Abstract: A control rod drop monitoring method and apparatus for use in a nuclear reactor including dedicated timers and displays to measure and display the drop time for each control rod when a reactor system trip occurs. The timers and displays may be made from CMOS technology, and may have millisecond resolution, if desired. The timers are initiated upon receipt of one or more reactor trip signals and are halted at the closure of switches for detecting the fully inserted or other preselected position of the control rods in the reactor core. The monitor system is designed for continuous operation. Reactor trip signals and limit switch signals are fed to latch circuits following a trip event or limit switch activation and are unaffected by subsequent changes in the state of the limit switches or trip signals. A manual reset is required to clear rod drop times. Also, a test mode is provided wherein all of the timers can be started and stopped, thereby providing a functional test of 100% of the monitor circuitry.

Abstract: The present invention is a component monitoring system which includes all sensor and plant component monitoring at distributed plant locations close to the sensors which are the source of the signals to be analyzed. Minimally preconditioned analog sensor signals are multiplexed and further conditioned by a sensor signal preprocessor that adjusts the gain and filters the sensor signals before being applied to an analog-to-digital converter controlled by a data acquisition computer. The analog sensor signals are converted into digital samples and stored in a data memory of a digital signal processor using a direct memory access technique. The data acquisition computer also controls the selection of sensors, gain, etc., based on commands from a controller. The digital signal processor, once sufficient sensor data has been collected, performs appropriate processing using known digital signal processing techniques.

Abstract: The device comprises a rod (28) on which is mounted a means (30, 32, 33) for support and displacement which is movable in an axial direction of the rod (28) and in two directions perpendicular to this axial direction. The support (33) of the means for displacement carries a device for pushing axially on a longitudinal end of the pencil, which may be activated by a control means (35), one end of which is located in a control station (25) disposed above a fuel assembly storage pool. Means for measuring the axial pushing force on the pencil and the length of axial displacement of the pencil under the effect of the pushing are disposed in the control station (25). A video camera (34), carried by the support (33), makes it possible to provide an image of a zone in the vicinity of the end of the pencil.

Abstract: Evaluation for deformations of the channel box of a fuel assembly in a nuclear reactor core in which deformations of the channel box which has been used for exposure so as to be deformed are estimated for future exposure through computation and process of data of core characteristics, data of material properties and shapes of channel boxes, data of initial deformations of the channel box, data of loading patterns and the like, thereby to judge whether the estimated deformations are agreeable for reusing the channel box in the future exposure, or not.

Abstract: The device comprises a control station (2) fixed on the edge of the pool (3), a rod (4) connected to the support of the control station (2) at its upper end (5), means for adjusting the inclination of the rod (4) relative to the vertical direction, a tool support (12) and an assembly for displacing the tool support (12) in two directions perpendicular to the axis of the rod (4), carried by the carriage (10). Control means (14, 15) may be activated from the control station (2) in order to displace the tool support (12) by means of the displacement assembly.

Abstract: The present invention seeks to obtain reliable detection of the fall of a control cluster into the core of a pressurized water nuclear reactor without giving rise to the risk of an erroneous detection causing an untimely emergency reactor stop. Four acquisition units (U1, U2, U3, U4) corresponding to the four quadrants of the reactor core receive neutron flux signals (F1), heat flux signals (P1), and cluster position signals (Z1A, Z1B, . . . ) from the various quadrants and process them to generate corresponding alarm signals. These units, together with a combination circuit (14, 18, 22), provide a cluster fall signal (24) in the presence of at least two alarm signals including at least one cluster position alarm signal. The invention is applicable to electricity power generation.

Abstract: A modified inspection apparatus for inspecting nuclear fuel pellets for surface defects includes a pellet guide chute assembly extending through an ultrasonic inspection chamber. The guide chute assembly has a support substrate and a pair of elongated plates attached on respective upper adjacent sides of the substrate. The substrate is composed of a pair of elongated wall sections rigidly attached together to have a V-shaped configuration in cross section. The plates are adjustably attached on the substrate wall sections for supporting a nuclear fuel pellet between and on adjacent longitudinal portions of the plates located in the inspection chamber.

