Abstract: In an industrial installation equipped with a sensor, a method is provided for checking the sensor and providing a measurement sequence associated with the sensor. Monitoring occurs for a thermodynamic parameter of a monitored fluid circulating in the industrial installation, by sensing the fluid using the sensor while the industrial installation is operating and by collecting and processing an electric signal emitted by the sensor using the measurement sequence. A perturbation is produced of fluid origin of the thermodynamic parameter while the installation is operating and the fluid circulates in the installation in order to act on a sensitive element of the sensor and cause a modification of the electric signal emitted by the sensor. The method performs collecting and analyzing of the modified signal from the sensor using the measurement sequence.

Abstract: A device for handling a guide tube assembly. The assembly has an upper tube and a lower tube in each of which there are fixed horizontal guide plates arranged such that they are spaced apart in an axial direction. The handling device itself includes a gripper having a tubular body and two opposing arms located at one end of the tubular body and moveable between a retracted position and a deployed position, the arms bearing against a lower surface of a contacted guide plate when in the deployed position. A control member is located at an opposite end of the tubular body for controlling the arms. The length of the tubular body of the gripper selected to be: a) greater than the distance between an upper end of the upper tube and a first guide plate of the lower tube; and b) less than the distance between the upper end of the upper tube and a third guide plate of the second tube.

Abstract: A method of determining a threshold value of a nuclear reactor operating parameter having the steps of simulating at least a transient operational occurrence of the nuclear reactor, calculating the value reached by a physical quantity during the transient operational occurrence in at least a cladding of a fuel rod of the reactor and establishing, as a limit value of an operational parameter of the reactor, the value of the operational parameter when the value calculated the step of calculating the value reached by a physical quantity during the transient operational occurrence which corresponds to a value of the physical quantity which characterizes a failure of the cladding.

Abstract: A device and method to move a fuel assembly of a nuclear reactor, wherein the device eliminates the potential for dropping the fuel assembly due to stress corrosion cracking of the upper guide thimble sleeves that attach the top nozzle to guide thimbles.

Abstract: A drum closure which establishes a connection between the body of a drum and lid, wherein the connection will prevent separation of the lid from the body under severe loading conditions such as dropping the drum from a height of thirty (30) feet.

Abstract: A control rod drive contains a drive housing, in which a control rod carrying element is moveable between a basic position and an end position. The control rod carrying element is guided over a portion in a throttle bush. Formed across the throttle bush is a free flow cross section for a pressure fluid which varies in dependence on a position of the control rod carrying element. During an emergency shutdown, in which the control rod carrying element is moved hydraulically via a pressure fluid, the flow resistance for the pressure fluid is reduced, when the control rod carrying element reaches the end position. Therefore, a braking of the control rod carrying element takes place before it reaches the end position, thus reducing mechanical loads during braking. A flow resistance change takes place via a variable outside diameter of the control rod carrying element and/or by a bypass orifice.

Abstract: Nuclear metal or alloy particle preparation method and device; said device comprising: means to prepare a fluid mass of the metal or alloy by melting; means to put the fluid mass of the metal or metal alloy in the form of a molten metal or alloy stream; means to impart a centrifugal rotation movement to a quenching fluid and carry out the atomisation of the molten metal or metal alloy stream particles and the quick quenching of the particles; means to melt the nuclear metal or metal alloy in an inert gas atmosphere, and means to surround the molten metal or metal alloy stream by a neutral gas envelope until quenching.

Abstract: The steam generator includes a feedwater delivery device at an upper axial end of an annular space between an outer shell and a bundle wrapper of the steam generator. The feedwater, delivery device includes a header (11) having a shell of toroidal general shape, at least one feed pipe for the header and feedwater distribution means (14) in the upper part of the annular space of the steam generator. The feedwater distribution means (14) include a plurality of openings (17) that pass through the upper part of the shell of the toroidal header (11) and are distributed circumferentially on the wall of the header and a plurality of tubes (15) fixed via a lower end part at an opening (17) of the header (11). Each of the straight tubes (15) includes an upper end part through which openings (20), such as water passage slots, pass and a cap (16) fixed to the upper end of the tube (15) surrounding, with a radial clearance, the upper end part of the tube (15).

Abstract: A system is provided for delivering tooling, suction pickup, and alignment platforms to a stay dome region of a stem generator or the tube-free lane regions adjacent to the stay dome region and includes a suction device positioned in the stay dome region of a steam generator capable of rotating 360° about a vertical axis which also acts as a platform for mounting inter-bundle tooling and alignment devices used to position the suction device radially at any given tube location around the stay dome region to align it with an annulus based spray nozzle tube cleaning system.

Abstract: In a nuclear reactor, cooled by pressurized water, an apparatus is employed for replacing a section of a primary circuit primary pipe that interconnects first and second components of a primary circuit. The apparatus cuts out and removes a section of the primary pipe. Supports are provided for the end parts of a new replacement section in a position held by the removed pipe. The apparatus bevel welds the end parts of the new replacement section of the primary pipe to confronting corresponding ends of the remaining primary pipe. Means are provided to work within the pipe along the entire length of the interior surface of the primary pipe and is introduced through one of first and second components of the primary circuit to a location within the welded replacement pipe section.

Abstract: A lifting support for a boiling water reactor nuclear fuel assembly comprising a grappling head configured to allow attachment to a lifting device, a body with an upper end and a lower end, the upper end connected to the grappling head, the body configured to be inserted into a water channel of a boiling water reactor nuclear fuel assembly, and an end connected to the lower end, the end configured to be accepted by the nuclear fuel assembly.

