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The Economic Simplified Boiling Water Reactor (ESBWR) is a passively safe generation III+ reactor design derived from its predecessor, the Simplified Boiling Water Reactor (SBWR) and from the Advanced Boiling Water Reactor (ABWR). All are designs by GE Hitachi Nuclear Energy (GEH), and are based on previous Boiling Water Reactor designs.
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The passive nuclear safety systems in an ESBWR operate without using any pumps, which creates increased design safety, integrity, and reliability, while simultaneously reducing overall reactor cost. It also uses natural circulation to drive coolant flow within the reactor pressure vessel (RPV); this results in fewer systems to maintain, and precludes significant BWR casualties such as recirculation line breaks. There are no circulation pumps or associated piping, power supplies, heat exchangers, instrumentation, or controls needed for these systems.
ESBWR's passive safety systems include a combination of three systems that allow for the efficient transfer of decay heat (created from nuclear decay) from the reactor to pools of water outside containment –the Isolation Condenser System, the Gravity Driven Cooling System, and the Passive Containment Cooling System. These systems utilize natural circulation based on simple laws of physics to transfer the decay heat outside containment while maintaining water levels inside the reactor, keeping the nuclear fuel submerged in water and adequately cooled.
In events where the reactor coolant pressure boundary remains intact, the Isolation Condenser System (ICS) is used to remove decay heat from the reactor and transfer it outside containment. The ICS system is a closed loop system that connects the reactor pressure vessel to a heat exchanger located in the upper elevation of the reactor building. Steam leaves the reactor through the ICS piping and travels to the ICS heat exchangers which are submerged in a large pool. The steam is condensed in the heat exchangers and the denser condensate then flows back down to the reactor to complete the cooling loop. Reactor coolant is cycled through this flow path to provide continuous cooling and to add water to the reactor core.
In cases where the reactor coolant pressure boundary does not remain intact and water inventory in the core is being lost, the Passive Containment Cooling System (PCCS) and Gravity Driven Cooling System (GDCS) work in concert to maintain the water level in the core and remove decay heat from the reactor by transferring it outside containment.
If the water level inside the reactor pressure vessel drops to a predetermined level, due to the loss of water inventory, the reactor is depressurized and the GDCS is initiated. It consists of large pools of water inside containment located above the reactor that are connected to the reactor pressure vessel. When the GDCS system is initiated, gravity forces water to flow from the pools into the reactor. The pools are sized to provide sufficient amounts of water to maintain the water at a level above the top of the nuclear fuel. After the reactor has been depressurized, the decay heat is transferred to the containment as water inside the reactor boils and exits the reactor pressure vessel into the containment in the form of steam.
The PCCS consists of a set of heat exchangers located in the upper portion of the reactor building. The steam from the reactor rises through the containment to the PCCS heat exchangers where the steam is condensed. The condensate then drains from the PCCS heat exchangers back to the GDCS pools where it completes the cycle and drains back to the reactor pressure vessel.
Both the ICS and PCCS heat exchangers are submerged in a pool of water large enough to provide 72 hours of reactor decay heat removal capability. The pool is vented to the atmosphere and is located outside of the containment. The combination of these features allows the pool to be refilled easily with low pressure water sources and installed piping.