Abstract: A tool 10 for creating a standard repeatable impact on a nuclear steam supply system (NSSS) structure to determine the response of an accelerometer to the impact of a known energy at a predetermined location on the structure. The tool 10 includes a barrel 12 and a spring 30 which biases a ram 20 reciprocally mounted in the barrel and driven by cam plates 66 and 68 on shaft 60. A flange 28 on ram 20 compresses spring 30 and upon movement of follower 72 to a maximum radial distance from shaft 60, the follower 72 drops down step 70 on the cam surface to allow spring 30 to uncompress and project ram 20 out of the barrel such that end 22 of the ram creates a standard impact on the NSSS structure. A hand crank drives shaft 60. A dampening spring 78 assists in the rebound of the ram 20.

Abstract: A nuclear reactor monitoring system includes incore and excore neutron detectors for producing neutron noise signals and system data including at least one of reactor coolant pump motor current, temperature and pressure of the reactor coolant system and pressurizer pressure and means for diagnosing conditions of the reactor internals and core support structures by comparing analyzed instrumentation signatures. The signatures may be derived from scale model test data, plant test data, analytical model results and trending analysis results. Analyses performed by the system include time and frequency domain analyses of the neutron noise signals and the system data. The analyzed data are compared with rules in a rules base of an expert system to generate a diagnosis of the condition of the reactor internals and core support structures and action recomendations for maintenance of the nuclear reactor.

Abstract: Lower internals for a reactor--typically a PWR--having a pressure vessel containing a core cooled by an upward flow of coolant and a core support include a shell defining a coolant volume between the support plate and the lower part of the vessel. The shell has a wall and ducts opening into passages of the support plate. Guide tubes passing through the volume receive glove fingers for in-core instrumentation probes. The internals further include a separation plate fixed to the wall means, situated below the core support plate. The separation plate is formed with holes connected to the coolant conveying ducts and is fixed to the lower support plate of the core by mechanical connection means spaced apart at the periphery of the separation plate and defining passages for radial access of the coolant to the distribution chamber.

Abstract: An apparatus for inspecting the quality of both ends of nuclear fuel rods includes a housing with longitudinally separated slots for receiving X-rays downwardly therethrough to define first and second operating positions so that the forward ends of nuclear fuel rods can be aligned below the first operating position and the rear ends of nuclear fuel rods can be aligned below the second operating position. An endless conveyor belt system, having at least three longitudinally spaced apart cleats mounted thereon to define first and second film cartridge receptacles, is adapted to advance X-ray film cartridges longitudinally through the housing. A table is included for selectively lifting the film cartridges from the belt system to maintain the film cartridges stationary while the conveyor belt system is advanced.

Abstract: A snubber monitoring system is provided for a nuclear power plant wherein the snubbers are disposed at various locations within a plant containment structure to support piping and equipment rigidly against seismic and vibratory loading while permitting thermal expansion and contraction thereof. The system includes displacement sensing means and temperature sensing means coupled to each of the snubbers to generate continuous snubber displacement and pipe and equipment temperature signals.A sensor data highway system has cable connected data collectors disposed within the plant containment to collect the snubber sensor signals. A receiver and a programmed computer system are located outside the containment to process the snubber displacement and pipe and equipment temperature signals to provide substantially continuous snubber monitoring. A data highway cable extends through the containment to couple the signal collectors to the receiver and computer system.

Abstract: An electrode probe suited for employment as a reference electrode in a high pressure, high temperature, and high radiation field environment such as adjacent the core of a nuclear reactor is described. The electrode is a brazed and welded assembly consisting of only ceramic and metal parts including a sapphire crucible which is brazed to a kovar/stainless steel housing, welded in turn, to a coaxial cable assembly for signal transfer. The crucible incorporates an integrally formed pedestal through which a conductor wire extends and over which is positioned a selectively coated cylindrically shaped sealing retainer. The device is particularly suited for employment with a silver/silver chloride electrode system.

