A process of incorporating technetium into an electroless deposit, forming an alloy that is extremely resistant to corrosion and reduces the mobility of technetium on a geologic time scale is disclosed and claimed. The process includes providing a liquid containing technetium, such as an aqueous waste stream generated during the used nuclear fuel reprocessing activities. The technetium is collected and concentrated, and provided into an electroless deposition bath. A substrate, such as suitably prepared zero valent iron or stainless steel, is introduced into the bath to initiate autocatalytic electroless deposition of the technetium onto the substrate due to the difference in electrochemical potential between the plating bath and the metals in solution. This causes a layer of technetium metal to form on the substrate. The electroless deposition is continued until virtually all of the technetium has been removed from the bath, and then continues to build a layer of technetium-free material on the outermost surface of the substrate. One or more additional deposition steps may be performed to armor the plated substrate against leaching/corrosion in a nuclear waste disposal facility.
C23C 18/16 - Chemical coating by decomposition of either liquid compounds or solutions of the coating forming compounds, without leaving reaction products of surface material in the coatingContact plating by reduction or substitution, i.e. electroless plating
C23C 18/34 - Coating with one of iron, cobalt or nickelCoating with mixtures of phosphorus or boron with one of these metals using reducing agents
This patent discloses and claims a system for providing in-plane stabilization to the tubes of a heat exchanger. The system increases the friction between the tubes and the existing anti-vibration bars by pressing the tube bundle together in the out-of-plane direction. The invention involves a structure that develops forces by reacting either off the inner diameter of the tube bundle wrapper, or off an assembled stressor hoop, which is self-supporting on top of the tube bundle, or involves purposely deforming connecting hoops to bias the anti-vibration bars in a way that the tube U-bend is no longer planar. It imposes a controlled clamping and/or deforming force on the tube bundle, either by direct contact with the outer tubes or indirectly via the existing anti -vibration bar structure.
The inventive tool is designed to ensure proper plant operation position of the core shroud head bolt assemblies of a boiling water nuclear reactor. The tool exerts an upward pulling force on stuck retainers to return them to the locking position around the nut. The tool includes a frame, an engagement member, and a collar. The engagement member is lowered toward and end of the bolt until it comes into contact therewith. Further movement of the engagement member causes the frame and collar to move upward, lifting the retainer and extending the spring.
The invention provides at least two electrochemical sensors positioned in a nuclear reactor or in a system adjacent to the nuclear reactor, wherein at least one of the at least two electrochemical sensors has a heated zirconium electrode, and the at least two electrochemical sensors produce voltages proportional to an electrochemical corrosion potential for a surface that each of the at least two electrochemical sensors are installed upon. The invention also provides an arrangement configured to accept the voltages produced by the at least two electrochemical sensors, wherein the arrangement is configured to determine an electrochemical corrosion potential of a zirconium fuel rod in the nuclear reactor based upon the voltages of the at least two electrochemical sensors.
A method for retrofitting a boiling water reactor is provided. The method includes removing a mixing chamber from a slip joint defined by a diffuser and the mixing chamber, the mixing chamber having an inner surface and a bottom edge directing flow to the diffuser such that a recirculation zone at an entrance to the slip joint creates a diverging effective path for the leakage flow entering the slip joint. The method also includes providing a new inner surface and new bottom edge, the new inner surface and the new bottom edge being reshaped to decrease the size of the recirculation zone. A jet pump is also provided.
The present invention provides a method and an apparatus for calibrating a first self-powered neutron detector for long term use in a nuclear reactor core with a second self-powered neutron detector, where the emitter material of the second self-powered neutron detector has a neutron absorption cross-section that is greater than the neutron absorption cross-section of the first emitter material for the first self-powered neutron detector.
The present invention provides a method and an apparatus for calibrating a first self-powered neutron detector for long term use in a nuclear reactor core with a second self-powered neutron detector, where the emitter material of the second self-powered neutron detector has a neutron absorption cross-section that is greater than the neutron absorption cross-section of the first emitter material for the first self-powered neutron detector.