The reactor core is shorter than in conventional BWR plants to reduce the pressure drop over the fuel, thereby enabling natural circulation. There are 1,132 fuel rod bundles and the thermal power is 4,500 MWth in the standardized SBWR. [1] The nominal output is rated at 1594 MWe gross and 1535 MWe net, yielding an overall plant Carnot efficiency of approximately 35%. [2]
In case of an accident, the ESBWR can remain in a safe, stable state for 72 hours without any operator action or even electrical power. ESBWR safety systems are designed to operate normally in the event of station blackout, which prevented proper functioning of the emergency core cooling systems at the Fukushima Daiichi Nuclear Power Plant. Below the vessel, there is a piping structure (core catcher) that allows for cooling of the core during any very severe accident. These pipes facilitate cooling above and below the molten core with water. The final safety evaluation report accepted by the NRC reports an overall core damage frequency of 1.65 * 10−8 per year (i.e., roughly once every 60 million years). [3]
Similarly to the ABWR, The containment is inerted with nitrogen before operation to prevent fires, and can be deinerted after reactor shutdown for maintenance. As this BWR can not be controlled using flow rate control as it lacks recirculation pumps, it can instead be controlled with the temperature of the feedwater entering the reactor. [4]
The ESBWR received a positive Safety Evaluation Report [5] and Final Design Approval [6] on March 9, 2011. On June 7, 2011, the NRC completed its public comment period. [7] Final rule was issued on September 16, 2014, after two outstanding problems with GE-Hitachi's modeling of loads on the steam dryer were solved. [8] [9]
In January 2014, GE Hitachi paid $2.7 million to resolve a lawsuit alleging it made false claims to the NRC about its analysis of the steam dryer. [10]
The NRC granted design approval in September 2014. [11]
However, in September 2015, at the request of owner Entergy, the NRC withdrew the Combined Construction and Operating License application for the first proposed ESBWR unit at Grand Gulf Nuclear Generating Station. [12]
On May 31, 2017, the Nuclear Regulatory Commission announced that it had authorized the issuance of a Combined License for North Anna Nuclear Generating Station unit 3. [13] [14]
A nuclear reactor is a device used to initiate and control a fission nuclear chain reaction or nuclear fusion reactions. Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nuclear fission is passed to a working fluid, which in turn runs through steam turbines. These either drive a ship's propellers or turn electrical generators' shafts. Nuclear generated steam in principle can be used for industrial process heat or for district heating. Some reactors are used to produce isotopes for medical and industrial use, or for production of weapons-grade plutonium. As of 2022, the International Atomic Energy Agency reports there are 422 nuclear power reactors and 223 nuclear research reactors in operation around the world.
The Three Mile Island accident was a nuclear meltdown of the Unit 2 reactor (TMI-2) of the Three Mile Island Nuclear Generating Station on the Susquehanna River in Londonderry Township, near the capital city of Harrisburg, Pennsylvania. The reactor accident began at 4:00 a.m. on March 28, 1979, and released radioactive gases and radioactive iodine into the environment. It is the worst accident in U.S. commercial nuclear power plant history. On the seven-point logarithmic International Nuclear Event Scale, the TMI-2 reactor accident is rated Level 5, an "Accident with Wider Consequences".
A pressurized water reactor (PWR) is a type of light-water nuclear reactor. PWRs constitute the large majority of the world's nuclear power plants. In a PWR, the primary coolant (water) is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms. The heated, high pressure water then flows to a steam generator, where it transfers its thermal energy to lower pressure water of a secondary system where steam is generated. The steam then drives turbines, which spin an electric generator. In contrast to a boiling water reactor (BWR), pressure in the primary coolant loop prevents the water from boiling within the reactor. All light-water reactors use ordinary water as both coolant and neutron moderator. Most use anywhere from two to four vertically mounted steam generators; VVER reactors use horizontal steam generators.
A boiling water reactor (BWR) is a type of light water nuclear reactor used for the generation of electrical power. It is the second most common type of electricity-generating nuclear reactor after the pressurized water reactor (PWR), which is also a type of light water nuclear reactor.
A nuclear meltdown is a severe nuclear reactor accident that results in core damage from overheating. The term nuclear meltdown is not officially defined by the International Atomic Energy Agency or by the United States Nuclear Regulatory Commission. It has been defined to mean the accidental melting of the core of a nuclear reactor, however, and is in common usage a reference to the core's either complete or partial collapse.
A loss-of-coolant accident (LOCA) is a mode of failure for a nuclear reactor; if not managed effectively, the results of a LOCA could result in reactor core damage. Each nuclear plant's emergency core cooling system (ECCS) exists specifically to deal with a LOCA.
The light-water reactor (LWR) is a type of thermal-neutron reactor that uses normal water, as opposed to heavy water, as both its coolant and neutron moderator; furthermore a solid form of fissile elements is used as fuel. Thermal-neutron reactors are the most common type of nuclear reactor, and light-water reactors are the most common type of thermal-neutron reactor.
