. The heat of combustion, in combination with decay heat, would probably drive "borderline aged" fuel into a molten condition. Moreover, if the fire becomes oxygen-starved (quite probable for a fire located in the bottom of a pit such as this), the hot zirconium would rob oxygen from the uranium dioxide fuel, forming a liquid mixture of metallic uranium, zirconium, oxidized zirconium, and dissolved uranium dioxide. This would cause a release of fission products from the fuel matrix quite comparable to that of molten fuel. In addition, although confined, BWR spent fuel pools are almost always located outside of the primary containment. Generation of hydrogen during the process would probably result in an explosion, damaging the secondary containment building. Thus, release to the atmosphere is more likely than for comparable accidents involving the reactor core.

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 supercritical; 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 or 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.
   3. This trip is only active above approximately 1/3 reactor power. Below this amount, the bypass steam system is capable of controlling reactor pressure without causing a reactivity transient in the core.
2. Loss of off-site power (LOOP)
   1. During normal operation, the reactor protection system (RPS) is powered by off-site power
      1. Loss of off-site power would open all relays in the RPS, 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 off-site power.
3. Neutron monitor trips – the purpose of these trips is to ensure an even increase in neutron and thermal power during startup.
   1. Source-range monitor (SRM) or 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 or 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 pre-set 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. Oscillation Power Range Monitor
      1. Prevents reactor power from rapidly oscillating during low flow high power conditions.
4. Low reactor water level:
   1. Loss of coolant contingency (LOCA)
   2. Loss of proper feedwater (LOFW)
   3. Protects the turbine from excessive moisture carryover if water level is below the steam separator and steam dryer stack.
5. High water level (in BWR6 plants)
   1. Prevents flooding of the main steam lines and protects turbine equipment.
   2. Limits the rate of cold water addition to the vessel, thus limiting reactor power increase during over-feed transients.
6. High drywell (primary containment) pressure
   1. Indicative of potential loss of coolant contingency
   2. Also initiates ECCS systems to prepare for core injection once the injection permissives are cleared.
7. Main steam isolation valve closure (MSIV)
   1. Protects from pressure transient in the core causing a reactivity transient
   2. Only triggers for each channel when the valve is greater than 8% closed
   3. One valve may be closed without initiating a reactor trip.
8. 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.
9. Low RPV pressure:
   1. Indicative of a line break in the steam tunnel or other location which does not trigger high drywell pressure
   2. Bypassed when the reactor is not in Run mode to allow for pressurization and cooldown without an automatic scram signal
10. Seismic event
    1. Generally only plants in high seismic areas have this trip enabled.
11. Scram Discharge Volume High
    1. In the event that the scram hydraulic discharge volume begins to fill up, this will scram the reactor prior to the volume filling. This prevents hydraulic lock, which could prevent the control rods from inserting. This is to prevent an ATWS (Anticipated Transient Without Scram).

Emergency core-cooling system (ECCS)

Diagram of a generic BWR reactor pressure vessel

While the reactor protection system is designed to shut down the reactor, ECCS is designed to maintain adequate core cooling. 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. The five criteria for ECCS are to prevent peak fuel cladding temperature from exceeding 2200 °F (1204 °C), prevent more than 17% oxidation of the fuel cladding, prevent more than 1% of the maximum theoretical hydrogen generation due the zircalloy metal-water reaction, maintain a coolable geometry, and allow for long-term cooling.  ECCS 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 three 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 main steam isolation valve 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 maintain reactor pressure within safety limits. Additionally, if reactor water level cannot be maintained with high-pressure coolant systems alone, the depressurization system can reduce reactor pressure to a level at which the low-pressure coolant systems can function.
3. Low-pressure systems: These systems are designed to function after the depressurization systems function. They have 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 automatic depressurization, core spray, and low-pressure coolant injection 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. HPCI is also able to be run in "pressure control mode", where the HPCI turbine is run without pumping water to the reactor vessel. This allows HPCI to remove steam from the reactor and slowly depressurize it without the need for operating the safety or relief valves. This minimizes the number of times the relief valves need to operate, and reduces the potential for one sticking open and causing a small LOCA.

Versioning note: Some BWR/5s and the BWR/6 replace the steam-turbine driven HPCI pump with the AC-powered high-pressure core spray (HPCS); ABWR replaces HPCI with high-pressure core flooder (HPCF), a mode of the RCIC system, as described below. (E)SBWR does not have an equivalent system as it primarily uses passive safety cooling systems, though ESBWR does offer an alternative active high-pressure injection method using an operating mode of the Control Rod Drive System (CRDS) to supplement the passive system.

Isolation Condenser (IC)

See also: Isolation Condensor

Some reactors, including some BWR/2 and BWR/3 plants, and the (E)SBWR series of reactors, have a passive system called the Isolation Condenser. This is a heat exchanger located above containment in a pool of water open to atmosphere. When activated, decay heat boils steam, which is drawn into the heat exchanger and condensed; then it falls by weight of gravity back into the reactor. This process keeps the cooling water in the reactor, making it unnecessary to use powered feedwater pumps. The water in the open pool slowly boils off, venting clean steam to the atmosphere. This makes it unnecessary to run mechanical systems to remove heat. Periodically, the pool must be refilled, a simple task for a fire truck. The (E)SBWR reactors provide three days' supply of water in the pool. Some older reactors also have IC systems, including Fukushima Dai-ichi reactor 1, however their water pools may not be as large.

