A sodium leak detection system for a nuclear reactor vessel includes a recirculation loop having an inlet and an outlet in communication with the annular space between the nuclear reactor vessel and the guard vessel. The recirculation loop and the annulus are filled with an inert gas, such as argon. The inert gas is doped with a known trace quantity of oxygen, typically in the single-digit ppm range up to about 1%. A recirculator forces the inert gas and oxygen to mix and flow throughout the annulus. The recirculation loop further includes a trace oxygen sensor that determines the concentration of oxygen in the inert gas. Because sodium reacts with oxygen, the trace oxygen sensor is monitored for a reduction in the oxygen level, which indicates a sodium leak into the annulus.
2.
MODULAR SODIUM PRESSURE SENSOR FOR NUCLEAR REACTOR
In a sodium fast reactor, a bypass pipe is fluidly coupled to the primary sodium pump discharge and diverts a portion of the primary sodium coolant to an instrument assembly. The instrument assembly includes a bypass tank and selectively swappable instrument modules. The instrument modules can be configured to measure flow, pressure, temperature, and fluid level, among other things. A modular pressure sensor that includes a diaphragm and a transducer may be easily swappable within the bypass tank. A guard pipe may extend above the reactor vessel head and house the transducer while the diaphragm is in contact with primary sodium, the transducer being located away from the thermal and radiological environment of the primary sodium. The modular pressure sensor is pre-calibrated and is accessible from above the reactor head for quick and efficient removal and replacement of the entire pressure sensor assembly.
G21C 17/022 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators
G01L 9/00 - Measuring steady or quasi-steady pressure of a fluid or a fluent solid material by electric or magnetic pressure-sensitive elementsTransmitting or indicating the displacement of mechanical pressure-sensitive elements, used to measure the steady or quasi-steady pressure of a fluid or fluent solid material, by electric or magnetic means
3.
CORROSION-RESISTANT COOLANT SALT AND METHOD FOR MAKING SAME
This document describes a method for reducing the corrosivity of certain magnesium salts. The salt product resulting from the method exhibits reduced corrosion of steels that come into contact with the salt relative to salt compositions that are not so treated. This makes such treated salts more efficient coolant salts as they will require less equipment replacement over time. The method uses magnesium metal to reduce unwanted impurities in the salts the reduced impurities are then removed as either gas or precipitate from the now purified salt. Without being bound to one particular theory, it is believed that the reduction of the level of impurities in the salt results in a salt with substantially reduced corrosiveness to steel.
A method of handling spent nuclear fuel assemblies immerses the spent nuclear fuel assemblies in water in a relatively short time period when compared to traditional methods. A spent nuclear fuel assembly is removed from a nuclear reactor core, inserted into a sodium removal machine having a receiver, a cleaning vessel, and an elevator. A cleaning fluid is applied to the cleaning vessel and fuel assembly, and the fuel assembly is flushed with water while in the cleaning vessel. The cleaning vessel is at least partially submerged in the spent fuel pool during cleaning to provide passive heat removal. The cleaning vessel is lowered by an elevator into the spent fuel pool. The fuel assembly may then be loaded into a rack and/or a cask for long-term storage.
G21C 19/32 - Apparatus for removing radioactive objects or materials from the reactor discharge area, e.g. to a storage placeApparatus for handling radioactive objects or materials within a storage place or removing them therefrom
G21C 19/08 - Means for heating fuel elements before introduction into the coreMeans for heating or cooling fuel elements after removal from the core
G21C 19/19 - Reactor parts specifically adapted to facilitate handling, e.g. to facilitate charging or discharging of fuel elements
Nuclear reactors; energy storage plants; nuclear power
plants; structural parts and fittings for nuclear power
plants. Demonstration of products in the fields of energy
production, nuclear energy and nuclear technology;
government advocacy, namely, promoting new and emerging
nuclear technology; new product commercialization services
in the fields of energy production, nuclear energy and
nuclear technology; promoting the benefits of nuclear
technology and nuclear energy to governments and
professionals in the energy industry; promoting public
awareness of carbon free energy initiatives, namely,
providing online information, news, and commentary in the
field of carbon-free energy initiatives; promoting public
awareness of nuclear energy by means of public advocacy.
6.
Anti-Proliferation Safeguards for Nuclear Fuel Salts
An anti-proliferation technique is disclosed to reduce the likelihood of nuclear proliferation due to the use fissionable fuel salts. The technique includes doping the fuel salt with one or more elements (referred to herein as activation dopants) that, upon exposure to neutrons such as would occur in the fuel salt when a reactor is in operation, undergo a nuclear reaction to, directly or indirectly, form highly active “protecting isotopes” (of the same element as the activation dopant or a different element). A sufficient mass of activation dopants is used so that the Figure of Merit (FOM) of the fuel salt is decreased to below 1.0 within some target number of days of fission. This allows the FOM of the fuel salt to be controlled so that the fuel becomes too dangerous to handle before to the creation of a significant amount of weaponizable isotopes.
G21C 3/54 - Fused salt, oxide, or hydroxide compositions
G21C 1/02 - Fast fission reactors, i.e. reactors not using a moderator
G21C 19/30 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products
Nuclear power plants and structural parts and fittings
therefor; energy storage plants; nuclear reactors. Demonstration of products in the fields of energy
production, nuclear energy and nuclear technology;
government advocacy, namely, promoting new and emerging
nuclear technology; new product commercialization services
in the fields of energy production, nuclear energy and
nuclear technology; promoting the benefits of nuclear
technology and nuclear energy to governments and
professionals in the energy industry; providing business
online information, news, and commentary in the field of
carbon-free energy initiatives; public advocacy to promote
awareness of nuclear energy.
A curvilinear electromagnetic pump is configured to follow a curve, such as by coupling multiple linear pump segments together that are offset by an angle with respect to each other. The curvilinear electromagnetic pump can curve within two dimensions, or within three dimensions. The curvilinear electromagnetic pump allows for more efficient arrangement of components and systems within a nuclear reactor vessel and allows a significantly reduced reactor vessel height as compared to a linear pump arranged vertically. The curvilinear electromagnetic pump may follow the curvature of the reactor vessel wall and may be entirely disposed near the bottom of the reactor vessel.
G21C 1/32 - Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles
G21C 5/10 - Means for supporting the complete structure
G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
G21C 13/024 - Supporting constructions for pressure vessels or containment vessels
G21C 13/04 - Arrangements for expansion and contraction
G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel
A nuclear reactor must be designed and constructed to withstand seismic events. A modular integrated reactor support structure includes seismic isolators that rely on viscous dampers and elastic support assemblies to provide three-dimensional seismic isolation to the reactor head, reactor vessel and reactor internal components. The gravitational load of the reactor head, reactor vessel, and reactor internals is supported through the reactor head, which transfers the load to a plurality of seismic isolators, which in turn, transfer the load to the basemat of the reactor building. A plurality of reactor support assembly blocks couple the reactor head to the seismic isolators while permitting thermal expansion and contraction of the reactor head while limiting rotational motion.
A method of removing cesium vapor from a cover gas stream in a nuclear reactor includes the steps of oxidizing the cesium vapor in the cover gas stream to yield cesium oxide particles and removing the cesium oxide particles using a particle filter. The method yields a filtered cover gas having zero to about 2% of the cesium vapor content of the initial cover gas stream, representing a reduction of at least about 98 percent.
B01D 53/82 - Solid phase processes with stationary reactants
B01D 39/20 - Other self-supporting filtering material of inorganic material, e.g. asbestos paper or metallic filtering material of non-woven wires
B01D 53/46 - Removing components of defined structure
G21C 19/303 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products specially adapted for gases
Nuclear reactors; Energy storage plants; Nuclear power plants; Structural parts and fittings for nuclear power plants Demonstration of products in the fields of energy production, nuclear energy and nuclear technology; Government advocacy, namely, promoting new and emerging nuclear technology; New product commercialization services in the fields of energy production, nuclear energy and nuclear technology; Promoting the benefits of nuclear technology and nuclear energy to governments and professionals in the energy industry; Promoting public awareness of carbon free energy initiatives, namely, providing online public policy information, news, and commentary in the field of carbon-free energy initiatives; Promoting public awareness of nuclear energy by means of public advocacy
(1) Nuclear reactors; energy storage plants; nuclear power plants; structural parts and fittings for nuclear power plants. (1) Demonstration of products in the fields of energy production, nuclear energy and nuclear technology; government advocacy, namely, promoting new and emerging nuclear technology; new product commercialization services in the fields of energy production, nuclear energy and nuclear technology; promoting the benefits of nuclear technology and nuclear energy to governments and professionals in the energy industry; promoting public awareness of carbon free energy initiatives, namely, providing online information, news, and commentary in the field of carbon-free energy initiatives; promoting public awareness of nuclear energy by means of public advocacy.