Abstract: A handling apparatus for a measuring instrument for the inside of a nuclear reactor has a pair of take-up drums each having a wire rope wrapped around it. A measuring instrument is secured to one end of each of the wire ropes. Each take-up drum rotates together with a corresponding drive wheel to raise or lower the measuring instrument by means of the cable. The take-up drums and the drive wheels are connected to a single motor by a gearbox equipped with a clutch which enables power to be selectively transmitted to either of the take-up drums and the corresponding drive wheel.

Abstract: An anomaly diagnosis system for a nuclear reactor core comprises an anomaly detecting unit incorporated into a fuel assembly of the nuclear reactor core, and a transmitter-receiver provided outside the reactor vessel, which transmits a signal by wireless to the anomaly detecting unit and receives an echo signal generated by the anomaly detecting unit by wireless. When the anomaly detecting unit detects an anomaly in the nuclear reactor core, such as an anomalous temperature rise in the fuel assembly, the mode of the echo signal deviates from a reference signal mode, and then the transmitter-receiver detects the deviation of the mode of the echo signal from the reference signal mode and gives an anomaly detection signal to a plant protection system, and thereby the nuclear reactor can be shut down without delay.

Abstract: This invention comprises electrochemical analytical devices capable of performing electrolysis in multiple pulse modes. It relates to a method of quantitative analysis and equipment for performing quantitative analysis which uses a combinaiton of these pulse modes to realize the measurement of concentrations of multiple substances to be measured, occurring in the test solution to be measured and also a nuclear reactor water quality control system employing such a quantitative analysis method.

Abstract: An apparatus is disclosed for determining pressurization in a nuclear fuel rod and the like by measuring the increase in diameter of the fuel rod occasioned by the internal pressurization of the rod, and which includes an elongate receiving member having front and rear ends and an axial opening extending therein. A gauge is positioned in the axial opening and is adapted to coaxially surround the fuel rod and then measure an increase in diameter of a fuel rod caused by a pressure increase within the rod. The gauge is formed from a substantially C-shaped bracket having inner and outer surfaces, spaced distal end portions and a more thin medial portion, and with a strain gauge mounted on both the inner and outer medial surfaces. A microprocessor is connected to the strain gauges so that changes in electrical output from the strain gauges can be measured to determine the increase in diameter of the fuel rod upon pressurization of the fuel rod.

Abstract: A sample of the pressure tube of a CANDU type of nuclear reactor is obtained for deuterium analyses without removing the pressure tube from the reactor. The preferred sampling tool comprises two cutters and means for capturing the removed material, wherein one cutter removes the surface oxide layer, and the second cutter removes a sample for analysis. The cutters and cutting operation are designed to avoid damaging the integrity of the pressure tube to allow it to remain in service. In the preferred embodiment, the sampling tool performs the surface and sample removal operation with the fueling equipment with simple linear movement.

Abstract: An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-bsed materials is disclosed. Samples of zirconium-based materials having different composition and/or fabrication are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280.degree. to 316.degree. C.). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hyriding for the same materials when subject to in-reactor (irradiated) corrision.

Abstract: A modified inspection apparatus for inspecting nuclear fuel pellets for surface defects includes a pellet guide chute assembly extending through an ultrasonic inspection chamber. The guide chute assembly has a support substrate and a pair of elongated plates attached on respective upper adjacent sides of the substrate. The substrate is composed of a pair of elongated wall sections rigidly attached together to have a V-shaped configuration in cross section. The plates are adjustably attached on the substrate wall sections for supporting a nuclear fuel pellet between and on adjacent longitudinal portions of the plates located in the inspection chamber.

Abstract: In a boiling water reactor, provision is made to sample the core bypass region immediate the top guide to determine the physical and chemical constituents of the moderating water. A conduit for a local power range monitor is fitted with a measurement assembly. The conduit and measurement assembly are inserted up to the vicinity of the top guide. A tube opening is provided to the bypass region immediate the top guide. During reactor operation, the saturated liquid in this region flashes to a steam water mixture (18% steam) at constant enthalpy and is rapidly removed from the reactor to measuring equipment in the reactor building. During removal, the radiolytic disassociated gases (namely hydrogen and oxygen) partition to the steam phase where their recombination is retarded and accurate measurement of their constituent content can be made. Also, temperature of the steam water mixture decreases thereby preserving unstable species like hydrogen peroxide that degrade much more rapidly a high temperatures.

