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115-472: 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

230-399: A CANDU reactor . In many cases, a scram is part of the routine shutdown procedure which serves to test the emergency shutdown system. There is no definitive origin for the term. United States Nuclear Regulatory Commission historian Tom Wellock notes that scram is English-language slang for leaving quickly and urgently, and he cites this as the original and most likely accurate basis for

345-623: A 1952 U.S. Atomic Energy Commission (AEC) report by Fermi, the AEC declassified information on the Chicago Pile. The report includes a section written by Wilson's team shortly after the Chicago Pile achieved a self-sustaining chain reaction on December 2, 1942. It includes a wiring schematic of the rod control circuitry with a clearly labeled "SCRAM" line (see image on the right and pages 37 and 48). The Russian name, AZ-5 ( АЗ-5 , in Cyrillic ),

460-471: 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

575-594: 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 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

690-404: A big button to push to drive in both the control rods and the safety rod. What to label it? 'What do we do after we punch the button?,' someone asked. 'Scram out of here!,' Wilson said. Bill Overbeck, another member of that group said, 'OK I'll label it SCRAM.'" The earliest references to "scram" among the Chicago Pile team were also associated with Wilson's shutdown circuitry and not Hilberry. In

805-433: A brief period. Often they are used to provide electrical power until the plant electrical supply can be switched to the batteries and/or diesel generators. Batteries often form the final redundant backup electrical system and are also capable of providing sufficient electrical power to shut down a plant. Containment systems are designed to prevent the release of radioactive material into the environment. The fuel cladding

920-458: A decrease in neutron multiplication , and thus shutting down the reactor without use of the control rods. In the PWR, these neutron absorbing solutions are stored in pressurized tanks (called accumulators) that are attached to the primary coolant system via valves. A varying level of neutron absorbent is kept within the primary coolant at all times, and is increased using the accumulators in the event of

1035-411: 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),

1150-430: A failure of all of the control rods to insert, which will promptly bring the reactor below the shutdown margin. In the BWR, soluble neutron absorbers are found within the standby liquid control system , which uses redundant battery-operated injection pumps, or, in the latest models, high pressure nitrogen gas to inject the neutron absorber solution into the reactor vessel against any pressure within. Because they may delay

1265-438: A full melt-down, the fuel would most likely end up on the concrete floor of the primary containment building. Concrete can withstand a great deal of heat, so the thick flat concrete floor in the primary containment will often be sufficient protection against the so-called China Syndrome . The Chernobyl plant didn't have a containment building, but the core was eventually stopped by the concrete foundation. Due to concerns that

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1380-512: 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

1495-521: A negative pressure within the secondary containment to limit the release of radioactive material. Each SGTS train generally consists of a mist eliminator/roughing filter; an electric heater; a prefilter; two absolute ( HEPA ) filters; an activated charcoal filter; an exhaust fan; and associated valves, ductwork, dampers, instrumentation and controls. The signals that trip the SGTS system are plant-specific; however, automatic trips are generally associated with

1610-712: 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

1725-500: 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

1840-516: 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

1955-445: A reactor that has not had a constant power history, the exact percentage is determined by the concentrations and half-lives of the individual fission products in the core at the time of the scram. The power produced by decay heat decreases as the fission products decay, but it is large enough that failure to remove decay heat may cause the reactor core temperature to rise to dangerous levels and has caused nuclear accidents , including

2070-405: A rope with a man with an axe standing next to it; cutting the rope would mean the rods would fall by gravity into the reactor core, shutting the reactor down. The axe man at the first chain reaction was Norman Hilberry . In a letter to Raymond Murray (January 21, 1981), Hilberry wrote: When I showed up on the balcony on that December 2, 1942 afternoon, I was ushered to the balcony rail, handed

2185-468: A scram to insert the control rods, as it is the most reliable method of completely inserting the control rods, and prevents the possibility of accidentally withdrawing them during or after the shutdown. Most neutrons in a reactor are prompt neutrons ; that is, neutrons produced directly by a fission reaction. These neutrons move at a high velocity , so they are likely to escape into the moderator before being captured . On average, it takes about 13 μs for