A method for removing corrosion products from a system, the method including: adjusting the system temperature to between 115° F. to 212° F.; injecting a cleaning dissolution solvent into the system; injecting a gas into the system after the system is filled with the cleaning dissolution solvent; the gas mixing with the solvent in the system; draining the solvent from the system after a predetermined period of time of dissolution; injecting a passivation composition into the system; injecting a gas into the system, the gas mixing the passivation composition; draining the system of the composition after a predetermined period of time of passivation; rinsing the system with a low volume solution; and rinsing the system at with a full volume solution.
A dry conversion reactor for converting uranium hexafluoride to uranium dioxide, the dry conversion reactor including a gas-phase reaction segment and a fluidized bed segment, wherein at least one of the gas-phase reaction segment and the fluidized bed segment is a replaceable segment. A method for operating a dry conversion reactor utilizing a uranium hexafluoride to uranium dioxide conversion process, the method including replacing at least one conversion reactor segment.
A method for retrofitting a boiling water reactor slip joint of a jet pump to reduce vibrations is provided. The method includes removing a mixing chamber from an existing slip joint defined by a diffuser and the mixing chamber, the existing slip joint defining an existing annular gap, and providing a new slip joint defining a new annular gap, the new annular gap being reshaped to permit reduced vibration. A jet pump and a method of operating a jet pump are also provided.
A method to perform an analysis of two types of CRUD on a nuclear fuel rod, including providing a nuclear fuel rod with first and second layers of CRUD on an exterior of the fuel rod; brushing the first layer of CRUD from a selected area on the fuel rod with a CRUD tool having a brushing device; a force applied to the brushing device to remove the first layer of CRUD; collecting the first layer of CRUD from the brushing device, scraping the second layer of CRUD from the fuel rod in the selected area with the tool, wherein the tool has a scraping device and a second force is applied to the tool for scraping, collecting the second layer of CRUD from the scraping device, and analyzing the first and second layers of CRUD separately with a scanning electron microscope.
A leak detecting apparatus for detecting flaws in a surface submerged in liquid is provided that includes a vacuum head for contacting the surface and forming a vacuum chamber filled with liquid between the surface and the vacuum head, the vacuum head including a multi-surface elastomer seal for contacting the surface; and a vacuum generator for pulling gas through a flaw in the surface into the vacuum head and to the gas separation chamber so as to displace the liquid from the gas separation chamber. A method for detecting flaws in a surface submerged in liquid is also provided.
A motor operated valve diagnostic data acquisition system including: a motor operated valve assembly; a transmitter, the transmitter being permanently connected in a housing of a motor operated valve assembly; a receiver, the transmitter transmitting signals to the receiver via motor operated valve cables, the cables being existing motor operated valve cables; and electronic circuitry connected to the receiver, the electronic circuitry processing the signals received by the receiver.
F16K 37/00 - Special means in or on valves or other cut-off apparatus for indicating or recording operation thereof, or for enabling an alarm to be given
G01L 5/00 - Apparatus for, or methods of, measuring force, work, mechanical power, or torque, specially adapted for specific purposes
G01L 5/12 - Apparatus for, or methods of, measuring force, work, mechanical power, or torque, specially adapted for specific purposes for measuring axial thrust in a rotary shaft, e.g. of propulsion plants
14.
PIPE ASSEMBLY WITH SCOOP FOR DIRECTING FLUID INTO A STANDPIPE
A piping assembly for directing fluid and mitigating acoustic and vortex coupled resonance is provided that includes a main pipe delivering fluid in a first direction; a standpipe coupled to the main pipe at an intersection; and a scoop positioned at the intersection directing the fluid towards the standpipe. A scooping insert and a method for disrupting vortex shedding in a piping assembly are also provided.
A method to determine corrective actions of a nuclear steam generator, having the steps of modeling of steam generator tube and deposit heat transfer characteristics by analytically deriving specific deposit characteristics and descriptive model parameters, wherein the modeling uses historical thermodynamic data for an operating plant under evaluation, identifying a set of one of preventive and corrective maintenance alternatives to accomplish steam generator deposit objectives, determining through the modeling a power production impact of each of the set of one of preventative and corrective maintenance alternatives to determine an economic cost for each of the set of preventative and corrective maintenance alternatives, and initiating a maintenance alternative with a lowest economic cost as compared to the maintenance evaluation alternatives with higher economic costs.