Abstract: Apparatus for monitoring an object piece of qualified equipment in a nuclear power plant includes passive means for detecting a total integrated dose of radiation received by the object piece, means for determining a thermal history of the object piece, and means for housing the passive detection means and determining means in close proximity to the object piece. In order to generate and maintain evidence to assure that qualified equipment used in such stations will operate on demand, the passive detection means and determining means mounted in close proximity to the object piece are exposed to the environment of a normally operating power plant and periodically analyzed to determine whether the qualified life of the object piece may be extended.

Abstract: In a hydraulic system of a control rod drive for insertion of control rods, a test apparatus and test process is disclosed for determining the integrity of the hydraulic system check valve for preventing inadvertent control rod ejection responsive to the reactor pressure. The check valve is located between a hydraulic valve for causing rod insertion and the hydraulic cylinder. In this interval downstream of the check valve, alteration is made to the hydraulic path by the installation of a conduit with a quick disconnect connected through an isolation valve. Test apparatus for temporary connection at the disconnect is disclosed consisting of a small positive displacement piston and cylinder. The small positive displacement piston and cylinder connects to the quick disconnect through a complementary quick disconnect fitting and a rapidly opening toggle valve. Provision is made for timing the excursion of piston in the piston and cylinder preferably by end of stroke microswitches.

Abstract: Apparatus is disclosed for locating and measuring wear in nuclear reactor control rods. Circumferential and radial eddy current test probes produce outputs corresponding respectively to volume and thickness of the control rod. A method is disclosed to determine cladding wear when the volume and thickness varies beyond preselected limits.

Abstract: In an automated first weld apparatus, a transporter conveys nuclear fuel cladding tubes successively to a welding station where a separate end plug is welded to an open end of each tube. Thereafter, the transporter indexes the tubes successsively through a cooldown station where the weld is cooled, to a reader station where a unique end plug serial number is read, and then to a succession of inspection stations where the internal and external weld characteristics are automatically examined. The resulting inspection data is correlated with the associated serial number for record purposes and tested against quality assurance standards pursuant to sorting the tubes into accepted and rejected lots.

Abstract: A method of compactly storing digital data includes the steps of sequentially entering a plurality of digital data entries into a shift register having a plurality of stages and summing successive groups of data entries as they are outputted from the shift register to produce singly compressed data entries that are subsequently entered into a second shift register. Sets of singly compressed data entries which are outputted from the second shift register are again added to obtain doubly compressed data entries which are entered into a third shift register. This process continues until a sufficient number of shift registers have been employed to accommodate all of the expected data entries.

Abstract: An apparatus for the inspection of nuclear reactor fuel rods combined in fuel rod clusters with spaces therebetween in a fuel assembly includes test probes each being disposed at a different respective level along the fuel assembly. Fingers are each part of a respective one of the test probes. Ultrasonic test heads are each disposed on a respective one of the fingers. The test heads are inserted into the spaces between the fuel rods, and the insertion position of each of the test probes is corrected independently of the insertion position of the others of the test probes, before insertion of the test probes.

Abstract: An apparatus for the inspection of a nuclear reactor fuel assembly in a water-filled pool having fuel rods with spaces therebetween includes an encapsulated housing having walls and being disposed in the water-filled pool. A lever drive mechanism is disposed in the housing. A rotary transmission connected to the lever drive mechanism extends through one of the walls of the housing toward the surface of water in the pool. A plurality of stationarily disposed lever linkages jut out from the rotary transmission outside the housing. Each of the lever linkages has an end facing away from the rotary transmission. Carriers are each connected to a respective one of the ends of the lever linkages. Probes are each disposed on a respective one of the carriers. Ultrasonic test heads are each disposed on a respective one of the probes for insertion into the spaces between the fuel rods.

Abstract: Nuclear reactor having upper internal equipments and lower internal equipments with probe guiding means including probe guide ducts for slidably guiding flexible elongated probes into the core. The upper internal equipments include a plenum chamber for collecting the coolant coming out of the core and separating means disposed between the core and the plenum chamber, said separating means including two plates, one upper and one lower, connected at their periphery by an envelope and comprising spacer tubes through which pass the coolant so as to avoid passing through cluster guides for guiding the clusters of control elements into and out of the core. Lower internal equipments placed between the core and the bottom of the vessel include an enclosure and ducts for the passage of the coolant entering the core, said enclosure thus defining a tranquilized volume in which the probe guide ducts are distributed.

