Abstract: A testing device for load-testing of specimens (3) in a nuclear reactor environment is attached to one of the pipes (1) of the reactor for conveying a first medium under pressure and provides a first space (14) in open communication with the pipe (1). In the first space (14) a movable pull rod (15) is arranged, one end of which is attached to one half (16) of a specimen (3) arranged in the space (14). The other end of the pull rod (15) is joined to a tensile force device capable of being influenced by the first medium for achieving a tensile stress in the specimen (3).

Abstract: Electrically controlled load activating mechanisms that can be used inside the containment vessel of a nuclear reactor in conjunction with bellows-loaded DCB crack growth sensors installed inside the reactor pressure vessel or piping of a nuclear reactor. One mechanism is a liquid-filled, double-bellows master/slave arrangement connected by a capillary tube to transmit the loading provided by a linear motion device. Another mechanism uses a heat-resistant gas bottle that can be heated in a furnace to increase the gas pressure to expand the bellows of the DCB sensor. A third mechanism uses a pump or compressor to provide the necessary expansion force. The loading is controlled via electrical connections that do not require special pressure boundary penetrations of the containment vessel.

Abstract: This invention concerns a plant diagnosis apparatus which comprises a first detecting unit for detecting a state of devices constituting a circulating system in which a fluid circulates and a second detecting unit for detecting a quality of the above-mentioned circulating fluid or a quality of the atmosphere surrounding the above-mentioned circulating system, and a unit for diagnosing the circulating system by output of the first and second detecting units, and therefore can diagnose anomaly of facilities such as a nuclear power plant with high reliability and rapidity.

Abstract: An instrumentation probe for measuring and recording one or more physical parameters within a fluid duct, such as a nuclear reactor fuel channel, comprises one or more sensors responsive to the parameters to be measured and recorded, each sensor actuating a scribe which cooperates with a recording chart. The recording chart is a drum, disc, tape or like recording device, and is volume driven at a constant speed by an impeller responsive to the fluid flow. No connections to external instrumentation are required.

Abstract: A device for creating a corrosive condition and monitoring that condition, adapted for use in active fluid heat exchangers, has a sealed tube having corrosion sites spaced apart on its exterior. Structure positioned on the outside of the tube at the corrosion sites provide crevices for retaining corrosive sludges in contact with the corrosion sites. The interior of the tube is pressurized with a gas to stress the tube thereby accelerating corrosion of the surface. Heat is provided to the corrosion sites from within the tube when placed within a fluid reservoir, such as the secondary side of a fluid heat exchanger. The heat concentrates corrosive chemicals at the corrosion sites, also accelerating corrosion. Electronic probes can be mounted in proximity to the corrosion sites for real time monitoring of the progress of corrosion.

Abstract: A tool for remotely obtaining an impression of a component surface. A support pole is used to position a mold plate having side walls against a component surface. One or more locating guides are provided with seals for sealing off bores in the component surface. A piston from a pneumatic cylinder attached to the support pole is used to force impression material into the space between the mold plate and component surface. An adhesive plate on the mold plate insures removal of the impression with the mold plate once the impression material has hardened. The surface condition of the component may be determined by inspection of the negative impression.

Abstract: The tank contains a liquid in which a radioactive element emitting gamma radiation is distributed substantially uniformly. A photon detector (15) is disposed in the vicinity of the external surface of a measurement, zone of the wall (12), a count is made of the photons emitted by the radioactive element distributed in the liquid (7), through the measurement zone of the wall (12), a number of photons determined by counting is compared with a reference number emitted by the radioactive element through a reference zone of the wall (12) and it is deduced therefrom whether or not a defect (8) is present in the measurement zone of the wall (12). The measurement device comprises a photon detector (15) disposed at one end of a collimator (16) made of heavy metal traversed by channels or windows (17).

Abstract: To measure the spring constants of double-acting, fuel rod-centering springs assembled with different pairs of ferrules in a nuclear fuel bundle spacer, a gauge is provided to include an alignment rod and a probe carried by a handle for insertions into the ferrules of a ferrule pair. The alignment rod loads the side of the spring acting in its ferrule, while the other spring side exerts its fuel rod-centering force on a load cell carried by the probe. A micrometer acts against the load cell to produce measured deflections of the other spring side. Spring constants are calculated from electrical readouts of spring force and spring deflection from the load cell and micrometer.