13.
TRACE OXYGEN SODIUM LEAK DETECTION IN NUCLEAR REACTOR ENCLOSURE
A sodium leak detection system for a nuclear reactor vessel includes a recirculation loop having an inlet and an outlet in communication with the annular space between the nuclear reactor vessel and the guard vessel. The recirculation loop and the annulus are filled with an inert gas, such as argon. The inert gas is doped with a known trace quantity of oxygen, typically in the single-digit ppm range up to about 1%. A recirculator forces the inert gas and oxygen to mix and flow throughout the annulus. The recirculation loop further includes a trace oxygen sensor that determines the concentration of oxygen in the inert gas. Because sodium reacts with oxygen, the trace oxygen sensor is monitored for a reduction in the oxygen level, which indicates a sodium leak into the annulus.
G01M 3/22 - Investigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using special tracer materials, e.g. dye, fluorescent material, radioactive material for pipes, cables, or tubesInvestigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using special tracer materials, e.g. dye, fluorescent material, radioactive material for pipe joints or sealsInvestigating fluid tightness of structures by using fluid or vacuum by detecting the presence of fluid at the leakage point using special tracer materials, e.g. dye, fluorescent material, radioactive material for valves
G01M 3/28 - Investigating fluid tightness of structures by using fluid or vacuum by measuring rate of loss or gain of fluid, e.g. by pressure-responsive devices, by flow detectors for pipes, cables, or tubesInvestigating fluid tightness of structures by using fluid or vacuum by measuring rate of loss or gain of fluid, e.g. by pressure-responsive devices, by flow detectors for pipe joints or sealsInvestigating fluid tightness of structures by using fluid or vacuum by measuring rate of loss or gain of fluid, e.g. by pressure-responsive devices, by flow detectors for valves
(1) Nuclear power plants and structural parts and fittings therefor; energy storage plants; nuclear reactors. (1) Demonstration of products in the fields of energy production, nuclear energy and nuclear technology; government advocacy, namely, promoting new and emerging nuclear technology; new product commercialization services in the fields of energy production, nuclear energy and nuclear technology; promoting the benefits of nuclear technology and nuclear energy to governments and professionals in the energy industry; providing business online information, news, and commentary in the field of carbon-free energy initiatives; public advocacy to promote awareness of nuclear energy.
Nuclear power plants and structural parts and fittings therefor; Energy storage plants; Nuclear reactors Demonstration of products in the fields of energy production, nuclear energy and nuclear technology; Government advocacy, namely, promoting new and emerging nuclear technology; New product commercialization services in the fields of energy production, nuclear energy and nuclear technology; Promoting the benefits of nuclear technology and nuclear energy to governments and professionals in the energy industry; Providing online public policy information, news, and commentary in the field of carbon-free energy initiatives; Promoting public awareness of nuclear energy by means of public advocacy
16.
AMBIENT TEMPERATURE THERMAL ADAPTER FOR SUPERCRITICAL CARBON DIOXIDE POWER CYCLE
A refrigeration cycle is connected to a supercritical carbon dioxide (sCO2) power cycle by incorporating a CO2 condenser which is also the refrigeration cycle evaporator. The heat rejected by the carbon dioxide is absorbed by the refrigerant in the CO2 condenser. Work is done on the refrigerant to raise its temperature above the local ambient temperature. The heat absorbed from the power cycle and the work required to raise the refrigerant temperature is rejected to the ambient environment, which may be either air, earth, or water. This results in improved efficiencies in carbon dioxide power cycles without regard to the ambient temperature by forcing a transcritical phase change even when ambient temperatures are above the carbon dioxide critical point. A two-phase heat transfer on both sides of the CO2 condenser further improves efficiency of the sCO2 power cycle.
F01K 25/10 - Plants or engines characterised by use of special working fluids, not otherwise provided forPlants operating in closed cycles and not otherwise provided for using special vapours the vapours being cold, e.g. ammonia, carbon dioxide, ether
17.
AMBIENT TEMPERATURE THERMAL ADAPTER FOR SUPERCRITICAL CARBON DIOXIDE POWER CYCLE
A refrigeration cycle is connected to a supercritical carbon dioxide (sCO2) power cycle by incorporating a CO2 condenser which is also the refrigeration cycle evaporator. The heat rejected by the carbon dioxide is absorbed by the refrigerant in the CO2 condenser. Work is done on the refrigerant to raise its temperature above the local ambient temperature. The heat absorbed from the power cycle and the work required to raise the refrigerant temperature is rejected to the ambient environment, which may be either air, earth, or water. This results in improved efficiencies in carbon dioxide power cycles without regard to the ambient temperature by forcing a transcritical phase change even when ambient temperatures are above the carbon dioxide critical point. A two-phase heat transfer on both sides of the CO2 condenser further improves efficiency of the sCO2 power cycle.
F01K 25/10 - Plants or engines characterised by use of special working fluids, not otherwise provided forPlants operating in closed cycles and not otherwise provided for using special vapours the vapours being cold, e.g. ammonia, carbon dioxide, ether
18.
CHLORIDE BASED VOLATILITY FOR THE RECOVERY OF URANIUM FROM USED NUCLEAR FUEL CONTAINING URANIUM METAL
Disclosed herein are methods and systems for the recovery of uranium from used nuclear fuel containing nuclear metal using nonaqueous chemistry without electrometallurgy techniques, such as pyroprocessing. The methods and systems use chloride-based volatility (CBV) as the basis for uranium recovery from the used nuclear fuel containing uranium metal. The solid used nuclear fuel containing uranium metal is maintained under UCl 4 sublimating conditions in a chlorinating environment for a period of time sufficient to convert at least some uranium metal into UCl 4.
In a sodium-cooled fast reactor, a breached fuel pin releases fission and activation products into the primary sodium coolant and into the cover gas. A portion of the primary sodium coolant is diverted to a distillation system where its constituents can be separated based on volatility. Condensing these constituents enables precise measurement of their concentration without delays caused by waiting for the decay of radioactive sodium background scatter, enhancing the detection and quantification of the dilute constituents. Additionally, the method continuously determines the initial concentration of constituents in the feed. Quantification is achieved using detectors, facilitating the continuous assessment of dilute constituent concentrations in the sodium coolant feed stream.
G21C 17/025 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators for monitoring liquid metal coolants
G21C 17/10 - Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain
A nuclear reactor core includes a plurality of core assemblies. At least some of the core assemblies are formed with bi-metallic assembly (BiMA) devices having two or more metallic materials with different coefficients of thermal expansion such that the BiMA devices are configured to preferentially deform in response to an increase in temperature. The preferential deformation is a bowing deformation and is directed toward the center of the core with a predetermined stroke such that the BiMA devices impart forces on adjacent core assemblies sufficient to achieve mechanical core lock-up during a startup routine and before the reactor reaches full power.
A nuclear reactor is designed and constructed to withstand seismic events. A mechanical hold down and discriminator may be provided as cooperating structure between the core receptacle and a nozzle of a core assembly. One or more receivers may be formed on the nozzle that cooperate with corresponding protrusions in a core receptacle. As the nozzle is inserted into the receptacle, the protrusion engages with the receiver to inhibit unintentional withdrawal of the nozzle from the receiver. A first pattern of receivers on a nozzle may align with a first pattern of protrusions within a receptacle to enable the nozzle to fit within a first receptacle. A second pattern of receivers formed on a second nozzle may inhibit the second nozzle from being inserted into the first receptacle and thereby provide a discrimination feature to ensure proper core assemblies are located in the correct location within a nuclear reactor core.