A nickel, chromium, iron alloy and method for use in producing weld deposits and weldments formed therefrom. The alloy comprises, in weight percent, about 28.5 to 31.0% chromium; about 0 to 16% iron; less than about 1.0% manganese; about 2.1 to 4.0% niobium plus tantalum; 1.0 to 6.5% molybdenum; less than 0.50% silicon; 0.01 to 0.35% titanium; 0 to 0.25% aluminum; less than 1.0% copper; less than 1.0% tungsten; less than 0.5% cobalt; less than about 0.10% zirconium; less than about 0.01% sulfur; less than 0.01% boron; less than 0.03% carbon; less than about 0.02% phosphorous; 0.002 to 0.015% magnesium plus calcium; and balance nickel and incidental impurities. The method includes the steps of forming a welding electrode from the above alloy composition and melting the electrode to form a weld deposit. A preferred weldment may be in the form of a tubesheet of a nuclear reactor.
B32B 15/04 - Layered products essentially comprising metal comprising metal as the main or only constituent of a layer, next to another layer of a specific substance
C22C 19/05 - Alloys based on nickel or cobalt based on nickel with chromium
B23K 35/22 - Rods, electrodes, materials, or media, for use in soldering, welding, or cutting characterised by the composition or nature of the material
17.
Boiling water reactor nuclear power plant with alcohol injection
A nuclear power plant is provided including a BWR, a reactor cooling system cooling the BWR, an HWC hydrogen injection system connected to the reactor cooling system and an alcohol injection system connected to the reactor cooling system. Methods for providing methanol and hydrogen are also provided.
A method to determine corrective actions of a nuclear steam generator, having the steps of modeling of steam generator tube and deposit heat transfer characteristics by analytically deriving specific deposit characteristics and descriptive model parameters, wherein the modeling uses historical thermodynamic data for an operating plant under evaluation, identifying a set of one of preventive and corrective maintenance alternatives to accomplish steam generator deposit objectives, determining through the modeling a power production impact of each of the set of one of preventative and corrective maintenance alternatives to determine an economic cost for each of the set of preventative and corrective maintenance alternatives, and initiating a maintenance alternative with a lowest economic cost as compared to the maintenance evaluation alternatives with higher economic costs.
A nuclear power plant (10) is provided including a BWR (20), a reactor cooling system cooling the BWR (20), an HWC hydrogen injection system (60) connected to the reactor cooling system and an alcohol injection system (100) connected to the reactor cooling system. Methods for providing methanol and hydrogen are also provided.
An improved manipulator system having a remotely actuated tube walker assembly for reciprocally moving the manipulator along a tube bundle for inspection and repair of the tubes of the tube bundle wherein the tube walker has a movable inner section and an outboard section with both sections having groups of hydraulically actuated locking fingers which are selectively engageable with the tubes of the tube bundle so that when the fingers of the outboard section are locked the fingers of the inner section are disengaged and the inner portion can be moved axially to an aligned position with new tube locations.
Methods for removing nuclear waste from a component are provided. In one embodiment, the method includes the steps of supplying oxalic acid to the nuclear waste on the component to form an oxalic acid/waste solution, feeding the oxalic acid/waste solution from the component, feeding an oxidant to the oxalic acid/waste solution to form carbon-dioxide, water and a precipitate, and separating the precipitate from the water. A system for removing nuclear waste is also provided.
A62D 3/30 - Processes for making harmful chemical substances harmless, or less harmful, by effecting a chemical change in the substances by reacting with chemical agents
A fuel channel fastener having a body with a hole, a bolt inserted into the hole, at least two anti-rotation supports attached to the body, a spring configured to attach to the body and to extend away from the body down the sides of a fuel assembly fuel channel, and at least two spring anti-rotation pads connected to the body.