Under normal conditions, the IC system is not activated, but the top of the IC condenser is connected to the reactor's steam lines through an open valve. The IC automatically starts on low water level or high steam pressure indications. Once it starts, steam enters the IC condenser and condenses until it is filled with water. When the IC system is activated, a valve at the bottom of the IC condenser is opened which connects to a lower area of the reactor. The water falls to the reactor by gravity, allowing the condenser to fill with steam, which then condenses. This cycle runs continuously until the bottom valve is closed.

Reactor core isolation cooling system (RCIC)

The reactor core isolation cooling system is not an emergency core cooling system proper, but it is included because it fulfills an important-to-safety function which can help to cool the reactor in the event of a loss of normal heat sinking capability; or when all electrical power is lost. It has additional functionality in advanced versions of the BWR.

RCIC is an auxiliary feedwater pump meant for emergency use. It is able to inject cooling water into the reactor at high pressures. It injects approximately 2,000 L/min (600 gpm) into the reactor core. It takes less time to start than the HPCI system, approximately 30 seconds from an initiating signal. It has ample capacity to replace the cooling water boiled off by residual decay heat, and can even keep up with small leaks.

The RCIC system operates on high-pressure steam from the reactor itself, and thus is operable with no electric power other than battery power to operate the control valves. Those turn the RCIC on and off as necessary to maintain correct water levels in the reactor. (If run continuously, the RCIC would overfill the reactor and send water down its own steam supply line.) During a station blackout (where all off-site power is lost and the diesel generators fail) the RCIC system may be "black started" with no AC and manually activated. The RCIC system condenses its steam into the reactor suppression pool. The RCIC can make up this water loss, from either of two sources: a makeup water tank located outside containment, or the wetwell itself. RCIC is not designed to maintain reactor water level during a LOCA or other leak. Similar to HPCI, the RCIC turbine can be run in recirculation mode to remove steam from the reactor and help depressurize the reactor. 

Versioning note: RCIC and HPCF are integrated in the ABWRs, with HPCF representing the high-capacity mode of RCIC. Older BWRs such as Fukushima Unit 1 and Dresden as well as the new (E)SBWR do not have a RCIC system, and instead have an Isolation Condenser system.

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 there is either a loss of high-pressure cooling to the vessel or if the high-pressure cooling systems cannot maintain the RPV water level. ADS can be manually or automatically initiated. When ADS receives an auto-start signal when water reaches the Low-Low-Low Water Level Alarm setpoint. ADS then confirms with the Low Alarm Water Level, verifies at least 1 low-pressure cooling pump is operating, and starts a 105-second timer. When the timer expires, or when the manual ADS initiate buttons are pressed, the system 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) to restore reactor water level. During an ADS blowdown, the steam being removed from the reactor is sufficient to ensure adequate core cooling even if the core is uncovered. The water in the reactor will rapidly flash to steam as reactor pressure drops, carrying away the latent heat of vaporization and providing cooling for the entire reactor. Low pressure ECCS systems will re-flood the core prior to the end of the emergency blowdown, ensuring that the core retains adequate cooling during the entire event.

Low-pressure core spray system (LPCS)

The Core Spray system, or Low-Pressure Core Spray system is designed to suppress steam generated by a major contingency and to ensure adequate core cooling for a partially or fully uncovered reactor core. LPCS can deliver up to 48,000 L/min (12,500 US gal/min) of water in a deluge from the top of the core. The core spray system collapses steam voids above the core, aids in reducing reactor pressure when the fuel is uncovered, and, in the event the reactor has a break so large that water level cannot be maintained, core spray is capable of preventing fuel damage by ensuring the fuel is adequately sprayed to remove decay heat. In earlier versions of the BWR (BWR 1 or 2 plants), the LPCS system was the only ECCS, and the core could be adequately cooled by core spray even if it was completely uncovered. Starting with Dresden units 2 and 3, the core spray system was augmented by the HPCI/LPCI systems to provide for both spray cooling and core flooding as methods for ensuring adequate core cooling. For most BWR models, core spray ensures the upper 1/3rd of the core does not exceed 17% cladding oxidation or 1% hydrogen production during a LOCA when used in combination with the LPCI system.

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 (LPCI)

Low-pressure coolant injection is the emergency injection mode of the Residual Heat Removal (RHR) system. LPCI can be operated at reactor vessel pressures below 375 psi. LPCI consists of several pumps which are capable of injecting up to 150,000 L/min (40,000 US gal/min) of water into the reactor. Combined with the Core Spray system, the LPCI is designed to rapidly flood the reactor with coolant. The LPCI system was first introduced with Dresden units 2 and 3. The LPCI system can also use the RHR heat exchangers to remove decay heat from the reactor and cool the containment to cold conditions. Early versions of the LPCI system injected through the recirculation loops or into the down comer. Later versions of the BWR moved the injection point directly inside the core shroud to minimize time to reflood the core, substantially reducing the peak temperatures of the reactor during a LOCA.

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 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 wetwell.

If Level 1 is not resubmerged within 50 seconds after the countdown started, DPVS fires and rapidly vents steam 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, depressurization reduces the saturation temperature enhancing the heat removal via phase transition. (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 still 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.