G21C 3/12 - Means forming part of the element for locating it within the reactor coreExternal spacers for this purpose
G21C 3/32 - Bundles of parallel pin-, rod-, or tube-shaped fuel elements
G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles
A nuclear reactor must be designed and constructed to withstand seismic events. A modular integrated reactor support structure includes seismic isolators that rely on viscous dampers and elastic support assemblies to provide three-dimensional seismic isolation to the reactor head, reactor vessel and reactor internal components. The gravitational load of the reactor head, reactor vessel, and reactor internals is supported through the reactor head, which transfers the load to a plurality of seismic isolators, which in turn, transfer the load to the basemat of the reactor building. A plurality of reactor support assembly blocks couple the reactor head to the seismic isolators while permitting thermal expansion and contraction of the reactor head while limiting rotational motion.
E04H 9/02 - Buildings, groups of buildings or shelters adapted to withstand or provide protection against abnormal external influences, e.g. war-like action, earthquake or extreme climate withstanding earthquake or sinking of ground
25.
Ambient temperature thermal adapter for supercritical carbon dioxide power cycle
A refrigeration cycle is connected to a supercritical carbon dioxide (sCO2) power cycle by incorporating a CO2 condenser which is also the refrigeration cycle evaporator. The heat rejected by the carbon dioxide is absorbed by the refrigerant in the CO2 condenser. Work is done on the refrigerant to raise its temperature above the local ambient temperature. The heat absorbed from the power cycle and the work required to raise the refrigerant temperature is rejected to the ambient environment, which may be either air, earth, or water. This results in improved efficiencies in carbon dioxide power cycles without regard to the ambient temperature by forcing a transcritical phase change even when ambient temperatures are above the carbon dioxide critical point. A two-phase heat transfer on both sides of the CO2 condenser further improves efficiency of the sCO2 power cycle.
F01K 25/10 - Plants or engines characterised by use of special working fluids, not otherwise provided forPlants operating in closed cycles and not otherwise provided for using special vapours the vapours being cold, e.g. ammonia, carbon dioxide, ether
F01K 3/18 - Plants characterised by the use of steam or heat accumulators, or intermediate steam heaters, therein having heaters
F01K 7/32 - Steam engine plants characterised by the use of specific types of enginePlants or engines characterised by their use of special steam systems, cycles or processesControl means specially adapted for such systems, cycles or processesUse of withdrawn or exhaust steam for feed-water heating the engines using steam of critical or over-critical pressure
F01K 9/00 - Steam engine plants characterised by condensers arranged or modified to co-operate with the engines
F25B 9/00 - Compression machines, plants or systems, in which the refrigerant is air or other gas of low boiling point
26.
TAG GAS CAPSULE WITH THERMALLY BASED RELEASE SYSTEM
A tag gas capsule utilizes a thermally based release system to allow the tag gas to flow through a fuel pin and to be detectable upon a leak in the fuel pin. A tag gas capsule is fabricated that has an escape pathway for the tag gas wherein the escape pathway is sealed by a thermally sensitive material. During fabrication, the thermally sensitive material is melted into the tag gas capsule and seals a flow hole in the tag gas capsule. The capsule is then filled with a tag gas and sealed to prevent escape of the tag gas. Once the capsule has been installed into a fuel pin, the capsule can be heated to melt the thermally sensitive material and expose the flow hole. The tag gas can then flow out of the flow hole and into the fuel pin.
G21C 3/18 - Internal spacers or other non-active material within the casing, e.g. compensating for expansion of fuel rods or for compensating excess reactivity
A tag gas capsule utilizes a thermally based release system to allow the tag gas to flow through a fuel pin and to be detectable upon a leak in the fuel pin. A tag gas capsule is fabricated that has an escape pathway for the tag gas wherein the escape pathway is sealed by a thermally sensitive material. During fabrication, the thermally sensitive material is melted into the tag gas capsule and seals a flow hole in the tag gas capsule. The capsule is then filled with a tag gas and sealed to prevent escape of the tag gas. Once the capsule has been installed into a fuel pin, the capsule can be heated to melt the thermally sensitive material and expose the flow hole. The tag gas can then flow out of the flow hole and into the fuel pin.
G21C 3/18 - Internal spacers or other non-active material within the casing, e.g. compensating for expansion of fuel rods or for compensating excess reactivity
Flow channels are formed into a nuclear reactor core receptacle and provide a flow path for leaking primary coolant. The flow path redirects the leaking coolant from an upwardly vertical direction, in which the mass flow would impart a lift off force to a core assembly seated within the receptacle, to a horizontal flow path direction which maintains hydraulic balance of the core assembly. The flow channels in the nuclear reactor core receptacle provide a passive measure to allow hydraulic hold down of the core assemblies even in the absence of mechanical hold downs.
G21C 5/06 - Means for locating or supporting fuel elements
G21C 15/02 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements
29.
SYSTEMS AND METHODS FOR NUCLEAR CORE ASSEMBLY RECEPTACLE LEAKAGE DIVERTER
Flow channels are formed into a nuclear reactor core receptacle and provide a flow path for leaking primary coolant. The flow path redirects the leaking coolant from an upwardly vertical direction, in which the mass flow would impart a lift off force to a core assembly seated within the receptacle, to a horizontal flow path direction which maintains hydraulic balance of the core assembly. The flow channels in the nuclear reactor core receptacle provide a passive measure to allow hydraulic hold down of the core assemblies even in the absence of mechanical hold downs.
G21C 5/06 - Means for locating or supporting fuel elements
G21C 15/02 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements
30.
CHLORIDE BASED VOLATILITY FOR THE RECOVERY OF URANIUM FROM NUCLEAR FUEL SALT
Disclosed herein are methods and systems for the recovery of uranium from used fuel salts using nonaqueous chemistry without electrometallurgy techniques. The used fuel salts are maintained in a molten state in a chlorination chamber, and are then chlorinated for a period of time sufficient to convert at least some UCl 3 into a different uranium chloride, thereby generating a uranium-containing gas phase. The uranium-containing gas phase is separated from the resulting residue, and is condensed for collection.
Disclosed herein are methods and systems for the recovery of uranium from used nuclear fuel using nonaqueous chemistry without electrometallurgy techniques, such as pyroprocessing. The methods and systems use chloride-based volatility (CBV) as the basis for uranium recovery from the used nuclear fuel. The solid used nuclear fuel containing solid uranium oxide is maintained under UCl 4 sublimating conditions in a chlorinating environment for a period of time sufficient to convert at least some solid uranium oxide into UCl 4 in a gas phase.
33 into a different uranium chloride, thereby generating a uranium-containing gas phase. The uranium-containing gas phase is separated from the resulting residue, and is condensed for collection.
A fuel element has a ratio of area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis to total area of the interior volume in the cross¬ section of the tubular fuel element (i.e.: smear density) that varies with position along the longitudinal axis. The ratio can vary with position along the longitudinal axis between a minimum of 0.30 and a maximum of 1.0. Increasing the ratio above and below the peak bum-up location associated with conventional systems reduces the peak bum-up and flattens and shifts the bum-up distribution, which is preferably Gaussian. The longitudinal variation can be implemented in fuel assemblies using fuel bodies, such as pellets, rods or annuli, or fuel in the form of metal sponge and meaningfully increases efficiency of fuel utilization.
A heat exchanger is formed of a shell, a mounting plate, a plurality of tube modules disposed within the shell through holes formed in the mounting plate, and nozzles coupled to each of the tube modules. The nozzles may be selectively removed from the tube modules to allow inspection or repair of the tube modules. Similarly, the tube modules may be selectively removable from the shell through the mounting plate to allow inspection, repair, or replacement of individual tube modules.
F22B 37/00 - Component parts or details of steam boilers
F28D 1/00 - Heat-exchange apparatus having stationary conduit assemblies for one heat-exchange medium only, the media being in contact with different sides of the conduit wall, in which the other heat-exchange medium is a large body of fluid, e.g. domestic or motor car radiators
G21C 1/32 - Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
A nuclear reactor is configured with a primary coolant loop for transferring heat away from the nuclear reactor core. In a shutdown event, the primary coolant pump may stop pumping primary coolant through the reactor core, resulting in decay heat buildup within the reactor core. An inertial energy coast down system can store kinetic energy while the nuclear reactor is operating and then release the stored kinetic energy to cause the primary coolant to continue to flow through the nuclear reactor core to remove decay heat. The inertial energy coast down system may include an impeller and a flywheel having a mass. During normal reactor operation, the flowing primary coolant spins up the impeller and flywheel, and upon a shutdown event where the primary coolant pump stops pumping, the flywheel and impeller can cause the primary coolant to continue to flow during a coast down of the flywheel and impeller.
