Supercritical water reactorThe supercritical water reactor (SCWR) is a concept Generation IV reactor, designed as a light water reactor (LWR) that operates at supercritical pressure (i.e. greater than 22.1 MPa). The term critical in this context refers to the critical point of water, and must not be confused with the concept of criticality of the nuclear reactor. The water heated in the reactor core becomes a supercritical fluid above the critical temperature of 374 °C, transitioning from a fluid more resembling liquid water to a fluid more resembling saturated steam (which can be used in a steam turbine), without going through the distinct phase transition of boiling.
Pressurized water reactorA pressurized water reactor (PWR) is a type of light-water nuclear reactor. PWRs constitute the large majority of the world's nuclear power plants (with notable exceptions being the UK, Japan and Canada). 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.
Light-water reactorThe 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. There are three varieties of light-water reactors: the pressurized water reactor (PWR), the boiling water reactor (BWR), and (most designs of) the supercritical water reactor (SCWR).
Fracture toughnessIn materials science, fracture toughness is the critical stress intensity factor of a sharp crack where propagation of the crack suddenly becomes rapid and unlimited. A component's thickness affects the constraint conditions at the tip of a crack with thin components having plane stress conditions and thick components having plane strain conditions. Plane strain conditions give the lowest fracture toughness value which is a material property.
Boiling water reactorA 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.
Nuclear reactorA 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 (water or gas), 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.
Breeder reactorA 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. Breeder reactors achieve this because their neutron economy is high enough to create more fissile fuel than they use.
FractureFracture is the separation of an object or material into two or more pieces under the action of stress. The fracture of a solid usually occurs due to the development of certain displacement discontinuity surfaces within the solid. If a displacement develops perpendicular to the surface, it is called a normal tensile crack or simply a crack; if a displacement develops tangentially, it is called a shear crack, slip band or dislocation. Brittle fractures occur without any apparent deformation before fracture.
Fracture mechanicsFracture mechanics is the field of mechanics concerned with the study of the propagation of cracks in materials. It uses methods of analytical solid mechanics to calculate the driving force on a crack and those of experimental solid mechanics to characterize the material's resistance to fracture. Theoretically, the stress ahead of a sharp crack tip becomes infinite and cannot be used to describe the state around a crack. Fracture mechanics is used to characterise the loads on a crack, typically using a single parameter to describe the complete loading state at the crack tip.
Pressure vesselA pressure vessel is a container designed to hold gases or liquids at a pressure substantially different from the ambient pressure. Construction methods and materials may be chosen to suit the pressure application, and will depend on the size of the vessel, the contents, working pressure, mass constraints, and the number of items required. Pressure vessels can be dangerous, and fatal accidents have occurred in the history of their development and operation.
Reactor pressure vesselA reactor pressure vessel (RPV) in a nuclear power plant is the pressure vessel containing the nuclear reactor coolant, core shroud, and the reactor core. Russian Soviet era RBMK reactors have each fuel assembly enclosed in an individual 8 cm diameter pipe rather than having a pressure vessel. Whilst most power reactors do have a pressure vessel, they are generally classified by the type of coolant rather than by the configuration of the vessel used to contain the coolant.
BrittlenessA material is brittle if, when subjected to stress, it fractures with little elastic deformation and without significant plastic deformation. Brittle materials absorb relatively little energy prior to fracture, even those of high strength. Breaking is often accompanied by a sharp snapping sound. When used in materials science, it is generally applied to materials that fail when there is little or no plastic deformation before failure. One proof is to match the broken halves, which should fit exactly since no plastic deformation has occurred.
Molten-Salt Reactor ExperimentThe Molten-Salt Reactor Experiment (MSRE) was an experimental molten salt reactor research reactor at the Oak Ridge National Laboratory (ORNL). This technology was researched through the 1960s, the reactor was constructed by 1964, it went critical in 1965, and was operated until 1969. The costs of a cleanup project were estimated at about $130 million. The MSRE was a 7.4 MWth test reactor simulating the neutronic "kernel" of a type of inherently safer epithermal thorium breeder reactor called the liquid fluoride thorium reactor.
Generation IV reactorGeneration IV reactors (Gen IV) 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.
Tensile testingTensile testing, also known as tension testing, is a fundamental materials science and engineering test in which a sample is subjected to a controlled tension until failure. Properties that are directly measured via a tensile test are ultimate tensile strength, breaking strength, maximum elongation and reduction in area. From these measurements the following properties can also be determined: Young's modulus, Poisson's ratio, yield strength, and strain-hardening characteristics.
Pebble-bed reactorThe pebble-bed reactor (PBR) is a design for a graphite-moderated, gas-cooled nuclear reactor. It is a type of very-high-temperature reactor (VHTR), one of the six classes of nuclear reactors in the Generation IV initiative. The basic design of pebble-bed reactors features spherical fuel elements called pebbles. These tennis ball-sized pebbles (approx. in diameter) are made of pyrolytic graphite (which acts as the moderator), and they contain thousands of micro-fuel particles called tristructural-isotropic (TRISO) particles.
Advanced heavy-water reactorThe 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. KAMINI is the world's first thorium-based experimental reactor. It is cooled and moderated by light water, fueled with uranium-233 metal produced by the thorium fuel cycle harnessed by the neighbouring FBTR reactor and produces 30 KW of thermal energy at full power.
Charpy impact testIn materials science, the Charpy impact test, also known as the Charpy V-notch test, is a standardized high strain rate test which determines the amount of energy absorbed by a material during fracture. Absorbed energy is a measure of the material's notch toughness. It is widely used in industry, since it is easy to prepare and conduct and results can be obtained quickly and cheaply. A disadvantage is that some results are only comparative. The test was pivotal in understanding the fracture problems of ships during World War II.
Molten salt reactorA 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. Two research MSRs operated in the United States in the mid-20th century. The 1950s Aircraft Reactor Experiment (ARE) was primarily motivated by the technology's compact size, while the 1960s Molten-Salt Reactor Experiment (MSRE) aimed to demonstrate a nuclear power plant using a thorium fuel cycle in a breeder reactor.
Lead-bismuth eutecticLead-Bismuth Eutectic or LBE is a eutectic alloy of lead (44.5 at%) and bismuth (55.5 at%) used as a coolant in some nuclear reactors, and is a proposed coolant for the lead-cooled fast reactor, part of the Generation IV reactor initiative. It has a melting point of 123.5 °C/255.3 °F (pure lead melts at 327 °C/621 °F, pure bismuth at 271 °C/520 °F) and a boiling point of 1,670 °C/3,038 °F. Lead-bismuth alloys with between 30% and 75% bismuth all have melting points below 200 °C/392 °F.