2300-413: A series of pumps and spargers that spray coolant into the upper portion of the primary containment structure. It is designed to condense the steam into liquid within the primary containment structure in order to prevent overpressure and overtemperature, which could lead to leakage, followed by involuntary depressurization. This system is often driven by a steam turbine to provide enough water to safely cool

2415-413: A solution containing boric acid , which acts as a neutron poison and rapidly floods the core in case of problems with the stopping of the chain reaction. Pressurized water reactors also can SCRAM the reactor completely with the help of their control rods. PWRs also use boric acid to make fine adjustments to reactor power level, or reactivity, using their Chemical and Volume Control System (CVCS). In

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2530-423: 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

2645-507: A well sharpened fireman's axe and told, "If the safety rods fail to operate, cut that manila rope ." The safety rods, needless to say, worked, the rope was not cut... I don't believe I have ever felt quite as foolish as I did then. ...I did not get the SCRAM [Safety Control Rod Axe Man] story until many years after the fact. Then one day one of my fellows who had been on Zinn's construction crew called me Mr. Scram. I asked him, "How come?" And then

2760-468: Is a mode of operation of a residual heat removal system, also known as an RHR or RHS but is generally called LPCI. It is also not a stand-alone valve or system. This system uses spargers (pipes fitted with an array of many small spray nozzles) within the reactor pressure vessel to spray water directly onto the fuel rods, suppressing the generation of steam. Reactor designs can include core spray in high-pressure and low-pressure modes. This system consists of

2875-495: 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. The typical steam turbine used in

2990-506: Is an abbreviation for аварийная защита 5-й категории ( avariynaya zashhchita 5-y kategorii ), which translates to "emergency protection of the 5th category" in English. In any reactor, a scram is achieved by inserting large amounts of negative reactivity mass into the midst of the fissile material, to immediately terminate the fission reaction. In light-water reactors , this is achieved by inserting neutron-absorbing control rods into

3105-508: 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

3220-424: 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

3335-399: Is designed to immediately terminate the nuclear reaction. By breaking the nuclear chain reaction , the source of heat is eliminated. Other systems can then be used to remove decay heat from the core. All nuclear plants have some form of reactor protection system. Control rods are a series of rods that can be quickly inserted into the reactor core to absorb neutrons and rapidly terminate

3450-590: 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. 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

3565-431: 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,

Boiling water reactor safety systems - Misplaced Pages Continue

3680-405: 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

3795-445: Is impossible, by directly flooding the core with coolant. These systems are of three major types: 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

3910-576: 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

4025-417: 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

4140-548: 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

4255-419: 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

4370-467: 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

4485-417: Is pressurized. It is designed to monitor the level of coolant in the reactor vessel and automatically inject coolant when the level drops below a threshold. This system is normally the first line of defense for a reactor since it can be used while the reactor vessel is still highly pressurized. The Automatic Depressurization System (ADS) consists of a series of valves which open to vent steam several feet under

4600-518: Is recirculated via a cooling tower . The failure of half of the ESWS pumps was one of the factors that endangered safety in the 1999 Blayais Nuclear Power Plant flood , while a total loss occurred during the Fukushima I and Fukushima II nuclear accidents in 2011. Emergency core cooling systems (ECCS) are designed to safely shut down a nuclear reactor during accident conditions. The ECCS allows

4715-425: 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", "torus" or "suppression pool". All BWRs utilize a number of safety/relief valves for overpressure; up to 7 of these are

Boiling water reactor safety systems - Misplaced Pages Continue

4830-503: 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

4945-665: Is the first layer of protection around the nuclear fuel and is designed to protect the fuel from corrosion that would spread fuel material throughout the reactor coolant circuit. In most reactors it takes the form of a sealed metallic or ceramic layer. It also serves to trap fission products, especially those that are gaseous at the reactor's operating temperature , such as krypton , xenon and iodine . Cladding does not constitute shielding, and must be developed such that it absorbs as little radiation as possible. For this reason, materials such as magnesium and zirconium are used for their low neutron capture cross sections. The reactor vessel