G21C 1/32 - Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles
G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
G21C 13/04 - Arrangements for expansion and contraction
G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel
A heat exchanger is formed of a shell, a mounting plate, a plurality of tube modules disposed within the shell through holes formed in the mounting plate, and nozzles coupled to each of the tube modules. The nozzles may be selectively removed from the tube modules to allow inspection or repair of the tube modules. Similarly, the tube modules may be selectively removable from the shell through the mounting plate to allow inspection, repair, or replacement of individual tube modules.
F28D 1/00 - Heat-exchange apparatus having stationary conduit assemblies for one heat-exchange medium only, the media being in contact with different sides of the conduit wall, in which the other heat-exchange medium is a large body of fluid, e.g. domestic or motor car radiators
G21C 1/32 - Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
F22B 37/00 - Component parts or details of steam boilers
Disclosed herein are methods and systems for the recovery of uranium from used fuel salts using nonaqueous chemistry without electrometallurgy techniques. The used fuel salts are maintained in a molten state in a chlorination chamber, and are then chlorinated for a period of time sufficient to convert at least some UCl3 into a different uranium chloride, thereby generating a uranium-containing gas phase. The uranium-containing gas phase is separated from the resulting residue, and is condensed for collection.
A method of handling spent nuclear fuel assemblies immerses the spent nuclear fuel assemblies in water in a relatively short time period when compared to traditional methods. A spent nuclear fuel assembly is removed from a nuclear reactor core, inserted into a sodium removal machine having a receiver, a cleaning vessel, and an elevator. A cleaning fluid is applied to the cleaning vessel and fuel assembly, and the fuel assembly is flushed with water while in the cleaning vessel. The cleaning vessel is at least partially submerged in the spent fuel pool during cleaning to provide passive heat removal. The cleaning vessel is lowered by an elevator into the spent fuel pool. The fuel assembly may then be loaded into a rack and/or a cask for long-term storage.
A method of handling spent nuclear fuel assemblies immerses the spent nuclear fuel assemblies in water in a relatively short time period when compared to traditional methods. A spent nuclear fuel assembly is removed from a nuclear reactor core, inserted into a sodium removal machine having a receiver, a cleaning vessel, and an elevator. A cleaning fluid is applied to the cleaning vessel and fuel assembly, and the fuel assembly is flushed with water while in the cleaning vessel. The cleaning vessel is at least partially submerged in the spent fuel pool during cleaning to provide passive heat removal. The cleaning vessel is lowered by an elevator into the spent fuel pool. The fuel assembly may then be loaded into a rack and/or a cask for long-term storage.
G21C 19/00 - Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
G21C 19/19 - Reactor parts specifically adapted to facilitate handling, e.g. to facilitate charging or discharging of fuel elements
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
G21C 19/24 - Arrangements for obtaining access to the interior of a pressure vessel whilst the reactor is operating by using an auxiliary vessel which is temporarily sealed to the pressure vessel
G21C 19/32 - Apparatus for removing radioactive objects or materials from the reactor discharge area, e.g. to a storage placeApparatus for handling radioactive objects or materials within a storage place or removing them therefrom
G21F 5/00 - Transportable or portable shielded containers
A fast neutron nuclear reactor contains a nuclear reactor core having an array of device locations. Some device locations in the nuclear reactor core contain fissile and fertile nuclear fuel assembly devices. One or more other device locations in the nuclear reactor core contain Doppler reactivity augmentation devices that amplify the negativity of the Doppler reactivity coefficient within the nuclear reactor core. In some implementations, a Doppler reactivity augmentation device can also reduce the coolant temperature coefficient within the nuclear reactor core. Accordingly, a Doppler reactivity augmentation device contributes to a more stable nuclear reactor core.
G21C 7/06 - Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section
G21C 1/02 - Fast fission reactors, i.e. reactors not using a moderator
G21C 3/42 - Selection of substances for use as reactor fuel
G21C 7/02 - Control of nuclear reaction by using self-regulating properties of reactor materials
G21C 7/08 - Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section by displacement of solid control elements, e.g. control rods
Energy storage plants; structural parts and fittings for
nuclear power plants; nuclear power plants; nuclear
reactors. Demonstration of products in the fields of energy
production, nuclear energy and nuclear technology;
government advocacy, namely, promoting new and emerging
nuclear technology; new product commercialization services
in the fields of energy production, nuclear energy and
nuclear technology; promoting the benefits of nuclear
technology and nuclear energy to governments and
professionals in the energy industry; providing online
commercial information and news in the field of carbon-free
energy initiatives; advertising services to promote public
awareness of nuclear energy by means of public advocacy.
A nuclear reactor includes a heat exchanger that transfers thermal energy from a primary reactor coolant to a secondary coolant. The heat exchanger is formed with a hot flow channel, a cold flow channel, and a porous layer between the hot flow channel and the cold flow channel. The porous layer may be thermally insulative to reduce the efficiency of thermal energy transfer from the hot flow channel to the cold flow channel. The porous layer may have a control gas passed therethrough that can be tailored to control the thermal energy transfer through the porous layer. The control gas can be tested for leakage within the heat exchanger. The control gas may also be used to sequester fission or activation products.
Disclosed embodiments include fuel assemblies, methods of making a fuel element, and methods of using a fuel element. A fuel element includes fuel, a fuel liner, and a cladding. The liner may be formed of one, two, three, or more layers of different materials, including different alloys have a different primary metallic component. The cladding may likewise be formed of one, two, three, or more layers of different materials. The different materials may include different alloys, different compositions, and/or different alloys in which the primary constituent of the alloy is a different material.
G21C 3/18 - Internal spacers or other non-active material within the casing, e.g. compensating for expansion of fuel rods or for compensating excess reactivity
G21C 3/07 - CasingsJackets characterised by their material, e.g. alloys
45.
Method of constructing a nuclear reactor having reactor core and control elements supported by reactor vessel head
A nuclear reactor is designed to couple the load path of the control elements with the reactor core, thus reducing the opportunity for differential movement between the control elements and the reactor core. A cartridge core barrel can be fabricated in a manufacturing facility to include the reactor core, control element supports, and control element drive system. The cartridge core barrel can be mounted to a reactor vessel head. Thus, any movement, such as through seismic forces, transmits an equal direction and magnitude to the control elements and the reactor core. This arrangement reduces the opportunity for differential movement between the control elements and the reactor core.
This document describes a method for reducing the corrosivity of certain magnesium salts. The salt product resulting from the method exhibits reduced corrosion of steels that come into contact with the salt relative to salt compositions that are not so treated. This makes such treated salts more efficient coolant salts as they will require less equipment replacement over time. The method uses magnesium metal to reduce unwanted impurities in the salts the reduced impurities are then removed as either gas or precipitate from the now purified salt. Without being bound to one particular theory, it is believed that the reduction of the level of impurities in the salt results in a salt with substantially reduced corrosiveness to steel.
(1) Energy storage plants; structural parts and fittings for nuclear power plants; nuclear power plants; nuclear reactors. (1) Demonstration of products in the fields of energy production, nuclear energy and nuclear technology; government advocacy, namely, promoting new and emerging nuclear technology; new product commercialization services in the fields of energy production, nuclear energy and nuclear technology; promoting the benefits of nuclear technology and nuclear energy to governments and professionals in the energy industry; providing online commercial information and news in the field of carbon-free energy initiatives; advertising services to promote public awareness of nuclear energy by means of public advocacy.
(Based on Use in Commerce) Energy storage plants; Structural parts and fittings for nuclear power plants; (Based on Intent To Use) ; Nuclear power plants; Nuclear reactors Demonstration of products in the fields of energy production, nuclear energy and nuclear technology; Government advocacy, namely, promoting new and emerging nuclear technology; New product commercialization services in the fields of energy production, nuclear energy and nuclear technology; Promoting the benefits of nuclear technology and nuclear energy to governments and professionals in the energy industry; Promoting public awareness of carbon free energy initiatives, namely, providing online information, news, and commentary in the field of carbon-free energy initiatives; Public advocacy to promote awareness of nuclear energy
51.