The invention provides at least two electrochemical sensors positioned in a nuclear reactor or in a system adjacent to the nuclear reactor, wherein at least one of the at least two electrochemical sensors has a heated zirconium electrode, and the at least two electrochemical sensors produce voltages proportional to an electrochemical corrosion potential for a surface that each of the at least two electrochemical sensors are installed upon. The invention also provides an arrangement configured to accept the voltages produced by the at least two electrochemical sensors, wherein the arrangement is configured to determine an electrochemical corrosion potential of a zirconium fuel rod in the nuclear reactor based upon the voltages of the at least two electrochemical sensors.
A method to assess light water reactor fuel integrity is presented having the steps of granting access in a nuclear reactor fuel pool to at least one of a discharged fuel rod and a nuclear fuel assembly, calculating an operating flux for the fuel rod, measuring a thickness of CRUD on the fuel rod, measuring a thickness of oxide on the fuel rod, calculating a maximized flux for the at least one fuel rod for a position of the one fuel rod in a nuclear reactor, calculating a maximized deposit for the fuel rod, calculating a maximized oxide thickness for the fuel rod, calculating a fuel condition index of the fuel rod, comparing the fuel condition index to an index constant, and removing the fuel rod from operation when the fuel condition index is greater than the index constant.
A method to characterize the power transfer of a nuclear component is provided including the steps of obtaining a sample of a deposit layer on a side of a nuclear component, obtaining a scanning electron microscope image of an outside surface of the sample, obtaining a scanning electron microscope image of an inside surface of the sample, analyzing the scanning electron microscope images of the outside and inside surfaces of the sample for a presence of capillaries and steam chimneys, and calculating the power transfer of the component based on a number of steam chimneys in the deposit layer.
G01N 23/225 - Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups , or by measuring secondary emission from the material using electron or ion microprobes
A method to determine deposits in a steam generator having the steps of creating a calibration standard having at least two rings of deposit material, subjecting the calibration standard to an eddy current signal, wherein an amplitude of the signal reflected from the calibration standard is used to obtain a polynomial equation fit of the reflected eddy current signals to actual thickness of the rings, obtaining a steam generator with tubes, initiating an eddy current signal into the tubes of the steam generator, detecting and recording reflections of the eddy current signal initiated into the tubes of the steam generator, and determining a thickness of the deposits in the steam generator from the recorded reflections of the eddy current signal and the polynomial equation.
G01B 7/06 - Measuring arrangements characterised by the use of electric or magnetic techniques for measuring length, width, or thickness for measuring thickness
27.
Apparatus and method for aiding in the removal of enriched uranium from soils
A method and apparatus for removing uranium (IV) and uranium (VI) from sands and soils. The method and device assays a volume of soil and determines a presence and position of uranium enrichment. A concentration of uranium enrichment is determined, and compared to a threshold concentration. The volume of soil is processed and transported to a coarse screen, deck screen and classified in a classifier. In the classifier, the volume of soil is separated into a washed and fines fraction, assayed and transported to a uranium recovery facility, then transported to a silicon removal process and subsequent ammonia and metals removal processes.
A method to characterize the power transfer of a nuclear component is provided including the steps of obtaining a sample of a deposit layer on a side of a nuclear component, obtaining a scanning electron microscope image of an outside surface of the sample, obtaining a scanning electron microscope image of an inside surface of the sample, analyzing the scanning electron microscope images of the outside and inside surfaces of the sample for a presence of capillaries and steam chimneys, and calculating the power transfer of the component based on a number of steam chimneys in the deposit layer.
G01N 23/225 - Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups , or by measuring secondary emission from the material using electron or ion microprobes
A nickel, chromium, iron alloy and method for use in producing weld deposits and weldments formed therefrom. The alloy comprises, in weight percent, about 28.5 to 31.0% chromium; about 0 to 16% iron, preferably 7.0 to 10.5% iron, less than about 1.0% manganese, preferably 0.05 to 0.35% manganese; about 2.1 to 4.0% niobium plus tantalum, preferably 2.1 to 3.5% niobium plus tantalum, and more preferably 2.2 to 2.8% niobium plus tantalum; 0 to 7.0% molybdenum, preferably 1.0 to 6.5%, and more preferably 3.0 to 5.0% molybdenum; less than 0.50% silicon, preferably 0.05 to 0.30% silicon; 0.01 to 0.35% titanium; 0 to 0.25% aluminum; less than 1.0% copper; less than 1.0% tungsten; less than 0.5% cobalt; less than about 0.10% zirconium; less than about 0.01% sulfur; less than 0.01% boron, preferably less than 0.0015% boron, and more preferably less than 0.001% boron; less than 0.03% carbon; less than about 0.02% phosphorous; 0.002 to 0.015% magnesium plus calcium; and balance nickel and incidental impurities. The method includes the steps of forming a welding electrode from the above alloy composition and melting the electrode to form a weld deposit. A preferred weldment made from the alloy and method includes an alloy substrate in the form of a tubesheet of a nuclear reactor.