Abstract: A safety control safeguard system includes sensors disposed in a quadruple array. First, second, third and fourth signal processing channels provided in parallel each include signal processing means for receiving output signals from the associated sensors for producing a trip signal. First and second actuating means operable independent of each other are provided for actuating an apparatus to be controlled. A first switch means operatively connected to the first, second, third and fourth processing channels for activating the first actuating means in response to reception of the trip signal produced by the signal processing channels and second switch means operatively connected to the first, second, third and fourth processing channels for activating the second actuating means in response to reception of the trip signal produced by the signal processing channels.

Abstract: A device for remotely measuring the diameter of wire rope. The device includes a linear variable differential tansducer, a mechanism to guide and clamp the rope in relation to the anvil of the transducer, an elongated handle for manually manipulating the transducer and the guide and clamp mechanism.

Abstract: The apparatus for simulation of the operation of a pressurized water reactor comprises a reactor vessel with a heater, a steam generating vessel divided into a primary portion and a secondary portion. The reactor vessel and the primary portion of the steam generating vessel are connected in a primary feed loop equipped with a coolant pump, while the secondary portion of the steam generating vessel and the condenser unit are joined together in a secondary feed loop provided with a feed pump. With such an apparatus a steam generating vessel heating pipe break or fracture can be simulated when the reactor vessel and a primary part of an additional steam generating unit are joined in another primary feed loop parallel to the primary feed loop having the cooling pump and a secondary part of the additional steam generating unit is connected with the aid of at least two branch pipes to the secondary feed loop.

Abstract: A sleeve-like sealing unit, particularly a thimble guide extender is adapted to be installed on and coaxially with a thimble guide mounted on a lower core plate of a nuclear reaction vessel and projects toward and terminates at a distance from an underface of a bottom nozzle forming part of a fuel assembly and supported on a top face of the lower core plate. The extender has a lower fitting adapted to surround the thimble guide at a bottom length portion thereof adjacent the top face of the lower core plate; and an upper fitting being axially displaceable relative to the lower fitting and being adapted to surround the thimble guide at a top length portion thereof. The upper fitting has an axial passage and a terminal opening defined by an outer annular radial face adapted to be in a face-to-face engagement with the underside of the bottom nozzle.

Abstract: An apparatus for the remote examination of peripheral shroud hold down bolts on steam separators used in boiling water reactors is disclosed. The testing apparatus is used while the steam separator is in a holding pool. The testing apparatus includes a depending pole having attached at the bottom thereof an aluminum shoe. The shoe has a flat, upwardly exposed bottom and opens to one side at gathering surfaces to receive the lower ends of the shroud bolts. The upwardly exposed bottom of the shoe defines an aperture through which an upwardly exposed piezoelectric device is exposed for direct contact with the bottom of the bolt. An overlying clamp member is provided to clamp the lug onto the bottom of the shoe for testing. The clamp member defines a receiving slot to receive the shaft of the bolt immediately overlying the rectangular lug. The clamp member slides over the shaft at the slot into contact with the top of the lug.

Abstract: An elongate selector member 16 having a Y-shaped passage 16a extending through it is rotatably mounted in a hollow cylindrical housing 12 and driven or switched between fixed limit positions by an actuator 17 to establish a continuous insertion/withdrawal path for a flexible cable between an inlet guide tube 11 and either one of two outlet guide tubes 19a, 19b.

Abstract: Device for flushing out a lance-housing tube in a reactor pressure vessel of a boiling-water reactor and for aligning therein a dry LVD lance which partly protrudes with a pressure-tight lance passthrough from an end flange on the lance-housing tube of the reactor pressure vessel, includes a tubular housing surrounding from below a part of the lance protruding from the reactor pressure vessel and sealed by a lance protection tube, the tubular housing being fastenable to the end flange; and a piston arranged in the tubular housing underneath the sealed lance, the piston being vertically displaceable and rotatable.