Passive nuclear safety is a design approach for safety features, implemented in a nuclear reactor, that does not require any active intervention on the part of the operator or electrical/electronic feedback in order to bring the reactor to a safe shutdown state, in the event of a particular type of emergency. Such design features tend to rely on the engineering of components such that their predicted behaviour would slow down, rather than accelerate the deterioration of the reactor state; they typically take advantage of natural forces or phenomena such as gravity, buoyancy, pressure differences, conduction or natural heat convection to accomplish safety functions without requiring an active power source. Many older common reactor designs use passive safety systems to a limited extent, rather, relying on active safety systems such as diesel-powered motors. Some newer reactor designs feature more passive systems; the motivation being that they are highly reliable and reduce the cost associated with the installation and maintenance of systems that would otherwise require multiple trains of equipment and redundant safety class power supplies in order to achieve the same level of reliability. However, weak driving forces that power many passive safety features can pose significant challenges to effectiveness of a passive system, particularly in the short term following an accident.
A containment building is a reinforced steel, concrete or lead structure enclosing a nuclear reactor. It is designed, in any emergency, to contain the escape of radioactive steam or gas to a maximum pressure in the range of 275 to 550 kPa. The containment is the fourth and final barrier to radioactive release, the first being the fuel ceramic itself, the second being the metal fuel cladding tubes, the third being the reactor vessel and coolant system.
The advanced boiling water reactor (ABWR) is a Generation III boiling water reactor. The ABWR is currently offered by GE Hitachi Nuclear Energy (GEH) and Toshiba. The ABWR generates electrical power by using steam to power a turbine connected to a generator; the steam is boiled from water using heat generated by fission reactions within nuclear fuel. Kashiwazaki-Kariwa unit 6 is considered the first Generation III reactor in the world.
The supercritical water reactor (SCWR) is a concept Generation IV reactor, designed as a light water reactor (LWR) that operates at supercritical pressure. 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.
A sodium-cooled fast reactor is a fast neutron reactor cooled by liquid sodium.
A steam generator is a heat exchanger used to convert water into steam from heat produced in a nuclear reactor core. They are used in pressurized water reactor between the primary and secondary coolant loops.
The water-water energetic reactor (WWER), or VVER is a series of pressurized water reactor designs originally developed in the Soviet Union, and now Russia, by OKB Gidropress. The idea of such a reactor was proposed at the Kurchatov Institute by Savely Moiseevich Feinberg. VVER were originally developed before the 1970s, and have been continually updated. As a result, the name VVER is associated with a wide variety of reactor designs spanning from generation I reactors to modern generation III+ reactor designs. Power output ranges from 70 to 1300 MWe, with designs of up to 1700 MWe in development. The first prototype VVER-210 was built at the Novovoronezh Nuclear Power Plant.
The three primary objectives of nuclear reactor safety systems as defined by the U.S. Nuclear Regulatory Commission are to shut down the reactor, maintain it in a shutdown condition and prevent the release of radioactive material.
GE Hitachi Nuclear Energy (GEH) is a provider of advanced reactors and nuclear services. It is headquartered in Wilmington, North Carolina, United States. Established in June 2007, GEH is a nuclear alliance created by General Electric and Hitachi. In Japan, the alliance is Hitachi-GE Nuclear Energy. In November 2015, Jay Wileman was appointed CEO.
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.
General Electric's BWR product line of boiling water reactors represents the designs of a relatively large (~18%) percentage of the commercial fission reactors around the world.
The Integral Molten Salt Reactor (IMSR) is a nuclear power plant design targeted at developing a commercial product for the small modular reactor (SMR) market. It employs molten salt reactor technology which is being developed by the Canadian company Terrestrial Energy. It is based closely on the denatured molten salt reactor (DMSR), a reactor design from Oak Ridge National Laboratory. In addition, it incorporates some elements found in the SmAHTR, a later design from the same laboratory. The IMSR belongs to the DMSR class of molten salt reactors (MSR) and hence is a "burner" reactor that employs a liquid fuel rather than a conventional solid fuel. This liquid contains the nuclear fuel as well as serving as the primary coolant.
The BWRX-300 is a design for a small modular nuclear reactor proposed by GE Hitachi Nuclear Energy (GEH). The BWRX-300 would feature passive safety, in that neither external power nor operator action would be required to maintain a safe state, even in extreme circumstances.
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| Heavy water by coolant |
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None (fast-neutron) |
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