Standby liquid control system (SLCS)

The SLCS is a backup to the reactor protection system. In the event that RPS is unable to scram the reactor for any reason, the SLCS will inject a liquid boron solution into the reactor vessel to bring it to a guaranteed shutdown state prior to exceeding any containment or reactor vessel limits. The standby liquid control system is designed to deliver the equivalent of 86 gpm of 13% by weight sodium pentaborate solution into a 251-inch BWR reactor vessel. SLCS, in combination with the alternate rod insertion system, the automatic recirculation pump trip and manual operator actions to reduce water level in the core will ensure that the reactor vessel does not exceed its ASME code limits, the fuel does not suffer core damaging instabilities, and the containment does not fail due to overpressure during high power scram failure.

The SLCS consists of a tank containing borated water as a neutron absorber, protected by explosively-opened valves and redundant pumps, allowing the injection of the borated water into the reactor against any pressure within; the borated water will shut down a reactor and maintain it shut down. The SLCS can also be injected during a LOCA or a fuel cladding failure to adjust the ph of the reactor coolant that has spilled, preventing the release of some radioactive materials.

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 reluctant about activating the SLCS, as these reactors have a reactor water cleanup system (RWCS) which is designed to remove boron – 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 Zircaloy shielding;
2. The reactor pressure vessel itself is manufactured out of 6-inch-thick (150 mm) steel, with extremely high temperature, vibration, and corrosion resistant surgical stainless steel 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 (3.9–7.9 ft) thick.
5. The reactor building (the shield wall/missile shield) is also made of steel-reinforced, pre-stressed concrete 0.3 to 1 m (0.98 to 3.28 ft) 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:

Garigliano Nuclear Power Plant, using the premodern "dry" containment

• 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.

Mark I Containment

Mark I Containment under construction at Browns Ferry Nuclear Plant unit 1. In the foreground is the lid of the drywell or primary containment vessel (PCV).

Schematic BWR inside Mark I containment.

• 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 structure of reinforced concrete.

BWR inside a Mark II containment.

• 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 PWR-like 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.

ESBWR Containment

• 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 stress than previous containments were, providing advanced safety. In particular, GE regards these containments as being able to withstand a direct hit by a tornado beyond Level 5 on the Old Fujita Scale with winds of 330+ miles per hour. Such a tornado has never been measured on earth. They are also designed to withstand seismic accelerations of .2 G, or nearly 2 meters per second² in any direction.

Containment Isolation System

Many valves passing in and out of the containment are required to be open to operate the facility. During an accident where radioactive material may be released, these valves must shut to prevent the release of radioactive material or the loss of reactor coolant. The containment isolation system is responsible for automatically closing these valves to prevent the release of radioactive material and is an important part of a plant's safety analysis. The isolation system is separated into groups for major system functions. Each group contains its own criteria to trigger an isolation. The isolation system is similar to reactor protection system in that it consists of multiple channels, it is classified as safety-related, and that it requires confirmatory signals from multiple channels to issue an isolation to a system. An example of parameters which are monitored by the isolation system include containment pressure, acoustic or thermal leak detection, differential flow, high steam or coolant flow, low reactor water level, or high radiation readings in the containment building or ventilation system. These isolation signals will lock out all of the valves in the group after closing them and must have all signals cleared before the lockout can be reset.

Isolation valves consist of 2 safety-related valves in series. One is an inboard valve, the other is an outboard valve. The inboard is located inside the containment, and the outboard is located just outside the containment. This provides redundancy as well as making the system immune to the single failure of any inboard or outboard valve operator or isolation signal. When an isolation signal is given to a group, both the inboard and outboard valves stroke closed. Tests of isolation logic must be performed regularly and is a part of each plant's technical specifications. The timing of these valves to stroke closed is a component of each plant's safety analysis and failure to close in the analyzed time is a reportable event.

Examples of isolation groups include the main steamlines, the reactor water cleanup system, the reactor core isolation cooling (RCIC) system, shutdown cooling, and the residual heat removal system. For pipes which inject water into the containment, two safety-related check valves are generally used in lieu of motor operated valves. These valves must be tested regularly as well to ensure they do indeed seal and prevent leakage even against high reactor pressures.

Hydrogen management

During normal plant operations and in normal operating temperatures, the hydrogen generation is not significant. When the nuclear fuel overheats, zirconium in Zircaloy cladding used in fuel rods oxidizes in reaction with steam:

Zr + 2H₂O → ZrO₂ + 2H₂

When mixed with air, hydrogen is flammable, and hydrogen detonation or deflagration may damage the reactor containment. In reactor designs with small containment volumes, such as in Mark I or II containments, the preferred method for managing hydrogen is pre-inerting with inert gas—generally nitrogen—to reduce the oxygen concentration in air below that needed for hydrogen combustion, and the use of thermal recombiners. Pre-inerting is considered impractical with larger containment volumes where thermal recombiners and deliberate ignition are used. Mark III containments have hydrogen igniters and hydrogen mixers which are designed to prevent the buildup of hydrogen through either pre-ignition prior to exceeding the lower explosive limit of 4%, or through recombination with Oxygen to make water.

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 reasonably expect. 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.

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 uninterruptible 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 likely collapse to liquid 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 (545 °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 (1,004 °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 (1,202 °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 (1,650 °F) (well below the maximum of 1,200 °C (2,190 °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 (1,112 °F), far below the NRC safety limit.