5060-408: Is the first layer of shielding around the nuclear fuel and usually is designed to trap most of the radiation released during a nuclear reaction. The reactor vessel is also designed to withstand high pressures. The primary containment system usually consists of a large metal and/or concrete structure (often cylindrical or bulb shaped) that contains the reactor vessel. In most reactors it also contains

5175-407: 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

5290-436: 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

5405-561: The AZ-5 shutdown system was initiated after a core overheat. RBMK reactors were subsequently either retrofitted to account for the flaw, or decommissioned. Not all of the heat in a nuclear reactor is generated by the chain reaction that a scram is designed to stop. For a reactor that is scrammed after holding a constant power level for an extended period (greater than 100 hrs), about 7% of the steady-state power will remain after initial shutdown due to fission product decay that cannot be stopped. For

5520-478: 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. The (E)SBWR has an additional ECCS capacity that

5635-594: The Control Rod Drive System (CRDS) to supplement the passive system. 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

5750-558: The ECCS and does not have a low coolant accident function. For pressurized water reactors, this system acts in the secondary cooling circuit and is called Turbine driven auxiliary feedwater system . Under normal conditions, nuclear power plants receive power from generator. However, during an accident a plant may lose access to this power supply and thus may be required to generate its own power to supply its emergency systems. These electrical systems usually consist of diesel generators and batteries . Diesel generators are employed to power

5865-542: 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

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5980-842: The HPCI systems are the "solid wheel" or "water wheel" Terry Steam Turbines manufactured by the Curtiss-Wright Corporation in Summerville, SC. 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

6095-456: 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. The reactor core isolation cooling system

6210-570: 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

6325-548: 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

6440-613: 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,

6555-424: The PWR. There are five major varieties of BWR containments: 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

6670-504: The RCIC systems are the "solid wheel" or "water wheel" Terry Steam Turbines manufactured by the Curtiss-Wright Corporation in Summerville, SC. 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. The Automatic depressurization system

6785-539: 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

6900-510: 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

7015-424: 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. Nuclear safety systems A reactor protection system

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7130-440: The air. SCRAM A scram or SCRAM is an emergency shutdown of a nuclear reactor effected by immediately terminating the fission reaction. It is also the name that is given to the manually operated kill switch that initiates the shutdown. In commercial reactor operations, this type of shutdown is often referred to as a "scram" at boiling water reactors , a "reactor trip " at pressurized water reactors and "EPIS" at

7245-462: 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

7360-413: The bottom of the containment building. The AREVA EPR , SNR-300, SWR1000, ESBWR, and Atmea I reactors have core catchers. The ABWR has a thick layer of basaltic concrete floor specifically designed to catch the core. A standby gas treatment system (SGTS) is part of the secondary containment system. The SGTS system filters and pumps air from secondary containment to the environment and maintains

7475-418: The case of LOCA, PWRs have three sources of backup cooling water, high pressure injection (HPI), low pressure injection (LPI), and core flood tanks (CFTs). They all use water with a high concentration of boron. The essential service water system (ESWS) circulates the water that cools the plant's heat exchangers and other components before dissipating the heat into the environment. Because this includes cooling

7590-421: 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

7705-407: The control rods upon any interruption of the electric current. In both the PWR and the BWR there are secondary systems (and often even tertiary systems) that will insert control rods in the event that primary rapid insertion does not promptly and fully actuate. Liquid neutron absorbers ( neutron poisons ) are also used in rapid shutdown systems for heavy and light water reactors. Following a scram, if

7820-456: The core would melt its way through the concrete, a " core catching device " was invented, and a mine was quickly dug under the plant with the intention to install such a device. The device contains a quantity of metal designed to melt, diluting the corium and increasing its heat conductivity; the diluted metallic mass could then be cooled by water circulating in the floor. Today, all new Russian-designed reactors are equipped with core-catchers in

