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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.
The RBMK is a class of graphite-moderated nuclear power reactor designed and built by the Soviet Union. The name refers to its design where, instead of a large steel pressure vessel surrounding the entire core, the core is surrounded by a cylindrical annular steel tank inside a concrete vault and each fuel assembly is enclosed in an individual 8 cm (inner) diameter pipe. The channels also contain the coolant, and are surrounded by graphite.
A breeder reactor is a nuclear reactor that generates more fissile material than it consumes. These reactors can be fuelled with more commonly available isotopes of uranium and thorium, such as uranium-238 or thorium-232, as opposed to the rare uranium-235 which is used in conventional reactors. These materials are called fertile materials since they can be bred into fuel by these breeder reactors.
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 molten salt reactor (MSR) is a class of nuclear fission reactor in which the primary nuclear reactor coolant and/or the fuel is a mixture of molten salt with a fissionable material.
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 CANDU reactor (ACR), or ACR-1000, was a proposed Generation III+ nuclear reactor design, developed by Atomic Energy of Canada Limited (AECL). It combined features of the existing CANDU pressurised heavy water reactors (PHWR) with features of light-water cooled pressurized water reactors (PWR). From CANDU, it took the heavy water moderator, which gave the design an improved neutron economy that allowed it to burn a variety of fuels. It replaced 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 nuclear reactor design technologies that are envisioned as successors of generation III reactors. The Generation IV International Forum (GIF) - an international organization that coordinates the development of generation IV reactors - specifically selected six reactor technologies as candidates for generation IV reactors. The designs target improved safety, sustainability, efficiency, and cost. The first commercial Gen IV plants are not expected before 2040–2050, although the World Nuclear Association in 2015 suggested that some might enter commercial operation before 2030.
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 especially lead, and to a lesser degree 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. In simple terms, if a neutron hits a particle with a similar mass, it tends to lose kinetic energy. In contrast, if it hits a much heavier atom such as lead, the neutron will "bounce off" without losing this energy. 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 will employ natural circulation emergency cooling. No operator interference is required, nor pumping of any kind to cool the residual heat of the reactor after shutdown. 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 theoretically high enough to support thermochemical production of hydrogen through the sulfur-iodine cycle, although this has not been demonstrated.
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 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.
The advanced heavy-water reactor (AHWR) or AHWR-300 is the latest Indian design for a next-generation nuclear reactor that burns thorium in its fuel core. It is slated to form the third stage in India's three-stage fuel-cycle plan. This phase of the fuel cycle plan was supposed to be built starting with a 300MWe prototype in 2016.
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.
The B&W mPower was a proposed small modular reactor designed by Babcock & Wilcox, and to be built by Generation mPower LLC, a joint venture of Babcock & Wilcox and Bechtel. It was a Generation III+ integral pressurized water reactor concept.
PRISM is a nuclear power plant design by GE Hitachi Nuclear Energy (GEH).
Small modular reactors (SMRs) are a proposed class of nuclear fission reactors, smaller than conventional nuclear reactors, which can be built in one location, then shipped, commissioned, and operated at a separate site. The term SMR refers to the size, capacity and modular construction only, not to the reactor type and the nuclear process which is applied. Designs range from scaled down versions of existing designs to generation IV designs. Both thermal-neutron reactors and fast-neutron reactors have been proposed, along with molten salt and gas cooled reactor models.
The BN-800 reactor is a sodium-cooled fast breeder reactor, built at the Beloyarsk Nuclear Power Station, in Zarechny, Sverdlovsk Oblast, Russia. The reactor is designed to generate 880 MW of electrical power. The plant was considered part of the weapons-grade Plutonium Management and Disposition Agreement signed between the United States and Russia, with the reactor being part of the final step for a plutonium-burner core The plant reached its full power production in August 2016. According to Russian business journal Kommersant, the BN-800 project cost 140.6 billion rubles.
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. It also incorporates 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 and also serves as primary coolant.
References
- Matzie, R.A.; Longo, J.; Bradbury, R.B.; Teare, K.R.; Hayns, M.R. (August 1992). "Design of the Safe Integral Reactor". Nuclear Engineering and Design. 136 (1–2): 73–83. doi:10.1016/0029-5493(92)90114-B.
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