Abstract: An inspection gage includes an expandable tubular member insertable within a fuel assembly guide thimble. The tubular member has embossments circumferentially spaced from one another about the tubular member and being the same in number as bulges on the guide thimble and configured to fit therewithin. An expansion member rotatably mounts the tubular member and is axially movable therewithin when the latter is inserted within the guide thimble. The expansion member is movable between a first position in which it permits inward contraction of the tubular member an displacement of its embossments away from the guide thimble bulges for removing the embossments from registry therewith and a second position in which it produces radial expansion of the tubular member and displacement of its embossments toward the guide thimble bulges for placing the embossments in registry therewith.

Abstract: A thimble guide extender for use in a nuclear power plant shields the thimble from turbulence and absorbs vibration from it to reduce wear. The extender includes an extension element which is secured to the thimble guide, a spring bushing which is attached to the extension element and which contacts the thimble, and a cap. In one embodiment the cap clamps the spring bushing, and in other embodiments it simply shields the bushing from turbulence. In one embodiment the spring bushing is an element having a cylindrical portion on one side and dimpled spring legs on the other, in another embodiment the spring bushing is an element having bent spring legs on either side, and in another embodiment the spring bushing is provided by dimpled spring legs that are integral with the extension element. The extension element is provided with locking fingers which secure the extension element to the thimble guide.

Abstract: A method for checking the dimensions of a nuclear reactor fuel assembly in a water tank uses two mutually parallel probes each having a first probe side carrying an ultrasonic test head at a free end thereof with acoustic directions directed towards each other and each having a second probe side facing away from the ultrasonic test head. The acoustic waves are transmitted with one of the ultrasonic test heads, the transmitted acoustic waves are received with the other ultrasonic test head, one of the second probe sides is brought into contact with a given region of the fuel assembly to be checked, the probes are moved towards each other in the direction of the acoustic waves due to contact pressure with the fuel assembly, the probe movement is indicated and assessed by a reduction of transit time of the acoustic waves between the test heads, and the actual dimension of the fuel assembly region to be checked is derived while accounting for the dimension of probe movement.

Abstract: For sampling a section of a tube in a nuclear fuel assembly comprising a skeleton formed by two end-pieces joined by tubes fixed to the end-pieces an opening for the extraction of the section through the upper end-piece is cut out with a milling cutter. Through the inside and by means of a cutting tool, the spacer tube is cut off at a level beneath the junction with the upper end-piece by means of a rotary tool introduced through the upper end-piece and the latter is extracted through the opening.

Abstract: A method and apparatus for monitoring and generating on-line, real time displays of two and three dimensional nuclear reactor core power distributions and off-line, periodically updated summaries of three dimensional nuclear reactor core power and burnup distributions is disclosed. The method and apparatus makes use of information obtained from inlet sensors and core-exit thermocouples to produce enthalpy rise values. Flux measurements are combined with enthalpy to produce power values. In one aspect of the invention, deviations from reference values are classified and displayed on a two-dimensional color graphics terminal where the various classifications are displayed according to a color code which enables a rapid and convenient method of analysis of the dynamics of the reactor operation.

Abstract: A method and apparatus for generating core-average and radially peaked, pointwise core axial power distributions of a nuclear reactor core from monitored responses of reactor flux obtained exterior of the core, such as responses obtained from ex-core power range nuclear detectors, and from responses of temperature obtained from selected core-exit thermocouples.

Abstract: A system which measures direct current drift and noise in sensor signals from redundant sensors utilizing a parity-space algorithm. Parity vector signals produced by the parity-space algorithm are averaged to provide a direct current (DC) drift signal. An instantaneous noise signal is found by subtracting the direct current drift signal from a parity vector signal for one of the samples of the sensor signals. The RMS value of the instantaneous noise signals are averaged to provide sensors noise signals.