7935-402: The core, although the mechanism by which rods are inserted depends on the type of reactor. In pressurized water reactors the control rods are held above a reactor's core by electric motors against both their own weight and a powerful spring. A scram is designed to release the control rods from those motors and allows their weight and the spring to drive them into the reactor core, rapidly halting

8050-530: 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

8165-551: The electric heaters and a high temperature condition in the charcoal filters. In case of a radioactive release, most plants have a system designed to remove radioactivity from the air to reduce the effects of the radioactivity release on the employees and public. This system usually consists of containment ventilation that removes radioactivity and steam from primary containment. Control room ventilation ensures that plant operators are protected. This system often consists of activated charcoal filters that remove radioactive isotopes from

8280-460: 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. 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

8395-416: The environment from occurring in nearly any circumstance. 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

8510-401: 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

8625-406: The fuel soon returns to decay heat power levels. Manual SCRAMs may be initiated by the reactor operators, while automatic SCRAMs are initiated upon: 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

8740-438: 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

8855-465: The high pressure systems. Some depressurization systems are automatic in function, while others may require operators to manually activate them. In pressurized water reactors with large dry or ice condenser containments, the valves of the system are called Pilot-operated relief valves . An LPCI is an emergency system which consists of a pump that injects a coolant into the reactor vessel once it has been depressurized. In some nuclear power plants an LPCI

8970-522: The hydrogen generation is not significant. When the nuclear fuel overheats, zirconium in Zircaloy cladding used in fuel rods oxidizes in reaction with steam: 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

9085-435: 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

9200-423: 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

9315-415: The lower explosive limit of 4%, or through recombination with Oxygen to make water. 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

9430-487: The neutrons to be slowed by the moderator enough to facilitate a sustained reaction, which allows the insertion of neutron absorbers to affect the reactor quickly. As a result, once the reactor has been scrammed, the reactor power will drop significantly almost instantaneously. A small fraction (about 0.65%) of neutrons in a typical power reactor comes from the radioactive decay of a fission product. These delayed neutrons , which are emitted at lower velocities, will limit

9545-423: The nuclear reaction by absorbing liberated neutrons. Another design uses electromagnets to hold the rods suspended, with any cut to the electric current resulting in an immediate and automatic control rod insertion. In boiling water reactors , the control rods are inserted up from underneath the reactor vessel. In this case a hydraulic control unit with a pressurized storage tank provides the force to rapidly insert

9660-416: The nuclear reaction. They are typically composed of actinides , lanthanides , transition metals , and boron , in various alloys with structural backing such as steel. In addition to being neutron absorbent, the alloys used also are required to have at least a low coefficient of thermal expansion so that they do not jam under high temperatures, and they have to be self-lubricating metal on metal, because at

9775-406: The plant to respond to a variety of accident conditions (e.g. LOCAs ) and additionally introduce redundancy so that the plant can be shut down even with one or more subsystem failures. In most plants, ECCS is composed of the following systems: The High Pressure Coolant Injection (HPCI) System consists of a pump or pumps that have sufficient pressure to inject coolant into the reactor vessel while it

9890-443: The plant. 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: 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

10005-414: 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

10120-406: 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

10235-485: 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

10350-423: 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. 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

10465-467: The radioactively contaminated systems. The primary containment system is designed to withstand strong internal pressures resulting from a leak or intentional depressurization of the reactor vessel. Some plants have a secondary containment system that encompasses the primary system. This is very common in BWRs because most of the steam systems, including the turbine, contain radioactive materials. In case of

10580-469: The rate at which a nuclear reactor will shut down. Due to flaws in its original control rod design, scramming an RBMK reactor could raise reactivity to dangerous levels before lowering it. This was noticed when it caused a power surge at the startup of Ignalina Nuclear Power Plant Unit number 1, in 1983. On April 26, 1986, the Chernobyl disaster happened due to a fatally flawed shutdown system, after