MODULAR MANUFACTURE, DELIVERY, AND ASSEMBLY OF NUCLEAR REACTOR BUILDING SYSTEMS
A nuclear reactor is constructed in sub-modules and super modules which are manufactured, packaged, and shipped to a construction site. At least some of the modules are packaged in suitable shielding containers or portions of containers, which may be steel. The modules are assembled on-site, and some of the modules remain within their respective shipping containers after assembly. One or more of the shipping containers may be used as concrete forms to support the pouring of concrete in between selected modules. The concrete may be used for structural support, shielding, or both.
E04H 5/02 - Buildings or groups of buildings for industrial purposes, e.g. for power-plants or factories
E04B 1/16 - Structures made from masses, e.g. concrete, cast or similarly formed in situ with or without making use of additional elements, such as permanent forms, sub-structures to be coated with load-bearing material
G21C 3/16 - Details of the construction within the casing
G21C 3/322 - Means to influence the coolant flow through or around the bundles
G21C 3/328 - Relative disposition of the elements in the bundle lattice
An in-vessel fuel transfer machine may be permanently affixed to a nuclear reactor and remain in place during power operations. The in-vessel fuel transfer machine may include a pantograph machine that positions a grapple in order to access any fuel socket location within the core and move any of the core assemblies between the core, an in-vessel fuel storage area, and a fuel elevator. The grapple may be positioned through a combination of movements, such as, rotating a rotating plug assembly, rotating the in-vessel fuel transfer machine, extending the pantograph arms, and shuttling the grapple along a leg. The grapple may be compliant to accommodate deformed core assemblies and may be configured to pivot to more closely align to an eccentric core assembly handling socket or be moveable in a horizontal plane to accommodate a deformed core assembly during insertion or withdrawal.
G21C 19/105 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements with grasping or spreading coupling elements
G21C 19/16 - Articulated or telescopic chutes or tubes for connection to channels in the reactor core
G21C 19/18 - Apparatus for bringing fuel elements to the reactor charge area, e.g. from a storage place
53.
METHOD FOR ONLINE RADIOISOTOPE MEASUREMENT FOR FAILED FUEL CHARACTERIZATION IN PRIMARY SODIUM SYSTEMS
A failed fuel pin emits cesium into the primary sodium coolant and xenon into the cover gas in a reactor vessel (402). A pipe (408) containing radioactive liquid sodium accepts flowing primary sodium from the reactor vessel. A radiation detector (416) is positioned adjacent the pipe such that gamma radiation emitted from the pipe can be measured. The pipe may be isolated to increase detection limits by allowing short-lived isotopes to decay. The isotopic ratio of 137Cs/134Cs can be measured, which can be used to determine the burnup of a fuel assembly from within the core, and therefore, the failed fuel assembly can be identified based at least in part on the burnup. Further, mass spectrometry may be used to measure the ratio of a stable and unstable xenon isotope. The identification techniques may be used in conjunction to quickly identify a failed fuel assembly in-situ and during reactor operation.
A failed fuel pin emits cesium into the primary sodium coolant and xenon into the cover gas in a reactor vessel. A pipe containing radioactive liquid sodium accepts flowing primary sodium from the reactor vessel. A radiation detector is positioned adjacent the pipe such that gamma radiation emitted from the pipe can be measured. The pipe may be isolated to increase detection limits by allowing short-lived isotopes to decay. The isotopic ratio of 137Cs/134Cs can be measured, which can be used to determine the burnup of a fuel assembly from within the core, and therefore, the failed fuel assembly can be identified based at least in part on the burnup. Further, mass spectrometry may be used to measure the ratio of a stable and unstable xenon isotope. The identification techniques may be used in conjunction to quickly identify a failed fuel assembly in-situ and during reactor operation.
A nuclear reactor core includes a plurality of core assemblies. The core assemblies have a cooperating structure formed at one or more load pads that mechanically couple the plurality of core assemblies together to limit relative motion between core assemblies in a kinematically determinate way. A shear key on one core assembly is configured to fit in a tab slot on an adjacent core assembly. Motion of one core assembly is transferred to a second core assembly and the core assemblies move together.
G21C 17/025 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators for monitoring liquid metal coolants
G21C 17/032 - Reactor-coolant flow measuring or monitoring
G21C 17/035 - Moderator- or coolant-level detecting devices
In a sodium fast reactor, a bypass pipe is fluidly coupled to the primary sodium pump discharge and diverts a portion of the primary sodium coolant to an instrument assembly. The bypass pipe routes flowing sodium upward toward the reactor head where it fluidly couples to the instrument assembly. The instrument assembly includes an instrument tank and selectively swappable instrument modules. The instrument modules can be configured to measure flow, pressure, temperature, and fluid level, among other things. The instrument assembly is located relatively close to the reactor head and close to the sodium level in the sodium pool and is accessible from above the reactor head for quick and efficient removal and replacement of the entire instrument assembly or individual instruments.
G21C 17/025 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators for monitoring liquid metal coolants
G21C 17/032 - Reactor-coolant flow measuring or monitoring
G21C 17/035 - Moderator- or coolant-level detecting devices
60.
System including a bypass pipe that routes some sodium coolant from a primary pump discharge upward to sodium measuring instruments located near a nuclear reactor head
In a sodium fast reactor, a bypass pipe is fluidly coupled to the primary sodium pump discharge and diverts a portion of the primary sodium coolant to an instrument assembly. The bypass pipe routes flowing sodium upward toward the reactor head where it fluidly couples to the instrument assembly. The instrument assembly includes an instrument tank and selectively swappable instrument modules. The instrument modules can be configured to measure flow, pressure, temperature, and fluid level, among other things. The instrument assembly is located relatively close to the reactor head and close to the sodium level in the sodium pool and is accessible from above the reactor head for quick and efficient removal and replacement of the entire instrument assembly or individual instruments.
G21C 17/025 - Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators for monitoring liquid metal coolants
A nuclear reactor core includes a plurality of core assemblies. The core assemblies have a cooperating structure formed at one or more load pads that mechanically couple the plurality of core assemblies together to limit relative motion between core assemblies in a kinematically determinate way. A shear key on one core assembly is configured to fit in a tab slot on an adjacent core assembly. Motion of one core assembly is transferred to a second core assembly and the core assemblies move together.
G21C 19/105 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements with grasping or spreading coupling elements
G21C 19/18 - Apparatus for bringing fuel elements to the reactor charge area, e.g. from a storage place
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
G21C 19/26 - Arrangements for removing jammed or damaged fuel elements or control elementsArrangements for moving broken parts thereof
63.
IN-VESSEL CORE COMPONENT HANDLING SYSTEMS AND METHODS
An in-vessel fuel transfer machine may be permanently affixed to a nuclear reactor and remain in place during power operations. The in-vessel fuel transfer machine may include a pantograph machine that positions a grapple in order to access any fuel socket location within the core and move any of the core assemblies between the core, an in-vessel fuel storage area, and a fuel elevator. The grapple may be positioned through a combination of movements, such as, rotating a rotating plug assembly, rotating the in-vessel fuel transfer machine, extending the pantograph arms, and shuttling the grapple along a leg. The grapple may be compliant to accommodate deformed core assemblies and may be configured to pivot to more closely align to an eccentric core assembly handling socket or be moveable in a horizontal plane to accommodate a deformed core assembly during insertion or withdrawal.
G21C 19/105 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements with grasping or spreading coupling elements
G21C 19/18 - Apparatus for bringing fuel elements to the reactor charge area, e.g. from a storage place
G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
G21C 19/26 - Arrangements for removing jammed or damaged fuel elements or control elementsArrangements for moving broken parts thereof
Energy storage plants; nuclear power plants and structural
parts and fittings therefor; nuclear reactors. Demonstration of products in the fields of energy
production, nuclear energy and nuclear technology;
Government advocacy, namely, promoting new and emerging
nuclear technology; new product commercialization services
in the fields of energy production, nuclear energy and
nuclear technology; promoting the benefits of nuclear
technology and nuclear energy to governments and
professionals in the energy industry; providing business
information via a website in the field of carbon-free energy
initiatives; public advocacy to promote awareness of nuclear
energy.