According to a report by the U.S. Nuclear Regulatory Commission into the Fukushima Daiichi nuclear disaster, the March 2011 Tōhoku earthquake and tsunami that caused that disaster was an event "far more severe than the design basis for the Fukushima Daiichi Nuclear Power Plant". The reactors at this plant were BWR 3 and BWR 4 models. Their primary containment vessels had to be flooded with seawater containing boric acid, which will preclude any resumption of operation and was not anticipated in the DBA scenario. In addition, nothing similar to the chemical explosions that occurred at the Fukushima Daiichi plant was anticipated by the DBA.

Prior to the Fukushima Daiichi disaster, no incident approaching the DBA or even a LBLOCA in severity had occurred with a BWR. There had been minor incidents involving the ECCS, but in those circumstances it had performed at or beyond expectations. The most severe incident that had previously 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 into new plants from that point forth.

Notable activations of BWR safety systems

General Electric defended the design of the reactor, stating that the station blackout caused by the 2011 Tōhoku earthquake and tsunami was a "beyond-design-basis" event which led to Fukushima I nuclear accidents. According to the Nuclear Energy Institute, "Coincident long-term loss of both on-site and off-site power for an extended period of time is a beyond-design-basis event for the primary containment on any operating nuclear power plant".

The reactors shut down as designed after the earthquake. However, the tsunami disabled four of the six sets of switchgear and all but three of the diesel backup generators which operated the emergency cooling systems and pumps. Pumps were designed to circulate hot fluid from the reactor to be cooled in the wetwell, but only units 5 and 6 had any power. Units 1, 2 and 3 reactor cores overheated and melted. Radioactivity was released into the air as fuel rods were damaged due to overheating by exposure to air as water levels fell below safe levels. As an emergency measure, operators resorted to using firetrucks and salvaged car batteries to inject seawater into the drywell to cool the reactors, but only achieved intermittent success and three cores overheated. Reactors 1–3, and by some reports 4 all suffered violent hydrogen explosions March 2011 which damaged or destroyed their top levels or lower suppression level (unit 2).

As emergency measures, helicopters attempted to drop water from the ocean onto the open rooftops. Later water was sprayed from fire engines onto the roof of reactor 3. A concrete pump was used to pump water into the spent fuel pond in unit 4.

According to NISA, the accident released up to 10 petabecquerels of radioactive iodine-131 per hour in the initial days, and up to 630 PBq total, about one eighth the 5200 PBq released at Chernobyl.

Generation III reactor

From Wikipedia, the free encyclopedia
https://en.wikipedia.org/wiki/Generation_III_reactor

Model of the Toshiba ABWR, which became the first operational Generation III reactor in 1996

Generation III reactors, or Gen III reactors, are a class of nuclear reactors designed to succeed Generation II reactors, incorporating evolutionary improvements in design. These include improved fuel technology, higher thermal efficiency, significantly enhanced safety systems (including passive nuclear safety), and standardized designs intended to reduce maintenance and capital costs. They are promoted by the Generation IV International Forum (GIF).

The first Generation III reactors to begin operation were Kashiwazaki 6 and 7 advanced boiling water reactors (ABWRs) in 1996 and 1997. From 2012, both have been shut down due to a less permissive political environment in the wake of the Fukushima nuclear accident. Due to the prolonged period of stagnation in the construction of new reactors and the continued (albeit declining) popularity of Generation II/II+ designs in new construction, relatively few third generation reactors have been built.

Overview

The older Gen II reactors comprise the vast majority of current nuclear reactors. Gen III reactors are so-called advanced light-water reactors (LWRs). Gen III+ reactors are labeled as "evolutionary designs". Though the distinction between Gen II and III reactors is arbitrary, few Gen III reactors have reached the commercial stage as of 2022. The Generation IV International Forum calls Gen IV reactors "revolutionary designs". These are concepts for which no concrete prognoses for realization existed at the time. The improvements in reactor technology in third generation reactors are intended to result in a longer operational life (designed for 60 years of operation, extendable to 100+ years of operation prior to complete overhaul and reactor pressure vessel replacement) compared with currently used Generation II reactors (designed for 40 years of operation, extendable to 60+ years of operation prior to complete overhaul and pressure vessel replacement).

The core damage frequencies for these reactors are designed to be lower than for Generation II reactors – 60 core damage events for the European Pressurized Reactor (EPR) and 3 core damage events for the Economic Simplified Boiling Water Reactor (ESBWR) per 100 million reactor-years are significantly lower than the 1,000 core damage events per 100 million reactor-years for BWR/4 Generation II reactors.

The third generation EPR reactor was also designed to use uranium more efficiently than older Generation II reactors, using approximately 17% less per unit of electricity generated than these older reactor technologies. An independent analysis conducted by environmental scientist Barry Brook on the greater efficiency and therefore lower material needs of Gen III reactors, corroborates this finding.

Developments

EPR core catching room designed to catch the corium in case of a meltdown. Some Generation III reactors include a core catcher in their design.

Gen III+ reactor designs are an evolutionary development of Gen III reactors, offering improvements in safety over Gen III reactor designs. Manufacturers began development of Gen III+ systems in the 1990s by building on the operating experience of the American, Japanese, and Western European light-water reactor.

The nuclear industry began to promote a nuclear renaissance suggesting that Gen III+ designs should solve three key problems: safety, cost and buildability. Construction costs of US$1,000/kW were forecast, a level that would make nuclear competitive with gas, and construction times of four years or less were expected. However, these estimates proved over-optimistic.

