A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

A method for dynamic pressure control during a multiphase injection is described wherein the pressures of fluids in the various reservoirs of a fluid delivery system are controlled to provide desired fluid delivery parameters. The methods include advancing the first drive member to expel the first fluid from the first reservoir into a conduit, wherein the fluid is pressurized to a first fluid pressure; measuring the first fluid pressure to provide a target value; while the second reservoir is in fluid isolation from the conduit, advancing or retracting the second drive member to increase or decrease the fluid pressure of the second fluid in the second reservoir to the target value; placing the second reservoir in fluid communication with the conduit; and advancing the second drive member to expel the second fluid from the second reservoir into the conduit.

A method for dynamic pressure control during a multiphase injection is described wherein the pressures of fluids in the various reservoirs of a fluid delivery system are controlled to provide desired fluid delivery parameters. The methods include advancing the first drive member to expel the first fluid from the first reservoir into a conduit, wherein the fluid is pressurized to a first fluid pressure; measuring the first fluid pressure to provide a target value; while the second reservoir is in fluid isolation from the conduit, advancing or retracting the second drive member to increase or decrease the fluid pressure of the second fluid in the second reservoir to the target value; placing the second reservoir in fluid communication with the conduit; and advancing the second drive member to expel the second fluid from the second reservoir into the conduit.

The present disclosure is directed to systems and methods useful for the construction and operation of a Modular Integrated Gas High-Temperature Reactor (MIGHTR). The MIGHTR includes a reactor core assembly disposed at least partially within a core baffle within a first high-pressure shell portion, a thermal transfer assembly disposed at least partially within a flow separation barrel within a second high-pressure shell portion. The longitudinal axes of the first high-pressure shell portion and the second high-pressure shell portion may be collinear. The reactor core assembly may be accessed horizontally for service, maintenance, and refueling. The core baffle may be flexibly displaceably coupled to the flow separation barrel. Coolant gas flows through the reactor core assembly and into the thermal transfer assembly where the temperature of the coolant gas is reduced. A plurality of coolant gas circulators circulate the cooled coolant gas from the thermal transfer assembly to the reactor core assembly.

The present disclosure is directed to systems and methods useful for the construction and operation of a Modular Integrated Gas High-Temperature Reactor (MIGHTR). The MIGHTR includes a reactor core assembly disposed at least partially within a core baffle within a first high-pressure shell portion, a thermal transfer assembly disposed at least partially within a flow separation barrel within a second high-pressure shell portion. The longitudinal axes of the first high-pressure shell portion and the second high-pressure shell portion may be collinear. The reactor core assembly may be accessed horizontally for service, maintenance, and refueling. The core baffle may be flexibly displaceably coupled to the flow separation barrel. Coolant gas flows through the reactor core assembly and into the thermal transfer assembly where the temperature of the coolant gas is reduced. A plurality of coolant gas circulators circulate the cooled coolant gas from the thermal transfer assembly to the reactor core assembly.

Nuclear reactor systems and associated devices and methods are described herein. A representative nuclear reactor system includes a reactor vessel having a barrier separating a core region from a shield region. A plurality of fuel rods containing a liquid nuclear fuel are positioned in the core region. A liquid moderator material is also positioned in the core region at least partially around the fuel rods. A plurality of heat exchangers can be positioned in the shield region, and a plurality of heat pipes can extend through the barrier. The moderator material is positioned to transfer heat received from the liquid nuclear fuel to the heat pipes, and the heat pipes are positioned to transfer heat received from the moderator material to the heat exchangers. The heat exchangers can transport the heat out of the system for use in one or more processes, such as generating electricity.

The invention refers to the area of nuclear power engineering, particularly to auxiliary devices for nuclear power plants, namely to the devices for installation of the outer heat insulation of a nuclear reactor vessel, and can be used at nuclear plants for recovery annealing of the VVER reactor vessel welds.

The objective to be achieved with the use of the claimed invention is to provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation.

Reduction of the temperature gradient through the thickness of the nuclear reactor vessel by heat insulation of the external reactor vessel surface, assurance of uniform physical properties for the reactor vessel metal and welds and reduction of thermal impacts on the surrounding structures during recovery annealing of the welds and/or base metal of the VVER reactor vessel shall be the technical result of this invention.

The technical result of the invention is ensured by the fact that the device for installation of the outer heat insulation of a nuclear reactor vessel includes a mobile transfer trolley equipped with the mechanism for its movement, a removable bearing rim located on the mobile transfer trolley with the reactor vessel heat insulation attached to it, at least two lifting devices located on the opposite sides of the reactor vessel at the level of its upper section, in this case the removable bearing rim is connected to the lifting devices with the possibility to lift and lower it.

Application of the claimed invention will provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation. Heat insulation of the external reactor vessel surface will ensure reduction of the temperature gradient through the thickness of the nuclear reactor vessel, uniformity of the physical properties for its metal and welds as well as reduction of thermal impacts on the surrounding structures during recovery annealing of the welds and/or base metal of the VVER reactor vessel.

The invention refers to the area of nuclear power engineering, particularly to auxiliary devices for nuclear power plants, namely to the devices for installation of the outer heat insulation of a nuclear reactor vessel, and can be used at nuclear plants for recovery annealing of the VVER reactor vessel welds.

The objective to be achieved with the use of the claimed invention is to provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation.

Reduction of the temperature gradient through the thickness of the nuclear reactor vessel by heat insulation of the external reactor vessel surface, assurance of uniform physical properties for the reactor vessel metal and welds and reduction of thermal impacts on the surrounding structures during recovery annealing of the welds and/or base metal of the VVER reactor vessel shall be the technical result of this invention.

The technical result of the invention is ensured by the fact that the device for installation of the outer heat insulation of a nuclear reactor vessel includes a mobile transfer trolley equipped with the mechanism for its movement, a removable bearing rim located on the mobile transfer trolley with the reactor vessel heat insulation attached to it, at least two lifting devices located on the opposite sides of the reactor vessel at the level of its upper section, in this case the removable bearing rim is connected to the lifting devices with the possibility to lift and lower it.

Application of the claimed invention will provide the possibility for installation and dismantling of heat insulation on the outer surface of the VVER reactor vessel in the confined space of the subpile room and at the increased level of ionizing radiation. Heat insulation of the external reactor vessel surface will ensure reduction of the temperature gradient through the thickness of the nuclear reactor vessel, uniformity of the physical properties for its metal and welds as well as reduction of thermal impacts on the surrounding structures during recovery annealing of the welds and/or base metal of the VVER reactor vessel.

A control rod for a nuclear fuel assembly is described herein that includes a neutron absorbing material having a melting point greater than 1500° C. that does not form a eutectic with a melting point less than 1500° C., and may further include a cladding material having a melting point greater than 1500° C. The cladding material is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material is selected from the group consisting of Gd.sub.2O.sub.3, Ir, B.sub.4C, Re, and Hf. The metal cladding or the absorbing material may be coated with an anti-oxidation coating of Cr with or without a Nb intermediate layer.