A masking element with an opening is disposed on the side of a core support structure. A flow stack wall defines a plurality of inlets. At least one inlet aligns with the masking element opening when the flow stack is mated with the masking element. A flow control assembly within the flow stack is configured to restrict flow of fluid within the flow stack.
A masking element with an opening is disposed on the side of a core support structure. A flow stack wall defines a plurality of inlets. At least one inlet aligns with the masking element opening when the flow stack is mated with the masking element. A flow control assembly within the flow stack is configured to restrict flow of fluid within the flow stack.
A method of removing cesium vapor from a cover gas stream in a nuclear reactor includes the steps of oxidizing the cesium vapor in the cover gas stream to yield cesium oxide particles and removing the cesium oxide particles using a particle filter. The method yields a filtered cover gas having zero to about 2% of the cesium vapor content of the initial cover gas stream, representing a reduction of at least about 98 percent.
B01D 53/82 - Solid phase processes with stationary reactants
B01D 39/20 - Other self-supporting filtering material of inorganic material, e.g. asbestos paper or metallic filtering material of non-woven wires
B01D 53/46 - Removing components of defined structure
G21C 19/303 - Arrangements for introducing fluent material into the reactor coreArrangements for removing fluent material from the reactor core with continuous purification of circulating fluent material, e.g. by extraction of fission products specially adapted for gases
71.
OXIDATION OF CESIUM AS METHOD FOR REMOVING CESIUM VAPOR FROM COVER GAS IN NUCLEAR REACTORS
A method of removing cesium vapor from a cover gas stream in a nuclear reactor includes the steps of oxidizing the cesium vapor in the cover gas stream to yield cesium oxide particles and removing the cesium oxide particles using a particle filter. The method yields a filtered cover gas having zero to about 2% of the cesium vapor content of the initial cover gas stream, representing a reduction of at least about 98 percent.
A method of removing cesium vapor from a cover gas stream in a nuclear reactor includes the steps of oxidizing the cesium vapor in the cover gas stream to yield cesium oxide particles and removing the cesium oxide particles using a particle filter. The method yields a filtered cover gas having zero to about 2% of the cesium vapor content of the initial cover gas stream, representing a reduction of at least about 98 percent.
A nuclear reactor is configured with an intermediate coolant loop for transferring thermal energy from the reactor core for a useful purpose. The intermediate coolant loop includes a bypass flowpath with an air heat exchanger for dumping reactor heat during startup and/or shutdown. A fluidic diode along the bypass flowpath asymmetrically restricts flow across the bypass flowpath, inhibiting flow in a first flow direction during a full power operating condition and allowing a relatively uninhibited flow in a second direction during a startup and/or shut down low power operating condition.
G21C 1/32 - Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles
G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel
(1) Energy storage plants; nuclear power plants and structural parts and fittings therefor; nuclear reactors. (1) Demonstration of energy production products, nuclear energy products, and nuclear technology products; Government advocacy, namely, promoting new and emerging nuclear technology; new product commercialization services in the fields of energy production, nuclear energy and nuclear technology; promoting the benefits of nuclear technology and nuclear energy to governments and professionals in the energy industry; providing business information via a website in the field of carbon-free energy initiatives; public advocacy to promote awareness of nuclear energy.
Demonstration of products in the fields of energy production, nuclear energy and nuclear technology; Government advocacy, namely, promoting new and emerging nuclear technology; New product commercialization services in the fields of energy production, nuclear energy and nuclear technology; Promoting the benefits of nuclear technology and nuclear energy to governments and professionals in the energy industry; Providing online information, news, and commentary in the field of carbon-free energy initiatives; Public advocacy to promote awareness of nuclear energy
A fuel element has a ratio of area of fissionable nuclear fuel in a cross-section of the tubular fuel element perpendicular to the longitudinal axis to total area of the interior volume in the cross-section of the tubular fuel element that varies with position along the longitudinal axis. The ratio can vary with position along the longitudinal axis between a minimum of 0.30 and a maximum of 1.0. Increasing the ratio above and below the peak burn-up location associated with conventional systems reduces the peak burn-up and flattens and shifts the burn-up distribution, which is preferably Gaussian. The longitudinal variation can be implemented in fuel assemblies using fuel bodies, such as pellets, rods or annuli, or fuel in the form of metal sponge and meaningfully increases efficiency of fuel utilization.
Embodiments of methods for improving mesophase pitch for carbon fiber production using supercritical carbon dioxide are described. The methods improve the relative amount and quality of mesophase pitch in feedstocks, such as coal tar, already having at least some mesophase pitch. One particular method includes performing a sCO2/toluene extraction on the coal tar to obtain a toluene insoluble fraction of the coal tar; mixing the toluene insoluble fraction with sCO2 to obtain a sCO2/toluene insoluble fraction mixture; and extruding the sCO2/toluene insoluble fraction mixture, thereby separating the sCO2 from the toluene insoluble fraction to obtain fibers of mesophase pitch.
C10C 3/08 - Working-up pitch, asphalt, bitumen by selective extraction
C01B 32/05 - Preparation or purification of carbon not covered by groups , , ,
D01F 9/15 - Carbon filamentsApparatus specially adapted for the manufacture thereof by decomposition of organic filaments from pitch or distillation residues from coal pitch
79.
HIGH ASSAY, LOW ENRICHED URANIUM DECONVERSION PROCESS
A novel semi-batch process for deconverting high assay low enriched uranium (HALEU) from its uranium hexafluoride state to uranium dioxide and other chemical states useful as feeds for nuclear fuel in a nuclear reactor is provided. The semi-batch process enables the use of equipment that is small enough, and production rates that are low enough, to meet nuclear criticality safety restraints for HALEU, while enabling the safe, dependable, and economical production of HALEU feed for nuclear fuel at a nominal capacity of up to about 20 MTU (metric tons of uranium metal) per year per deconversion reactor.
A novel semi-batch process for deconverting high assay low enriched uranium (HALEU) from its uranium hexafluoride state to uranium dioxide and other chemical states useful as feeds for nuclear fuel in a nuclear reactor is provided. The semi-batch process enables the use of equipment that is small enough, and production rates that are low enough, to meet nuclear criticality safety restraints for HALEU, while enabling the safe, dependable, and economical production of HALEU feed for nuclear fuel at a nominal capacity of up to about 20 MTU (metric tons of uranium metal) per year per deconversion reactor.
A novel semi-batch process for deconverting high assay low enriched uranium (HALEU) from its uranium hexafluoride state to uranium dioxide and other chemical states useful as feeds for nuclear fuel in a nuclear reactor is provided. The semi-batch process enables the use of equipment that is small enough, and production rates that are low enough, to meet nuclear criticality safety restraints for HALEU, while enabling the safe, dependable, and economical production of HALEU feed for nuclear fuel at a nominal capacity of up to about 20 MTU (metric tons of uranium metal) per year per deconversion reactor.
A nuclear fuel assembly is constructed with fuel assembly components that are wire wrapped and positioned in hexagonal rings within a fuel assembly duct. The fuel assembly components positioned in an outermost ring of the fuel assembly are wire wrapped with a pitch that is shorter than fuel assembly components positioned at an interior ring of the fuel assembly. The shorter pitch at the outer ring of the fuel assembly increases pressure drop of a coolant fluid at the edge and corner subchannels and thereby reduces the temperature gradient across the fuel assembly, which provides a higher output temperature of the nuclear reactor without substantially increasing peak temperature of the fuel cladding.
A vaporizer includes an outer tube configured to receive a flow of heated gas and an inner tube disposed at least partially within the outer tube. The inner tube is spaced apart from the outer tube such that the flow of heated gas is channeled through an annular space therebetween. The vaporizer also includes a crucible disposed at least partially within the inner tube. The crucible is extendable and retractable relative to the inner tube and within the outer tube. The crucible is configured to hold a molten metal such that a surface area of the molten metal exposed to the flow of heated gas is adjustable based on the position of the crucible relative to the inner tube. A heater is configured to vaporize the molten material and the vapor mixes with the flow of heated gas.