Abstract: Process for carrying out an inspection or monitoring round of a nuclear site (5) consisting of recording in a central computer (22) connected to a radiofrequency transmitter/receiver or transceiver (24) a combination or sequence of different collection or reading points (12) grouping sensors (10) constituting the round to be performed. This sequence is copied again into a memory of a portable microcomputer (26) carried by the watchman during the round and connected to a radiofrequency transceiver (28). Each connection point (12) is validated and this validation is transmitted to the central computer (22). The sensors are checked and any abnormal condition is indicated to the central computer (22). All the checks carried out are recorded in the portable microcomputer (26) and copied again at the end of the round into a central computer (22) in such a way that they can be processed.

Abstract: The duplex tube (1) comprises a tubular core (2) and a cladding or covering layer (3) made from an alloy, the base metal of which is identical to the base metal of the alloy constituting the tubular core (2).

Abstract: Calibration arrangement and method for calibrating an inspection instrument, such as an ultrasonic inspection instrument of the kind typically used to inspect nuclear reactor pressure vessels. The calibration arrangement comprises a frame, the frame having a slot therethrough for receiving a sonic coupling medium therein. A plurality of channels extend through the frame and are in communication with the slot. A plurality of solid reflector rods matingly extend through respective ones of the channels and into the slot for reflecting sonic energy produced by a pulsed ultrasonic transducer mounted on the frame and spanning the slot. A pulser stimulates the transducer so that the transducer transmits pulsed sonic energy through the coupling medium which resides in the slot. The sonic energy is reflected by the rods that pass through the slot. The reflected sonic energy or echoes are received by the transducer, which generates transducer output signals that are received by a receiver.

Abstract: A method and system for electrochemically measuring the sensitization to stress corrosion cracking of small pipes of plant structural members in a very short period of time. A micro electrochemical cell for measuring the sensitization is movably disposed within a BWR plant ICM housing or an instrumentation pipe. An electrochemical instrumentation system for performing the electrochemical potential variation such as pulse voltammetry and a remote control system for remotely controlling the electrochemical cell are provided.

Abstract: A tool system features an EDM crack removal head which can be positioned within a nozzle/thermal sleeve and then oscillated in a vertical and horizontal directions to enable precise material removal, leaving the surface free of cracks and in a condition for rewelding if required. If a crack is found in a surface which is covered by a thermal sleeve, then a small specialized EDM cutting head, which can be delivered from the under side of the reactor head and has an ability to pass through the ID of the thermal sleeve, is used. The head enables the sleeve to be cut and allow a portion of the sleeve to be removed. The crack can then be removed using a crack removal type of head.

Abstract: A method and an apparatus for estimating the remaining service life of an object by measuring the physical quantity of a sample and which are applicable to a nuclear reactor. A plurality of model samples are experimentally prepared by preliminarily disposing a material having substantially the same composition as the object to be subjected to the estimation in an environment substantially the same as the object and measuring the physical quantities of the model samples in relation with exposure times. From this experiment, a relationship between the exposure time and the physical quantity under such an environment is obtained. Further, a critical exposure time which will cause an unstable fracture of a material of an actual sample, which is made of a material substantially the same as the object and placed in substantially the same environment as the object, is preliminarily obtained from the relationship between the exposure time and the physical quantity for the model samples.

Abstract: A gap between adjoining fuel rods arranged in a fuel assembly is measured by setting a measuring device into a fluid such as cooling water or a specific gas such as CF filling a container, placing a fuel assembly to be inspected to a predetermined portion the container, measuring a distribution of the fluid distributed in the fuel assembly by utilizing a nuclear magnetic resonance imaging system, and measuring the gap in accordance with the fluid distribution as an image. The measuring device includes a magnet unit and a coil unit for generating magnetic fields, and the coil unit includes a coil assembly for generating gradient magnetic fields and a coil assembly for generating a high-frequency magnetic field and detecting a nuclear magnetic resonance signal.

Abstract: A method and apparatus of sensing obstructions in a tubular nuclear fuel rod or the like is disclosed. A nozzle is positioned to engage one end of the fuel rod in an air tight sealed relationship. A first gas flows through the nozzle and then into and through the fuel rod. The pressure is sensed and compared with a standardized pressure to determine any excess back pressure in the nozzle indicating an obstruction present therein.