10695-519: The reactor (or section(s) thereof) are not below the shutdown margin (that is, they could return to a critical state due to insertion of positive reactivity from cooling, poison decay, or other uncontrolled conditions), the operators can inject solutions containing neutron poisons directly into the reactor coolant. Neutron poison solutions are water-based solutions that contain chemicals that absorb neutrons, such as common household borax , sodium polyborate , boric acid , or gadolinium nitrate , causing

10810-411: 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

10925-445: 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. During normal plant operations and in normal operating temperatures,

11040-424: 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

11155-446: 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

11270-399: The reactor if the reactor building is isolated from the control and turbine buildings. Steam turbine driven cooling pumps with pneumatic controls can run at mechanically controlled adjustable speeds, without battery power, emergency generator, or off-site electrical power. The Isolation cooling system is a defensive system against a condition known as station blackout. This system is not part of

11385-537: 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

11500-497: 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

11615-422: 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. The typical steam turbine used in

11730-442: 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

11845-539: 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

11960-404: 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

12075-455: 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

12190-418: The restart of a reactor, these systems are only used to shut down the reactor if control rod insertion fails. This concern is especially significant in a BWR, where injection of liquid boron would cause precipitation of solid boron compounds on fuel cladding, which would prevent the reactor from restarting until the boron deposits were removed. In most reactor designs, the routine shutdown procedure also uses

12305-417: The rods had "scrammed" into the pile. Other witnesses that day agreed with Libby's crediting "scram" to Wilson. Wellock wrote that Warren Nyer, a student who worked on assembling the pile, also attributed the word to Wilson: "The word arose in a discussion Dr. Wilson, who was head of the instrumentation and controls group, was having with several members of his group," Nyer wrote. "The group had decided to have

12420-498: 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

12535-607: The site during emergency situations. They are usually sized such that a single one can provide all the required power for a facility to shut down during an emergency. Facilities have multiple generators for redundancy. Additionally, systems that are required to shut down the reactor have separate electrical sources (often separate generators) so that they do not affect shutdown capability. Loss of electrical power can occur suddenly and can damage or undermine equipment. To prevent damage, motor-generators can be tied to flywheels that can provide uninterrupted electrical power to equipment for

12650-399: The story. Leona Marshall Libby , who was present that day at the Chicago Pile, recalled that the term was coined by Volney Wilson who led the team that designed the control rod circuitry: The safety rods were coated with cadmium foil, and this metal absorbed so many neutrons that the chain reaction was stopped. Volney Wilson called these "scram" rods. He said that the pile had "scrammed,"

12765-532: The surface of a large pool of liquid water (known as the wetwell or torus) in pressure suppression type containments (typically used in boiling water reactor designs), or directly into the primary containment structure in other types of containments, such as large-dry or ice-condenser containments (typically used in pressurized water reactor designs). The actuation of these valves depressurizes the reactor vessel and allows lower pressure coolant injection systems to function, which have very large capacities in comparison to

12880-412: The systems that remove decay heat from both the primary system and the spent fuel rod cooling ponds, the ESWS is a safety-critical system. Since the water is frequently drawn from an adjacent river, the sea, or other large body of water, the system can be fouled by seaweed, marine organisms, oil pollution, ice and debris. In locations without a large body of water in which to dissipate the heat, water

12995-446: The temperatures experienced by nuclear reactor cores oil lubrication would foul too quickly. Boiling water reactors are able to SCRAM the reactor completely with the help of their control rods. In the case of a loss of coolant accident (LOCA), the water-loss of the primary cooling system can be compensated with normal water pumped into the cooling circuit. On the other hand, the standby liquid control (SLC) system (SLCS) consists of

13110-470: The use of scram in the technical context. Scram is sometimes cited as being an acronym for safety control rod axe man or safety cut rope axe man . This was supposedly coined by Enrico Fermi when he oversaw the construction of the world's first nuclear reactor . The core , which was built under the spectator seating at the University of Chicago's Stagg Field , had an actual control rod tied to

13225-479: 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

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