Related Research Articles
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 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.
A breeder reactor is a nuclear reactor that generates more fissile material than it consumes. These reactors can be fueled with more-commonly available isotopes of uranium and thorium, such as uranium-238 and 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.
Superphénix or SPX was a nuclear power station prototype on the Rhône river at Creys-Malville in France, close to the border with Switzerland. Superphénix was a 1,242 MWe fast breeder reactor with the twin goals of reprocessing nuclear fuel from France's line of conventional nuclear reactors, while also being an economical generator of power on its own. As of 2022, Superphénix remains the largest breeder reactor ever built.
A fast-neutron reactor (FNR) or fast-spectrum reactor or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons, as opposed to slow thermal neutrons used in thermal-neutron reactors. Such a fast reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor. Around 20 land based fast reactors have been built, accumulating over 400 reactor years of operation globally. The largest of this was the Superphénix Sodium cooled fast reactor in France that was designed to deliver 1,242 MWe. Fast reactors have been intensely studied since the 1950s, as they provide certain advantages over the existing fleet of water cooled and water moderated reactors. These are:
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 integral fast reactor is a design for a nuclear reactor using fast neutrons and no neutron moderator. IFR would breed more fuel and is distinguished by a nuclear fuel cycle that uses reprocessing via electrorefining at the reactor site.
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.
Generation IVreactors 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 World Nuclear Association in 2015 suggested that some might enter commercial operation before 2030.
The gas-cooled fast reactor (GFR) system is a nuclear reactor design which is currently in development. Classed as a Generation IV reactor, it features a fast-neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The reference reactor design is a helium-cooled system operating with an outlet temperature of 850 °C (1,560 °F) using a direct Brayton closed-cycle gas turbine for high thermal efficiency. Several fuel forms are being considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations are being considered based on pin- or plate-based fuel assemblies or prismatic blocks, which allows for better coolant circulation than traditional fuel assemblies.
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.
A sodium-cooled fast reactor is a fast neutron reactor cooled by liquid sodium.
The Fast Breeder Test Reactor (FBTR) is a breeder reactor located at Kalpakkam, Tamil Nadu, India. The Indira Gandhi Center for Atomic Research (IGCAR) and Bhabha Atomic Research Centre (BARC) jointly designed, constructed, and operate the reactor.
The BN-600 reactor is a sodium-cooled fast breeder reactor, built at the Beloyarsk Nuclear Power Station, in Zarechny, Sverdlovsk Oblast, Russia. It has a 600 MWe gross capacity and a 560 MWe net capacity, dispatched o the Middle Urals power grid. It has been in operation since 1980 and represents an evolution on the preceding BN-350 reactor. In 2014, its larger sister reactor, the BN-800 reactor began operation.
The Beloyarsk Nuclear Power Station was the third of the Soviet Union's nuclear plants. It is situated by Zarechny in Sverdlovsk Oblast, Russia. Zarechny township was created to service the station, which is named after the Beloyarsky District. The closest city is Yekaterinburg.
The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe fast breeder nuclear reactor presently being constructed at Kalpakkam, India. The Indira Gandhi Centre for Atomic Research (IGCAR) is responsible for the design of this reactor. The facility builds on the decades of experience gained from operating the lower power Fast Breeder Test Reactor (FBTR). At first, the reactor's construction was supposed to be completed in September 2010, but there were several delays. The Prototype Fast Breeder Reactor is scheduled to be put into service in December 2024, which is more than 20 years after construction began and 14 years after the original commissioning date, as of December 2023. The project's cost has doubled from ₹3,500 crore to ₹7,700 crore due to the multiple delays.The project currently holds an infamous record of being the longest under-construction nuclear reactor project in the world.
The liquid fluoride thorium reactor is a type of molten salt reactor. LFTRs use the thorium fuel cycle with a fluoride-based molten (liquid) salt for fuel. In a typical design, the liquid is pumped between a critical core and an external heat exchanger where the heat is transferred to a nonradioactive secondary salt. The secondary salt then transfers its heat to a steam turbine or closed-cycle gas turbine.
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. The reactor is 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 BN-1200 reactor is a sodium-cooled fast breeder reactor project, under development by OKBM Afrikantov in Zarechny, Russia. The BN-1200 is based on the earlier BN-600 and especially BN-800, with which it shares a number of features. The reactor's name comes from its electrical output, nominally 1220 MWe.
The IPHWR-700 is an Indian pressurized heavy-water reactor designed by the NPCIL. 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 5 units under construction and 10 more units planned, at a cost of ₹1.05 lakh crore (US$13 billion).
References
- ↑ Design completed for prototype fast reactor, World Nuclear News. 2 September 2014
- ↑ Lead-Cooled Fast-Neutron Reactor (BREST), IAEA. 2015
- ↑ Russia Accelerates Efforts to Build Advanced Nuclear Reactors, Powermag. 1 October 2016
- ↑ Russia notes progress with fast reactor technology, World Nuclear News. 23 August 2016
- ↑ Preparatory construction for Brest-300 reactor begins in Russia, Nuclear Engineering International. 22 May 2020
- ↑ ""Росатом" начал строительство реактора Брест-300 в Северске". 8 June 2021.
- ↑ Proctor, Darrell (8 June 2021). "Nuclear First—Work Starts on Russian Fast Neutron Reactor". POWER Magazine. Retrieved 21 June 2021.
