See also
- Atmea
- Mitsubishi FBR Systems
- AP1000, the Westinghouse nuclear power plant
The Mitsubishi advanced pressurized water reactor (APWR) is a generation III nuclear reactor design developed by Mitsubishi Heavy Industries (MHI) based on pressurized water reactor technology. It features several design enhancements including a neutron reflector, improved efficiency and improved safety systems. It has safety features advanced over the last generation, including a combination of passive and active systems. None are currently under construction.
The standard APWR is going through the licensing process in Japan and two (of 1538 MWe) are being constructed at the Tsuruga plant. The next APWR+ will be of a 1700 MWe power and have full MOX core abilities.
The US-APWR was developed by MHI to modify their APWR design to comply with US regulations. TXU selected the US-APWR for use at multiple sites, including the Comanche Peak Nuclear Generating Station. [1] However, in 2013, Mitsubishi slowed U.S. certification work, and the application to build two units at Comanche was suspended. [2]
The reactors are intended for use in nuclear power plants to produce nuclear power from nuclear fuel.
| Electric Power | 1,700 MWe [3] |
| Core Thermal Power | 4,451 MWt |
| Reactor Fuel Assemblies | 257 |
| Reactor Fuel | Advanced 17x17, 14 ft. |
| Active Core Length | 4.2 meters |
| Coolant System Loops | 4 |
| Coolant Flow | 7.64 m3/s/loop |
| Coolant Pressure | 15.5 MPa |
| Steam Generator Type | 90TT-1 |
| Number of Steam Generators | 4 |
| Reactor Coolant Pump Type | 100A |
| Number of Reactor Coolant Pumps | 4 |
| Reactor Coolant Pump Motor Output | 6 MWe |
The US-APWR has several design features to improve plant economics. The core is surrounded by a steel neutron reflector which increases reactivity and saves ~0.1wt% U-235 enrichment. In addition, the US-APWR uses more advanced steam generators (compared to the APWR) which creates drier steam allowing for the use of higher efficiency (and more delicate) turbines. This leads to a ~10% efficiency increase compared to the APWR.
Several safety improvements are also notable. The safety systems have enhanced redundancy, utilizing 4 trains each capable of supplying 50% of the needed high pressure makeup water instead of 2 trains capable of 100%. Also, more reliance is placed on the accumulators which have been redesigned and increased in size. The improvements in this passive system have led to the elimination of the Low Pressure Safety Injection system, an active system.
In 2013, plans to build units in the U.S. were suspended: [2]
On 10 May 2011, Japanese Prime Minister Naoto Kan announced that Japan was cancelling plans for new nuclear construction, including the 2 proposed new APWR reactors at Tsuruga Nuclear Power Plant. [4] As of 2014, under a new government, plans for Tsuruga were uncertain. [5] In March 2015 the Nuclear Regulation Authority (NRA) accepted an expert report that concluded Tsuruga is on an active geological fault. [6]
The CANDU is a Canadian pressurized heavy-water reactor design used to generate electric power. The acronym refers to its deuterium oxide moderator and its use of uranium fuel. CANDU reactors were first developed in the late 1950s and 1960s by a partnership between Atomic Energy of Canada Limited (AECL), the Hydro-Electric Power Commission of Ontario, Canadian General Electric, and other companies.
A nuclear reactor, formerly known as an atomic pile, 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 early 2019, the IAEA reports there are 454 nuclear power reactors and 226 nuclear research reactors in operation around the world.
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. The main difference between a BWR and PWR is that in a BWR, the reactor core heats water, which turns to steam and then drives a steam turbine. In a PWR, the reactor core heats water, which does not boil. This hot water then exchanges heat with a lower pressure water system, which turns to steam and drives the turbine. The BWR was developed by the Argonne National Laboratory and General Electric (GE) in the mid-1950s. The main present manufacturer is GE Hitachi Nuclear Energy, which specializes in the design and construction of this type of reactor.
A breeder reactor is a nuclear reactor that generates more fissile material than it consumes. Breeder reactors achieve this because their neutron economy is high enough to create more fissile fuel than they use, by irradiation of a fertile material, such as uranium-238 or thorium-232, that is loaded into the reactor along with fissile fuel. Breeders were at first found attractive because they made more complete use of uranium fuel than light water reactors, but interest declined after the 1960s as more uranium reserves were found, and new methods of uranium enrichment reduced fuel costs.