A notable improvement of Gen III+ systems over second-generation designs is the incorporation in some designs of passive safety features that do not require active controls or operator intervention but instead rely on gravity or natural convection to mitigate the impact of abnormal events.

Kakrapar Atomic Power Station Unit 3 and 4 under construction. India's first Generation III+ reactor

Generation III+ reactors incorporate extra safety features to avoid the kind of disaster suffered at Fukushima in 2011. Generation III+ designs, passive safety, also known as passive cooling, requires no sustained operator action or electronic feedback to shut down the plant safely in the event of an emergency. Many of the Generation III+ nuclear reactors have a core catcher. If the fuel cladding and reactor vessel systems and associated piping become molten, corium will fall into a core catcher which holds the molten material and has the ability to cool it. This, in turn protects the final barrier, the containment building. As an example, Rosatom installed a 200-tonne core catcher in the VVER reactor as the first large piece of equipment in the reactor building of Rooppur 1, describing it as "a unique protection system". In 2017, Rosatom has started commercial operations of the NVNPP-2 Unit 1 VVER-1200 reactor in central Russia, marking the world's first full start-up of a generation III+ reactor.

First reactors

Novovoronezh Nuclear Power Plant II with the first Generation III+ nuclear reactor in the world

The first Generation III reactors were built in Japan, in the form of advanced boiling water reactors. On 5 August 2016, a Generation III+ VVER-1200/392M reactor became operational (first grid connection) at Novovoronezh Nuclear Power Plant II in Russia, which was the first operational Generation III+ reactor.

Several other Generation III+ reactors are under late-stage construction in Europe, China, India, and the United States. The next Generation III+ reactors to come online were an AREVA EPR reactor at the Taishan Nuclear Power Station (first grid connection on 2018-06-29) and a Westinghouse AP1000 reactor at the Sanmen Nuclear Power Station (first grid connection on 2018-06-30) in China.

In the United States, reactor designs are certified by the Nuclear Regulatory Commission (NRC). As of August 2020, the commission has approved seven new designs, and is considering one more design as well as renewal of an expired certification.

Response and criticism

Proponents of nuclear power and some who have historically been critical have acknowledged that third generation reactors as a whole are safer than older reactors.

Edwin Lyman, a senior staff scientist at the Union of Concerned Scientists, has challenged specific cost-saving design choices made for two Generation III reactors, both the AP1000 and ESBWR. Lyman, John Ma (a senior structural engineer at the NRC), and Arnold Gundersen (an anti-nuclear consultant) are concerned about what they perceive as weaknesses in the steel containment vessel and the concrete shield building around the AP1000 in that its containment vessel does not have sufficient safety margins in the event of a direct airplane strike. Other engineers do not agree with these concerns, and claim the containment building is more than sufficient in safety margins and factors of safety.

The Union of Concerned Scientists in 2008 referred to the EPR as the only new reactor design under consideration in the United States that "...appears to have the potential to be significantly safer and more secure against attack than today's reactors."

There have also been issues in fabricating the precision parts necessary to maintain safe operation of these reactors, with cost overruns, broken parts, and extremely fine steel tolerances causing issues with new reactors under construction in France at the Flamanville Nuclear Power Plant.