A method to analyze crystals in a deposit on a surface of a nuclear generating station heating surface, wherein the method extracts a sample of material from the surface of the nuclear generating station heating surface and also includes conducting at least one of a high resolution scanning electron microscope/energy dispersive X-ray spectrometry of the sample and a scanning transmission electron microscope/selected area electron diffraction/spot and elemental mapping analysis of the sample.
A method to determine deposits in a steam generator having the steps of creating a calibration standard having at least two rings of deposit material, subjecting the calibration standard to an eddy current signal, wherein an amplitude of the signal reflected from the calibration standard is used to obtain a polynomial equation fit of the reflected eddy current signals to actual thickness of the rings, obtaining a steam generator with tubes, initiating an eddy current signal into the tubes of the steam generator, detecting and recording reflections of the eddy current signal initiated into the tubes of the steam generator, and determining a thickness of the deposits in the steam generator from the recorded reflections of the eddy current signal and the polynomial equation.
A method to determine corrective actions of a nuclear steam generator, having the steps of modeling of steam generator tube and deposit heat transfer characteristics by analytically deriving specific deposit characteristics and descriptive model parameters, wherein the modeling uses historical thermodynamic data for an operating plant under evaluation, identifying a set of one of preventive and corrective maintenance alternatives to accomplish steam generator deposit objectives, determining through the modeling a power production impact of each of the set of one of preventative and corrective maintenance alternatives to determine an economic cost for each of the set of preventative and corrective maintenance alternatives, and initiating a maintenance alternative with a lowest economic cost as compared to the maintenance evaluation alternatives with higher economic costs.
G05B 19/418 - Total factory control, i.e. centrally controlling a plurality of machines, e.g. direct or distributed numerical control [DNC], flexible manufacturing systems [FMS], integrated manufacturing systems [IMS] or computer integrated manufacturing [CIM]
A method for determining the inner surface and an outer surface of a sample flake through the use of digital image photography wherein digital images of a sample flake first surface and second surface are compared to a known inner surface and known outer surface of a reference flake.
A method to load a nuclear fuel rod, comprising, providing nuclear fuel pellets in a fuel plate transfer unit; transferring the nuclear fuel pellets from the fuel plate transfer unit to a fuel pellet column through the use of a belt, indexing the nuclear fuel pellets in the fuel pellet column to a nuclear fuel pellet loading machine; and pushing the fuel pellet column into a fuel rod cladding.
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
A method to load a nuclear fuel rod, comprising, providing nuclear fuel pellets in a bottomless tray; transferring the nuclear fuel pellets from the bottomless tray to a fuel pellet column through the use of a belt, indexing the nuclear fuel pellets in the fuel pellet column to a nuclear fuel pellet loading machine; and pushing the fuel pellet column into a fuel rod cladding.
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
An apparatus and method to remove uranium from a body of material wherein the method includes the steps of depositing the body of solid material in an ultrasonic extractor and depositing an amount of acid in the ultrasonic extractor. The method also provides for the steps of heating a jacket of the ultrasonic extractor, transporting the body of solid material in the ultrasonic extractor and the amount of acid such that the body of solid material and the acid contact each other inside the heated ultrasonic extractor while the ultrasonic extractor provides ultrasonic energy to both the body of solid material and the amount of acid, wherein the amount of acid strips uranium from the body of solid material. The method further provides for collecting the amount of acid and the body of solid material in the ultrasonic extractor in different positions, transporting the amount of acid with the stripped uranium to an extraction mixer settler, and settling uranium product from the extraction mixer settler.
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