Abstract: Marking of the cuts which have to be made, cutting of the section at two ends, removal of the section which has to be replaced, bevelling of the joint ends of the parts remaining after the section has been cut out from the pipe, adjustment of a new or replacement section for length and bevelling of its joining ends and positioning and narrow bevel welding of the ends joining the replacement section to the ends of the remaining parts of the pipe are performed outside the pipe. Within the pipe operations of machining and inspecting an internal part of the joining ends welded together are performed by remote control in a programmed way by introducing and positioning means for working within the pipe from a component of the primary circuit.

Abstract: The invention provides a method and tool to perform an analysis of CRUD on a nuclear fuel rod. The method recites providing a nuclear fuel rod with a layer of CRUD on an exterior of the fuel rod, scraping the CRUD from the fuel rod with a CRUD scraping tool and collecting CRUD flakes from the CRUD scraping tool. The method also provides for sorting the CRUD flakes into particle fractions, and analyzing the CRUD with a scanning electron microscope, wherein the scraping tool has a blade with a rigidity that is matched to an anticipated CRUD deposit shear strength.

Abstract: A connector interconnect two coaxial cables each having a central conductor surrounded by an outer conductor. A respective connector piece is assigned to each coaxial cable. The invention ensures the easy and rapid use of the connector, while reliably guaranteeing the electrical contact between the interconnected coaxial cables. To achieve this, each connector piece has one connection region that is electrically connected to the respective central conductor of the corresponding coaxial cable, a connection head of the second connector piece being supported on the connection region of said piece by means of a spring element, in order to make electrical contact with the connection region of the first connector piece.

Abstract: In a method for catalytic oxidation of a gas, the outlay required is kept at a low level for reliably treating even relatively large quantities of gas and/or high concentrations of the gas fraction that needs to be treated. For this purpose, a gas stream including the gas being treated is circulated through a reaction zone and a return-flow zone, which is in communication with the inlet and outlet sides of the reaction zone. The circulation of the gas stream can be maintained effectively in the form of a passive system if the gas stream which includes the gas to be treated is guided in the upward direction in the reaction zone and the upward flow is assisted by convection resulting from the heat released during the oxidation reaction. Furthermore, the gas stream is preferably cooled locally in the return-flow zone by spray cooling.

Abstract: A fuel element for a pressurized water reactor is described. The fuel element contains a laterally open skeleton having control-rod guide tubes each with a first end and a second end, spacers fastened to the control-rod guide tubes, a fuel element head disposed at the first end of the control-rod guide tubes, and a fuel element foot disposed at the second end of the control-rod guide tubes. Gastight cladding tubes are inserted into the skeleton and each is filled with a column of fuel pellets. At least some of the gastight cladding tubes have a multilayer wall. The multilayer wall is formed of a mechanically stable matrix containing a first zirconium alloy disposed in a middle of the multiplayer wall; and a thinner protective layer of a second zirconium alloy alloyed to a lesser extent than the first zirconium alloy. The thinner protective layer is bound metallurgically to the matrix and is disposed on an inside of the matrix facing the fuel pellets.

Abstract: Marking of the cuts which have to be made, cutting of the section at two ends, removal of the section which has to be replaced, bevelling of the joint ends of the parts remaining after the section has been cut out from the pipe, adjustment of a new or replacement section for length and bevelling of its joining ends and positioning and narrow bevel welding of the ends joining the replacement section to the ends of the remaining parts of the pipe are performed outside the pipe. Within the pipe operations of machining and inspecting an internal part of the joining ends welded together are performed by remote control in a programmed way by introducing and positioning means for working within the pipe from a component of the primary circuit.

Abstract: The maintenance device comprises a robot (2) for carrying out maintenance work in a remote-controlled or programmed way in the maintenance region. The robot (2) comprises a supporting part (2a) connected to a clamping support (3) secured to a means (4) of transporting the maintenance device (1) between an introductory region and the maintenance region. The clamping support (3) comprises clamping shoes (11, 13a, 13b) which can be moved in roughly parallel and opposite directions so as to clamp the maintenance device (1) in the maintenance region. The transport means (4) may be constructed in the form of a carriage comprising two motorized wheelsets. The device and the method of the invention make it possible in particular to carry out machining or inspection work inside primary pipes of a pressurized water nuclear reactor.

Abstract: The present invention provides a motor signature measuring system comprising three current transformers which are placed around the power cables leading to the high voltage motor, an E-Plug module connected to the current transformers and located in a motor control case to reduce the motor voltage from 600V RMS to 5V RMS, jack connectors for connecting the 5V output of the module to a connector mounted through the door of the motor control enclosure, and a cable connector for connecting the door connector to a monitoring system conducting the motor signature testing operations.

Abstract: The maintenance device comprises a robot (2) for carrying out maintenance work in a remote-controlled or programmed way in the maintenance region. The robot (2) comprises a supporting part (2a) connected to a clamping support (3) secured to a means (4) of transporting the maintenance device (1) between an introductory region and the maintenance region. The clamping support (3) comprises clamping shoes (11, 13a, 13b) which can be moved in roughly parallel and opposite directions so as to clamp the maintenance device (1) in the maintenance region. The transport means (4) may be constructed in the form of a carriage comprising two motorized wheelsets. The device and the method of the invention make it possible in particular to carry out machining or inspection work inside primary pipes of a pressurized water nuclear reactor.