G21C 17/028 - Devices or arrangements for monitoring coolant or moderator for monitoring gaseous coolants
B01B 1/00 - BoilingBoiling apparatus for physical or chemical purposes
B01J 8/02 - Chemical or physical processes in general, conducted in the presence of fluids and solid particlesApparatus for such processes with stationary particles, e.g. in fixed beds
B05C 11/11 - Vats or other containers for liquids or other fluent materials
G01N 27/626 - Investigating or analysing materials by the use of electric, electrochemical, or magnetic means by investigating the ionisation of gases, e.g. aerosolsInvestigating or analysing materials by the use of electric, electrochemical, or magnetic means by investigating electric discharges, e.g. emission of cathode using heat to ionise a gas
84.
FUEL-CLADDING CHEMICAL INTERACTION RESISTANT NUCLEAR FUEL ELEMENTS AND METHODS FOR MANUFACTURING THE SAME
This disclosure describes fuel-cladding chemical interaction (FCCI) resistant nuclear fuel elements and their manufacturing techniques. The nuclear fuel elements include two or more layers of different materials (i.e., adjacent barriers are of different base materials) provided on a steel cladding to reduce the effects of FCCI between the cladding and the nuclear material. Depending on the embodiment, a layer may be the structural element (i.e., a layer thick enough to provide more than 50% of the strength of the overall component consisting of the cladding and the barriers) or may be more appropriately described as a liner or coating that is applied in some fashion to a surface of the structural component (e.g., to the cladding, or to a structural form of the fuel).
G21C 3/20 - Details of the construction within the casing with coating on fuel or on inside of casingDetails of the construction within the casing with non-active interlayer between casing and active material
G21C 21/02 - Manufacture of fuel elements or breeder elements contained in non-active casings
G21C 21/14 - Manufacture of fuel elements or breeder elements contained in non-active casings by plating in a fluid
Disclosed herein is a nuclear reactor system comprising a vessel encompassing a reactor core. An external loop line is connected to top and bottom portions of the vessel. A pump can circulate molten fuel through the reactor core and the external loop line. An external neutron reflector encompasses the vessel. External heaters heat the loop line and the reflector. The loop line, the reactor, or both, may be covered with insulation.
A nuclear reactor is designed to couple the load path of control elements with the reactor core, thus reducing opportunity for differential movement between the control elements and the reactor core. A core barrel can be fabricated in a manufacturing facility to include the reactor core, control element supports, and control element drive system. The core barrel can be mounted to a reactor vessel head. Movement, such as through seismic forces, transmits an equal direction and magnitude to the control elements and the reactor core, thus inhibiting the opportunity for differential movement.
G21C 9/00 - Emergency protection arrangements structurally associated with the reactor
G21C 3/33 - Supporting or hanging of elements in the bundleMeans forming part of the bundle for inserting it into, or removing it from, the coreMeans for coupling adjacent bundles
G21C 1/02 - Fast fission reactors, i.e. reactors not using a moderator
G21C 15/247 - Promoting flow of the coolant for liquids for liquid metals
G21C 19/04 - Means for controlling flow of coolant over objects being handledMeans for controlling flow of coolant through channel being serviced
G21C 1/32 - Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel
87.
Fission product getter formed by additive manufacturing
A getter element includes a getter material reactive with a fission product contained within a stream of liquid and/or gas exiting a fuel assembly of a nuclear reactor. At least one transmission pathway passes through the getter element that is sufficiently sized to maintain a flow of the input stream through the getter element at above a selected flow level. At least one transmission pathway includes a reaction surface area sufficient to uptake a pre-identified quantity of the fission product.
G21C 3/17 - Means for storage or immobilisation of gases in fuel elements
G21C 3/18 - Internal spacers or other non-active material within the casing, e.g. compensating for expansion of fuel rods or for compensating excess reactivity
G21C 3/32 - Bundles of parallel pin-, rod-, or tube-shaped fuel elements
88.
MODIFIED LOW POWER, FAST SPECTRUM MOLTEN FUEL REACTOR DESIGNS HAVING IMPROVED NEUTRONICS
A simple nuclear reactor in which most of the reflector material is outside of the reactor vessel is described. The reactor vessel is a cylinder that contains all of the fuel salt and a displacement component, which may be a reflector, in the upper section of the reactor vessel. Other than the displacement component, the reflector elements including a radial reflector and a bottom reflector are located outside the vessel. The salt flows around the outside surface of the displacement component through a downcomer heat exchange duct defined by the exterior of the displacement component and the interior surface of the reactor vessel. This design reduces the overall size of the reactor vessel for a given volume of salt relative to designs with internal radial or bottom reflectors.
G21C 1/03 - Fast fission reactors, i.e. reactors not using a moderator cooled by a coolant not essentially pressurised, e.g. pool-type reactors
G21C 5/02 - Moderator or core structureSelection of materials for use as moderator Details
G21C 7/08 - Control of nuclear reaction by application of neutron-absorbing material, i.e. material with absorption cross-section very much in excess of reflection cross-section by displacement of solid control elements, e.g. control rods
G21C 7/32 - Control of nuclear reaction by varying flow of coolant through the core
G21C 11/06 - Reflecting shields, i.e. for minimising loss of neutrons
G21C 15/04 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from fissile or breeder material
G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel
G21D 7/04 - Arrangements for direct production of electric energy from fusion or fission reactions using thermoelectric elements
G21C 3/24 - Fuel elements with fissile or breeder material in fluid form within a non-active casing
G21C 5/10 - Means for supporting the complete structure
G21C 15/243 - Promoting flow of the coolant for liquids
G21C 15/253 - Promoting flow of the coolant for gases, e.g. blowers
89.
MODIFIED LOW POWER, FAST SPECTRUM MOLTEN FUEL REACTOR DESIGNS HAVING IMPROVED NEUTRONICS
A simple nuclear reactor in which most of the reflector material is outside of the reactor vessel is described. The reactor vessel is a cylinder that contains all of the fuel salt and a displacement component, which may be a reflector, in the upper section of the reactor vessel. Other than the displacement component, the reflector elements including a radial reflector and a bottom reflector are located outside the vessel. The salt flows around the outside surface of the displacement component through a downcomer heat exchange duct defined by the exterior of the displacement component and the interior surface of the reactor vessel. This design reduces the overall size of the reactor vessel for a given volume of salt relative to designs with internal radial or bottom reflectors.
G21C 1/03 - Fast fission reactors, i.e. reactors not using a moderator cooled by a coolant not essentially pressurised, e.g. pool-type reactors
G21C 3/24 - Fuel elements with fissile or breeder material in fluid form within a non-active casing
G21C 5/02 - Moderator or core structureSelection of materials for use as moderator Details
G21C 5/10 - Means for supporting the complete structure
G21C 7/32 - Control of nuclear reaction by varying flow of coolant through the core
G21C 11/06 - Reflecting shields, i.e. for minimising loss of neutrons
G21C 15/04 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from fissile or breeder material
G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel
G21C 15/243 - Promoting flow of the coolant for liquids
G21C 15/253 - Promoting flow of the coolant for gases, e.g. blowers
G21D 7/04 - Arrangements for direct production of electric energy from fusion or fission reactions using thermoelectric elements
90.
Fuel handling system, layout, and process for nuclear reactor
A method of handling spent nuclear fuel assemblies immerses the spent nuclear fuel assemblies in water in a relatively short time period when compared to traditional methods. A spent nuclear fuel assembly is removed from a nuclear reactor core, inserted into a sodium removal machine having a receiver, a cleaning vessel, and an elevator. A cleaning fluid is applied to the cleaning vessel and fuel assembly, and the fuel assembly is flushed with water while in the cleaning vessel. The cleaning vessel is at least partially submerged in the spent fuel pool during cleaning to provide passive heat removal. The cleaning vessel is lowered by an elevator into the spent fuel pool. The fuel assembly may then be loaded into a rack and/or a cask for long-term storage.
G21C 19/19 - Reactor parts specifically adapted to facilitate handling, e.g. to facilitate charging or discharging of fuel elements
G21C 19/08 - Means for heating fuel elements before introduction into the coreMeans for heating or cooling fuel elements after removal from the core
G21C 19/32 - Apparatus for removing radioactive objects or materials from the reactor discharge area, e.g. to a storage placeApparatus for handling radioactive objects or materials within a storage place or removing them therefrom
91.