Lists of Generation III reactors

Generation III reactors currently operational or under construction

Generation III designs not adopted or built yet

Lists of Generation III+ reactors

Generation III+ reactors currently operational or under construction

Generation III+ designs not adopted or built yet

Developer(s)	Reactor name(s)	Type	MWe (net)	MWe (gross)	MWth	Notes
General Electric, Toshiba, Hitachi	ABWR; US-ABWR	BWR	1350	1420	3926	In operation at Kashiwazaki since 1996. NRC certified in 1997.
KEPCO	APR-1400	PWR	1383	1455	3983	In operation at Kori since Jan 2016.
CGNPG	ACPR-1000		1061	1119	2905	Improved version of the CPR-1000. The first reactor came online in 2018 at Yangjiang-5.
CGNPG, CNNC	Hualong One (HPR-1000)		1090	1170	3050	In part a merger of the Chinese ACPR-1000 and ACP-1000 designs, but ultimately an incrementally developed improvement on the prior CNP-1000 and CP-1000 designs. It was initially intended to be named the "ACC-1000", but was ultimately named as the "Hualong One" or "HPR-1000". Fangchenggang Units 3–6 will be the first to utilize the HPR-1000 design, with Units 3 & 4 currently under construction as of 2017.
OKBM Afrikantov	VVER-1000/428		990	1060	3000	First version of the AES-91 design, designed and used for Tianwan Units 1 & 2, which came online in 2007.
	VVER-1000/428M		1050	1126	3000	Another version of the AES-91 design, also designed and used for Tianwan (this time for Units 3 & 4, which came online in 2017 and 2018, respectively).
	VVER-1000/412		917	1000	3000	First constructed AES-92 design, used for the Kudankulam.
Developer(s)	Reactor name(s)	Type	MWe (net)	MWe (gross)	MWth	Notes
General Electric, Hitachi	ABWR-II	BWR	1638	1717	4960	Improved version of the ABWR. Uncertain development status.
Mitsubishi	APWR; US-APWR; EU-APWR; APWR+	PWR	1600	1700	4451	Two units planned at Tsuruga cancelled in 2011. US NRC licensing for two units planned at Comanche Peak was suspended in 2013. The original APWR and the updated US-APWR/EU-APWR (also known as the APWR+) differ significantly in their design characteristics, with the APWR+ having higher efficiency and electrical output.
Westinghouse	AP600		600	619	?	NRC certified in 1999. Evolved into the larger AP1000 design.
Combustion Engineering	System 80+		1350	1400	?	NRC certified in 1997. Provided a basis for the Korean APR-1400.
OKBM Afrikantov	VVER-1000/466(B)		1011	1060	3000	This was the first AES-92 design to be developed, originally intended to be built at the proposed Belene Nuclear Power Plant, but construction was later halted.
Candu Energy Inc.	EC6	PHWR	?	750	2084	The EC6 (Enhanced CANDU 6) is an evolutionary upgrade of previous CANDU designs. Like other CANDU designs, it is capable of using unenriched natural uranium as fuel.
	AFCR		?	740	2084	The Advanced Fuel CANDU Reactor is a modified EC6 design that has been optimized for extreme fuel flexibility with the ability to handle numerous potential reprocessed fuel blends and even thorium. It is currently undergoing late-stage development as part of a joint venture between SNC-Lavalin, CNNC, and Shanghai Electric.
Various (see MKER Article.)	MKER	BWR	1000	?	2085	A Development of the RBMK nuclear power reactor. Fixes all of the RBMK reactor's design errors and flaws and adds a full containment building and Passive nuclear safety features such as a passive core cooling system. The physical prototype of the MKER-1000 is the 5th unit of the Kursk Nuclear Power Plant. The construction of Kursk 5 was cancelled in 2012 and a VVER-TOI whose construction is ongoing since 2018 is being built instead as of 2018.
Developer(s)	Reactor name(s)	Type	MWe (net)	MWe (gross)	MWth	First grid connection	Notes
Westinghouse, Toshiba	AP1000	PWR	1117	1250	3400	2018-06-30 Sanmen	NRC certified Dec 2005.
SNPTC, Westinghouse	CAP1400		1400	1500	4058		The first Chinese co-developed and upsized "native" version/derivative of the AP1000. Westinghouse's co-development agreement gives China the IP rights for all co-developed plants >1350 MWe. First two units currently under construction at Shidao Bay. The CAP1400 is planned to be followed by a CAP1700 and/or a CAP2100 design if the cooling systems can be scaled up by far enough.
Areva	EPR		1660	1750	4590	2018-06-29 Taishan
OKB Gidropress	VVER-1200/392M		1114	1180	3200	2016-08-05 Novovoronezh II	The VVER-1200 series is also known as the AES-2006/MIR-1200 design. This particular model was the original reference model used for the VVER-TOI project.
	VVER-1200/491		1085	1199	3200	2018-03-09 Leningrad II
	VVER-1200/509		1114	1200	3200		Under construction in Akkuyu NPP, as Akkuyu 1 & 2. Grid connections due 2023 & 2024.
	VVER-1200/523		1080	1200	3200		2.4 GWe Rooppur Nuclear Power Plant of Bangladesh is under construction.The two units of VVER- 1200/523 generating 2.4 GWe are planned to be operational in 2023 and 2024.
	VVER-1200/513		?	1200	3200		Standardized version of the VVER-1200 based in part on the VVER-1300/510 design (which is the current reference design for the VVER-TOI project). First unit expected to be completed by 2022 at Akkuyu, as Akkuyu 3.
	VVER-1300/510		1115	1255	3300		The VVER-1300 design is also known as the AES-2010 design, and is sometimes mistakenly designated as the VVER-TOI design. The VVER-1300/510 is based on the VVER-1200/392M that was originally used as the reference design for the VVER-TOI project, although the VVER-1300/510 now serves that role (which has led to confusion between the VVER-TOI plant design and the VVER-1300/510 reactor design). Multiple units are currently planned for construction at several Russian nuclear plants. First units under construction at Kursk Nuclear Power Plant.
BARC	IPHWR-700	PHWR	630	700	2166	2021	Successor of indigenous 540MWe PHWR with increased output and additional safety features. Under construction and due to come online in 2020. Unit 3 at Kakrapar Atomic Power Station achieved first criticality on 22 July 2020. The Unit 3 was connected to the grid on 10 January 2021.
Developer(s)	Reactor name(s)	Type	MWe (net)	MWe (gross)	MWth	Notes
Toshiba	EU-ABWR	BWR	?	1600	4300	Updated version of the ABWR designed to meet EU guidelines, increase reactor output, and improve design generation to III+.
Areva	Kerena		1250	1290	3370	Previously known as the SWR-1000. Based on German BWR designs, mainly that of Gundremmingen units B/C. Co-developed by Areva and E.ON.
General Electric, Hitachi	ESBWR		1520	1600	4500	Based on the unreleased SBWR design which in turn was based on the ABWR. Being considered for North Anna-3. Eschews the use of recirculation pumps entirely in favor of a design completely reliant on natural circulation (which is very unusual for a boiling water reactor design).
KEPCO	APR+	PWR	1505	1560	4290	APR-1400 successor with increased output and additional safety features.
Areva, Mitsubishi	ATMEA1		1150	?	3150	Proposed Sinop plant did not proceed
OKB Gidropress	VVER-600/498		?	600	1600	Essentially a scaled-down VVER-1200. Commercial deployment planned by 2030 at Kola.
Candu Energy Inc.	ACR-1000	PHWR	1085	1165	3200	The Advanced CANDU Reactor is a hybrid CANDU design that retains the heavy water moderator but replaces the heavy water coolant with conventional light water coolant, significantly reducing heavy water costs compared to traditional CANDU designs but losing the characteristic CANDU capability of using unenriched natural uranium as fuel.
BARC	IPWR-900	PWR	900	?	2700	India's first light water reactor, a Gen 3+ design based on the CLWR-B1 reactor of Arihant-class submarine.