Magnox is a type of nuclear power/production reactor that was designed to run on natural uranium with graphite as the moderator and carbon dioxide gas as the heat exchange coolant. It belongs to the wider class of gas-cooled reactors. The name comes from the magnesium-aluminium alloy used to clad the fuel rods inside the reactor. Like most other "Generation I nuclear reactors", the Magnox was designed with the dual purpose of producing electrical power and plutonium-239 for the nascent nuclear weapons program in Britain. The name refers specifically to the United Kingdom design but is sometimes used generically to refer to any similar reactor.
Monju (もんじゅ) was a Japanese sodium-cooled fast reactor, located near the Tsuruga Nuclear Power Plant, Fukui Prefecture. Its name is a reference to Manjusri.
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.
The San Onofre Nuclear Generating Station (SONGS) is a permanently closed nuclear power plant located south of San Clemente, California, on the Pacific coast, in Nuclear Regulatory Commission Region IV. The plant was shut down in 2013 after replacement steam generators failed; it is currently in the process of decommissioning.
The Advanced CANDU reactor (ACR), or ACR-1000, is a Generation III+ nuclear reactor designed by Atomic Energy of Canada Limited (AECL). It combines features of the existing CANDU pressurised heavy water reactors (PHWR) with features of light-water cooled pressurized water reactors (PWR). From CANDU, it takes the heavy water moderator, which gives the design an improved neutron economy that allows it to burn a variety of fuels. It replaces the heavy water cooling loop with one containing conventional light water, reducing costs. The name refers to its design power in the 1,000 MWe class, with the baseline around 1,200 MWe.
Generation IV reactors are a set of nuclear reactor designs currently being researched for commercial applications by the Generation IV International Forum. They are motivated by a variety of goals including improved safety, sustainability, efficiency, and cost.
The lead-cooled fast reactor is a nuclear reactor design that features a fast neutron spectrum and molten lead or lead-bismuth eutectic coolant. Molten lead or lead-bismuth eutectic can be used as the primary coolant because lead and bismuth have low neutron absorption and relatively low melting points. Neutrons are slowed less by interaction with these heavy nuclei and therefore, help make this type of reactor a fast-neutron reactor. The coolant does, however, serve as a neutron reflector, returning some escaping neutrons to the core. Fuel designs being explored for this reactor scheme include fertile uranium as a metal, metal oxide or metal nitride. Smaller capacity lead-cooled fast reactors can be cooled by natural convection, while larger designs use forced circulation in normal power operation, but with natural circulation emergency cooling. The reactor outlet coolant temperature is typically in the range of 500 to 600 °C, possibly ranging over 800 °C with advanced materials for later designs. Temperatures higher than 800 °C are high enough to support thermochemical production of hydrogen through the sulfur-iodine cycle.
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.
Carolinas–Virginia Tube Reactor (CVTR), also known as Parr Nuclear Station, was an experimental pressurized tube heavy water nuclear power reactor at Parr, South Carolina in Fairfield County. It was built and operated by the Carolinas Virginia Nuclear Power Associates. CVTR was a small test reactor, capable of generating 17 megawatts of electricity. It was officially commissioned in December 1963 and left service in January 1967.
Comanche Peak Nuclear Power Plant is located in Somervell County, Texas. The nuclear power plant is located 40 miles (64 km) southwest of Ft. Worth and about 60 miles (97 km) southwest of Dallas. It relies on nearby Squaw Creek Reservoir for cooling water. The plant has about 1,300 employees and is operated by Luminant Generation, a subsidiary of Vistra Energy.
The Tsuruga Nuclear Power Plant is located in the city of Tsuruga, Fukui Prefecture, Japan. It is operated by the Japan Atomic Power Company (JAPC). The total site area is 5.12 square kilometres (1.98 sq mi) with 94% of it being green area that the company is working to preserve. The Tsuruga site is a dual site with the decommissioned prototype Fugen Nuclear Power Plant.
Atmea is a joint venture between Mitsubishi Heavy Industries (MHI) and EDF Group that develops, markets, licenses and sells the ATMEA1 reactor, a new generation III+, medium-power pressurized water reactor (PWR). The company is headquartered in Paris.
The Energy Multiplier Module is a nuclear fission power reactor under development by General Atomics. It is a fast-neutron version of the Gas Turbine Modular Helium Reactor (GT-MHR) and is capable of converting spent nuclear fuel into electricity and industrial process heat.
The IPHWR-700 is an Indian pressurized heavy-water reactor designed by the Bhabha Atomic Research Centre. It is a Generation III+ reactor developed from earlier CANDU based 220 MW and 540 MW designs. It can generate 700 MW of electricity. Currently there are 6 units under construction and 10 more units planned, at a cost of INR 1.05 trillion.
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