Mesophase pitch for carbon fiber production using supercritical carbon dioxide
Embodiments of methods for improving mesophase pitch for carbon fiber production using supercritical carbon dioxide are described. The methods improve the relative amount and quality of mesophase pitch in feedstocks, such as coal tar, already having at least some mesophase pitch. One particular method includes performing a sCO2/toluene extraction on the coal tar to obtain a toluene insoluble fraction of the coal tar; mixing the toluene insoluble fraction with sCO2 to obtain a sCO2/toluene insoluble fraction mixture; and extruding the sCO2/toluene insoluble fraction mixture, thereby separating the sCO2 from the toluene insoluble fraction to obtain fibers of mesophase pitch.
C10C 3/08 - Working-up pitch, asphalt, bitumen by selective extraction
D01F 9/15 - Carbon filamentsApparatus specially adapted for the manufacture thereof by decomposition of organic filaments from pitch or distillation residues from coal pitch
C01B 32/05 - Preparation or purification of carbon not covered by groups , , ,
92.
THORIUM PEROXIDE-BASED GENERATOR FOR AC-225 GENERATION
The actinium generator described herein is based on peroxide precipitation of thorium from its daughter products radium and actinium. In this system, the "actinium generator" is a quantity of solid thorium peroxide stored under a cover solution. The thorium peroxide is stored, as a suspension to allow for the buildup of the decay products radium and actinium in the suspension. The suspension is then treated with a peroxide solution and the solid and liquid phases are separated. The thorium remains in the solid peroxide form while the soluble actinium and radium are removed with the liquid phase in a rinsing step. After rinsing, an amount of the rinsing solution is retained with the thorium peroxide solid as a fresh cover solution to form another suspension for storage. This new suspension is then stored to allow actinium and radium to again build up in the suspension for a subsequent separation cycle.
G21G 1/00 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes
C22B 60/02 - Obtaining thorium, uranium or other actinides
Radiation cured composite materials are greatly improved by enhancing the fiber to matrix bond by prewetting the fibers with an interface resin that has a curing agent mixed in with the interface resin. Furthermore, radiation curing the composite material at or near an expected operating temperature of the composite material improves the mechanical properties of the material by reducing thermally induced strains and stresses caused by thermally curing a material and subsequently cooling the material. Adding an interface resin with a curing agent to the fibers allows relatively thick parts, a must faster curing process, a wide variety of inexpensive and easily workable molding materials, the ability to maintain tight tolerances and reduce or eliminate springback, and a radiation cured material that approaches or exceeds the material characteristics of thermally cured composite materials.
In one aspect, the technology relates to a method of producing Ac, the method including preparing a phosphate-modified titania material to produce an ion-exchange material, contacting a solution including 229Th with the ion-exchange material to produce a Th-loaded titania material, eluting the Th-loaded titania material with a wash solution to produce an eluted solution containing eluted compounds including 225Ac, concentrating the eluted solution to generate eluted compounds including the 225Ac, and separating the 225Ac from the eluted compounds.
C22B 60/02 - Obtaining thorium, uranium or other actinides
G21G 4/08 - Radioactive sources other than neutron sources characterised by constructional features specially adapted for medical applications
G21G 1/00 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes
95.
ZAMAK STABILIZATION OF SPENT SODIUM-COOLED REACTOR FUEL ASSEMBLIES
Methods and systems for stabilizing spent fuel assemblies from sodium-cooled nuclear reactors using Zamak are described herein. It has been determined that there is a synergism between Zamak and sodium that allows Zamak to form thermally-conductive interface with the sodium-wetted surfaces of the fuel assemblies. In the method, one or more spent fuel assemblies are removed from the sodium coolant pool and placed in a protective sheath. The remaining volume of the sheath is then filled with liquid Zamak. To a certain extent Zamak will dissolve and alloy with sodium remaining on the fuel assemblies. Excess sodium that remains undissolved is displaced from the sheath by the Zamak fill. The Zamak is then cooled until solid and the sheath sealed. The resulting Zamak-stabilized spent fuel assembly is calculated to have sufficient internal thermal conductivity to allow it to be stored and transported without the need for liquid cooling.
G21C 19/32 - Apparatus for removing radioactive objects or materials from the reactor discharge area, e.g. to a storage placeApparatus for handling radioactive objects or materials within a storage place or removing them therefrom
Methods and systems for stabilizing spent fuel assemblies from sodium-cooled nuclear reactors using Zamak are described herein. It has been determined that there is a synergism between Zamak and sodium that allows Zamak to form thermally-conductive interface with the sodium-wetted surfaces of the fuel assemblies. In the method, one or more spent fuel assemblies are removed from the sodium coolant pool and placed in a protective sheath. The remaining volume of the sheath is then filled with liquid Zamak. To a certain extent Zamak will dissolve and alloy with sodium remaining on the fuel assemblies. Excess sodium that remains undissolved is displaced from the sheath by the Zamak fill. The Zamak is then cooled until solid and the sheath sealed. The resulting Zamak-stabilized spent fuel assembly is calculated to have sufficient internal thermal conductivity to allow it to be stored and transported without the need for liquid cooling.
C22C 18/04 - Alloys based on zinc with aluminium as the next major constituent
G21F 1/08 - MetalsAlloysCermets, i.e. sintered mixtures of ceramics and metals
G21C 19/32 - Apparatus for removing radioactive objects or materials from the reactor discharge area, e.g. to a storage placeApparatus for handling radioactive objects or materials within a storage place or removing them therefrom
Methods and systems for stabilizing spent fuel assemblies from sodium-cooled nuclear reactors using Zamak are described herein. It has been determined that there is a synergism between Zamak and sodium that allows Zamak to form thermally-conductive interface with the sodium-wetted surfaces of the fuel assemblies. In the method, one or more spent fuel assemblies are removed from the sodium coolant pool and placed in a protective sheath. The remaining volume of the sheath is then filled with liquid Zamak. To a certain extent Zamak will dissolve and alloy with sodium remaining on the fuel assemblies. Excess sodium that remains undissolved is displaced from the sheath by the Zamak fill. The Zamak is then cooled until solid and the sheath sealed. The resulting Zamak-stabilized spent fuel assembly is calculated to have sufficient internal thermal conductivity to allow it to be stored and transported without the need for liquid cooling.
G21F 5/10 - Heat-removal systems, e.g. using circulating fluid or cooling fins
G21C 19/32 - Apparatus for removing radioactive objects or materials from the reactor discharge area, e.g. to a storage placeApparatus for handling radioactive objects or materials within a storage place or removing them therefrom
A nuclear reactor includes a heat exchanger that transfers thermal energy from a primary reactor coolant to a secondary coolant. The heat exchanger is formed with a hot flow channel, a cold flow channel, and a porous layer between the hot flow channel and the cold flow channel. The porous layer may be thermally insulative to reduce the efficiency of thermal energy transfer from the hot flow channel to the cold flow channel. The porous layer may have a control gas passed therethrough that can be tailored to control the thermal energy transfer through the porous layer. The control gas can be tested for leakage within the heat exchanger. The control gas may also be used to sequester fission or activation products.
A nuclear reactor includes a heat exchanger that transfers thermal energy from a primary reactor coolant to a secondary coolant. The heat exchanger is formed with a hot flow channel, a cold flow channel, and a porous layer between the hot flow channel and the cold flow channel. The porous layer may be thermally insulative to reduce the efficiency of thermal energy transfer from the hot flow channel to the cold flow channel. The porous layer may have a control gas passed therethrough that can be tailored to control the thermal energy transfer through the porous layer. The control gas can be tested for leakage within the heat exchanger. The control gas may also be used to sequester fission or activation products.
A nuclear reactor includes a heat exchanger that transfers thermal energy from a primary reactor coolant to a secondary coolant. The heat exchanger is formed with a hot flow channel, a cold flow channel, and a porous layer between the hot flow channel and the cold flow channel. The porous layer may be thermally insulative to reduce the efficiency of thermal energy transfer from the hot flow channel to the cold flow channel. The porous layer may have a control gas passed therethrough that can be tailored to control the thermal energy transfer through the porous layer. The control gas can be tested for leakage within the heat exchanger. The control gas may also be used to sequester fission or activation products.