Supercritical water reactor

From Wikipedia, the free encyclopedia
https://en.wikipedia.org/wiki/Supercritical_water_reactor

Supercritical water reactor scheme.

The supercritical water reactor (SCWR) is a concept Generation IV reactor, designed as a light water reactor (LWR) that operates at supercritical pressure (i.e. greater than 22.1 megapascals [3,210 psi]). The term critical in this context refers to the critical point of water, and should not be confused with the concept of criticality of the nuclear reactor.

The water heated in the reactor core becomes a supercritical fluid above the critical temperature of 374 °C (705 °F), transitioning from a fluid more resembling liquid water to a fluid more resembling saturated steam (which can be used in a steam turbine), without going through the distinct phase transition of boiling.

In contrast, the well-established pressurized water reactors (PWR) have a primary cooling loop of liquid water at a subcritical pressure, transporting heat from the reactor core to a secondary cooling loop, where the steam for driving the turbines is produced in a boiler (called the steam generator). Boiling water reactors (BWR) operate at even lower pressures, with the boiling process to generate the steam happening in the reactor core.

The supercritical steam generator is a proven technology.

The development of SCWR systems is considered a promising advancement for nuclear power plants because of its high thermal efficiency (~45 % vs. ~33 % for current LWRs) and simpler design. As of 2012 the concept was being investigated by 32 organizations in 13 countries.

History

The super-heated steam cooled reactors operating at subcritical-pressure were experimented with in both Soviet Union and in the United States as early as the 1950s and 1960s such as Beloyarsk Nuclear Power Station, Pathfinder and Bonus of GE's Operation Sunrise program. These are not SCWRs. SCWRs were developed from the 1990s onwards. Both a LWR-type SCWR with a reactor pressure vessel and a CANDU-type SCWR with pressure tubes are being developed.

A 2010 book includes conceptual design and analysis methods such as core design, plant system, plant dynamics and control, plant startup and stability, safety, fast reactor design etc.

A 2013 document saw the completion of a prototypical fueled loop test in 2015. A Fuel Qualification Test was completed in 2014.

A 2014 book saw reactor conceptual design of a thermal spectrum reactor (Super LWR) and a fast reactor (Super FR) and experimental results of thermal hydraulics, materials and material-coolant interactions.

Design

Moderator-coolant

The SCWR operates at supercritical pressure. The reactor outlet coolant is supercritical water. Light water is used as a neutron moderator and coolant. Above the critical point, steam and liquid become the same density and are indistinguishable, eliminating the need for pressurizers and steam generators (PWR), or jet/recirculation pumps, steam separators and dryers (BWR). Also, by avoiding boiling, SCWR does not generate chaotic voids (bubbles) with less density and moderating effect. In a LWR this can affect heat transfer and water flow, and the feedback can make the reactor power harder to predict and control. Neutronic and thermal hydraulic coupled calculation is needed to predict the power distribution. SCWR's simplification should reduce construction costs and improve reliability and safety.

A LWR type SCWR adopts water rods with thermal insulation and a CANDU type SCWR keeps water moderator in a Calandria tank. A fast reactor core of the LWR type SCWR adopts tight fuel rod lattice as a high conversion LWR. The fast neutron spectrum SCWR has advantages of a higher power density, but needs plutonium and uranium mixed oxides fuel which will be available from reprocessing.　

Control

SCWRs would likely have control rods inserted through the top, as is done in PWRs.

Material

The temperature inside an SCWR is higher than those in LWRs. Although supercritical fossil fuel plants have much experience in the materials, it does not include the combination of high temperature environment and intense neutron radiation. SCWRs need core materials (especially fuel cladding) to resist the environment. R&D focuses on:

• The chemistry of supercritical water under radiation (preventing stress corrosion cracking, and maintaining corrosion resistance under neutron radiation and high temperatures)
• Dimensional and microstructural stability (preventing embrittlement, retaining strength and creep resistance also under radiation and high temperatures)
• Materials that both resist the high temperature conditions and do not absorb too many neutrons, which affects fuel economy

In the once-through coolant cycles, such as SCWRs and supercritical fossil fired power plants, the entire reactor coolant is processed at low temperature after condensation. It is an advantage in managing water chemistry and stress corrosion cracking of structural materials. It is not possible in LWRs due to the recirculation of hot reactor coolant. Materials and water chemistry R&D should be done with the once-through characteristics in mind.