Abstract: An automatic indexer assembly provides a constant axial and rotational movement of a signal-producing-sensor probe through a zone of a tube to be inspected with minimum signal distortion by means of threadedly connected fixed and movable hollow shafts. A probe tail is clamped to move with the movable shaft as it is driven by a constant speed drive motor and gear reducer mounted on a platform and guide plates which move with the movable shaft along a track parallel to the shaft axis. A water swivel and electrical slip ring also move on a platform with the movable shaft. Twisting of a fluid supply line and an electrical supply line is minimized. Constant sensor movement can be accomplished to permit accurate recording of the signal produced on a recording means with a minimum of backlash, acceleration, deceleration, or intermittent variables distorting the signal.

Abstract: A device is provided concerning probe guide ducts for allowing a probe to pass for carrying out measurements on the coolant at the outlet of the assemblies (6) of a nuclear reactor (1). It comprises sleeves (9) through which pass upper columns (12) grouping together the guide ducts (10). After running horizontally along the periphery of the reactor, the ducts (10) are grouped together in lower columns (15) in the form of bundles before running towards the outlets of the coolant from the asemblies (6) in contact with the upper plate of the core (7). This device provides an excellent protection of the probe guide ducts with respect of the flow of the coolant.

Abstract: In the device, over at least a portion of the length of the internal guide passage (5) of the tubular conduit (15) extends a retaining tube (16) exhibiting radial undulations (17) such that the cylinder having straight generatrices which is inscribed in the tube (16) corresponding to the free passage in a straight line, within this tube, has a diameter less than the diameter of the tubular conduit (15). In the case of a pressurized water nuclear reactor, the retaining tube (16) is preferably placed, at least in part, within the lower core plate (2).

Abstract: This reservoir is provided with a fluidtight covering of metal sheets (6) welded together by their edges and supported by beams (2) each comprising two section elements (16, 17), and three longitudinal channels. In cross-section, the first and second section elements each have the shape of a U, the second section element being substantially narrower than the first element and being fixed by the edge of its branches (18) to the bottom (19) of the first section element so as to define an inspection channel (20) having a closed cross-section, and two draining channels (21) having an open cross-section. The surface of the bottom of the second section element (17) which faces toward the exterior of the beam, is contained in the same plane as the edges (11) of the branches (22) of the first section element (16) so as to form the bearing surface (14) for two sheets (6) in the region of their weld head (15) which interconnects them.

Abstract: A pipe end sealing element for use in hydrotesting of plain end pipes, such as instrument tubes of a nuclear reactor, and a method of hydrotesting of such plain end pipes. The pipe sealing element comprises a hollow tubular member with a bevelled flange at one end and a threaded portion at the other end, a hollow conical bushing of decreasable interior diameter having a bevelled surface complementary to the bevelled surface of the tubular member, a sealing plug and threaded bolt engageable with the threaded portion of the hollow tubular member. The hollow tubular member is placed about the open end of the pipe, with the threaded portion thereof adjacent and beyond the pipe end, and inserting the conical bushing into the hollow tubular member about the pipe wall. The sealing plug is placed on the pipe end and the threaded bolt engaged with the threads of the hollow tubular member.

Abstract: A device for detecting the presence of a gaseous phase in the tank of a nuclear reactor closed by a cover on which a plurality of vertical tubes are fixed for adapting the control rod actuating mechanism of the reactor, the adapter tubes not provided with mechanisms being closed by a plug. A plurality of ultrasonic emitter-receivers are fixed to the upper surface of the plugs of at least a part of the adapter tubes, so that each emitter-receiver directs an ultrasonic beam to the inside of the tank parallel to the axis of the corresponding tube, the latter being provided with at least one reflector placed inside the tube in the path of the ultrasonic beam. The invention is especially applicable to reactors cooled by pressurized water.

Abstract: An electromechanical measurement system for acquiring dimensional data from a nuclear fuel assembly includes an underwater measurement assembly, support means for mounting same on the edge of a spent fuel pool, and a control and data acquisition and processing unit. The underwater measurement assembly includes an elongated strongback disposable vertically along a pool wall, adjustment mechanism on a support assembly permitting accurate positioning of the strongback. A support plate carries hydraulic clamps for clamping the top nozzle of a fuel assembly while it is supported from overhead, a clamp carried by the lower end of the strongback being engageable with the bottom nozzle for holding the fuel assembly in a selected orientation with the faces of the top nozzle vertical.