Advantages

• Supercritical water has excellent heat transfer properties allowing a high-power density, a small core, and a small containment structure.
• The use of a supercritical Rankine cycle with its typically higher temperatures improves efficiency (would be ~45 % versus ~33 % of current PWR/BWRs).
• This higher efficiency would lead to better fuel economy and a lighter fuel load, lessening residual (decay) heat.
• SCWR is typically designed as a direct cycle, whereby steam or hot supercritical water from the core is used directly in a steam turbine. This makes the design simple. As a BWR is simpler than a PWR, a SCWR is a lot simpler and more compact than a less-efficient BWR having the same electrical output. There are no steam separators, steam dryers, internal recirculation pumps, or recirculation flow inside the pressure vessel. The design is a once-through, direct-cycle, the simplest type of cycle possible. The stored thermal and radiologic energy in the smaller core and its (primary) cooling circuit would also be less than that of either a BWR's or a PWR's.
• Water is liquid at room temperature, cheap, non-toxic and transparent, simplifying inspection and repair (compared to liquid metal cooled reactors).
• A fast SCWR could be a breeder reactor, like the proposed Clean and Environmentally Safe Advanced Reactor and could burn the long-lived actinide isotopes.
• A heavy-water SCWR could breed fuel from thorium (4x more abundant than uranium). Similar to a CANDU it could also use unenriched natural uranium if enough moderation is provided
• Process heat can be delivered at higher temperatures than other water-cooled reactors allow

Disadvantages

• Lower water inventory (due to compact primary loop) means less heat capacity to buffer transients and accidents (e.g., loss of feedwater flow or large break loss-of-coolant accident) resulting in accident and transient temperatures that are too high for conventional metallic cladding.

However, it is not too high for stainless steel cladding. Safety analysis of LWR type SCWR showed that safety criteria are met with margins at accidents and abnormal transients including total loss of flow and loss of coolant accident. No double ended break occurs because of the once-through coolant cycle. Core is cooled by the induced flow at the loss of coolant accident. The water inventory in the top dome of the reactor vessel serves as an in-vessel accumulator. The SCWR safety principle is not to maintain coolant inventory, but to maintain core coolant flow rate. It is easier to monitor than water level at accidents. There was an error in the water level signal in the Three Mile Island accident and the operators shut down the ECCS.

• Higher pressure combined with higher temperature and also a higher temperature rises across the core (compared to PWR/BWRs) result in increased mechanical and thermal stresses on vessel materials that are difficult to solve.

However, a LWR type design, reactor pressure vessel inner wall is cooled by the inlet coolant as PWR. Outlet coolant nozzles are equipped with thermal sleeves. A pressure-tube design, where the core is divided up into smaller tubes for each fuel channel, has potentially fewer issues here, as smaller diameter tubing can be much thinner than massive single pressure vessels, and the tube can be insulated on the inside with inert ceramic insulation so it can operate at low (calandria water) temperature.

• The coolant greatly reduces its density at the end of the core, resulting in a need to place extra moderator there.

However, a LWR type SCWR design adopts water rods in the fuel assemblies as BWRs. The coolant density in water rods is kept high with thin thermal insulation, not fully insulated. Most designs of CANDU type SCWR use an internal calandria where part of the feedwater flow is guided through top tubes through the core, that provide the added moderation (feedwater) in that region. This has the added advantage of being able to cool the entire vessel wall with feedwater, but results in a complex and materially demanding (high temperature, high temperature differences, high radiation) internal calandria and plena arrangement. A pressure-tube design has the characteristics as most of the moderator is in the calandria at low temperature and pressure, reducing the coolant density effect on moderation, and the actual pressure tube can be kept cool by the calandria water.

• Extensive material development and research on supercritical water chemistry under radiation is needed.

However, the entire SCWR coolant is cleaned after condensation. This is an advantage in managing water chemistry and Stress corrosion cracking of structural materials. It is not possible in LWRs where hot coolant circulates.

• Special start-up procedures needed to avoid instability before the water reaches supercritical conditions.

However, Instability is managed by power to coolant flow rate ratio as a BWR. The coolant density change is smaller in SCWRs than BWRs.

• A fast SCWR needs a relatively complex reactor core to have a negative void coefficient.

However, single coolant flow pass core is feasible.

• As with all alternatives to currently widespread designs (mostly subcritical water cooled, water moderated thermal reactors of some kind) there will be fewer suppliers of technology and parts and less expertise at least initially than for decades old proven technology or its evolutionary improvements such as generation III+ reactors.

However, LWRs were developed in the 1950s based on the subcritical fossil fired power technologies. The success of LWRs is based on that experience. Supercritical fossil fired power plants were developed after 1950s. Components such as valves, piping, turbines, feedwater pumps and heaters for operation at turbine throttle pressure up to 30 MPa (4,400 psi) and temperature up to 630 °C (903 K; 1,166 °F) are present for commercial applications. SCWRs are natural evolution of LWRs. The competitiveness of LWRs in the electricity market is being challenged in the US due to Shale gas from historical summaries of U.S. Energy Information Administration’s (EIA’s) Levelized cost of electricity (LCOE) projections (2010-2020) in Cost of electricity by source. LWRs are the dominant design with the largest share of nuclear power generation and are the current offering for new construction in the world. Innovation dynamics show that innovation does not come from companies with the largest market share. Comparing SCWRs and LWRs is not relevant in terms of innovation dynamics. If Small modular reactor (SMR) is competitive, a SMR version of SCWRs will increase its advantage.

• The chemical shim might behave drastically different as the solution properties of supercritical water are vastly different from those of liquid water. Currently most pressurized water reactors employ boric acid to control reactivity early in burnup.

However, chemical shim cannot be used in SCWRs as well as BWRs, due to the positive coolant void coefficient. SCWRs use borated water as the secondary shut-down similar to BWRs.

• Depending on design online refuelling may be impossible. While CANDUs are capable of online refuelling, other water moderated reactors are not.

However, the Capacity factor of LWRs is already high in USA, over 90%. Pressure vessel type SCWRs do not require online refuelling.