Abstract: A method for detecting and identifying faulty component candidates during off-normal operations of nuclear power plants involves the qualitative analysis of macroscopic imbalances in the conservation equations of mass, energy and momentum in thermal-hydraulic control volumes associated with one or more plant components and the functional classification of components. The qualitative analysis of mass and energy is performed through the associated equations of state, while imbalances in momentum are obtained by tracking mass flow rates which are incorporated into a first knowledge base. The plant components are functionally classified, according to their type, as sources or sinks of mass, energy and momentum, depending upon which of the three balance equations is most strongly affected by a faulty component which is incorporated into a second knowledge base. Information describing the connections among the components of the system forms a third knowledge base.

Abstract: A device for inspecting the support plates and tube sheet of a nuclear steam generator. The device comprises a boom for extending into an access port of a steam generator and into a lane separating two rows of members, the boom being uprightable within the lane; and a video camera attached to the boom for inspecting the tube members and support plates within the lane when the boom is uprighted. In a preferred embodiment the video camera comprises a charge coupled device affixed to a remotely controlled pan and tilt mechanism.

Abstract: In the nuclear fuel assembly, a plurality of plate-shaped straps are assembled perpendicularly to each other into a grid shape so as to construct a supporting grid providing with a plurality of cells each of which inserts a nuclear fuel rod. A cell-size inspection device which inspects the size of each cell of the supporting grid is constructed by a base, a movable plate, a plurality of pin gauges and a sensor. The supporting grid is mounted on the base, while the movable plate is provided such that it can relatively move with respect to the base in a longitudinal direction of the cells. The pin gauges are provided on the movable plate such that they can freely slide along the longitudinal direction of the cells. When the movable plate is moved toward the supporting grid, each of the pin gauges is to be inserted into each of the cells. The sensor senses whether the pin gauge is relatively moved from the predetermined position of the movable plate.

Abstract: Segmented instrumentation tube including a locking sleeve for interlocking the segments of the instrumentation tube, so that the threaded ends of the instrumentation tube do not unthread when subjected to vibration, such an instrumentation tube being suitable for use in a nuclear reactor pressure vessel. The instrumentation tube has a first member having a threaded end portion that has a plurality of first holes circumferentially around the outside surface thereof. The instrumentation tube also has a second member having a threaded end portion that has a plurality of second holes circumferentially around the outside surface thereof. The threads of the second member are caused to threadably engage the threads of the first member for defining a threaded joint therebetween. A sleeve having an inside surface surrounds the end portion of the first member and the end portion of the second member and thus surrounds the threaded joint.

Abstract: The measurement device (3) comprises a metallic body (5) fixed in the wall of a reactor coolant pipe (1). The metallic body (5) is pierced in order to constitute a first channel (8) communicating with the inner volume of the pipe (1) by a plurality of coolant inlet openings (14) and a second channel (12) communicating with the inner volume of the pipe (1) by at least one coolant outlet opening (15) and enclosing the measurement probe (20) in a zone situated between the end of the second channel (12), which end is connected to the first channel (8), and the coolant outlet opening (15). The coolant inlet openings (14) are disposed on one frontal face of the metallic body (5) in the flow direction (6) of the coolant, and the outlet opening (15) is located on an opposite face thereof.

Abstract: A method of ultrasonic inspection of a reactor nozzle involves positioning emitter and receiver transducers on the exterior vessel wall near the nozzle. The emitter transducer is positioned and oriented so that it generates both primary longitudinal and transverse beams. The primary longitudinal beam is directed at the inner radius of the nozzle. The transverse beam is directed toward the interior vessel wall at an angle of about 30.degree. to the normal. Most of the energy of the transverse beam is converted to a secondary longitudinal beam also directed toward the inner radius of the nozzle. The receiver transducer is positioned on the exterior vessel wall about 4" from the emitter transducer so as to be able to detect reflections of both the primary and second longitudinal beams from any potential defect in the nozzle volume adjacent the inner radius.

Abstract: The present invention is a detector assembly for a nuclear reactor which includes platinum detector segments (30-40) axially distributed end to end within a reactor assembly and vanadium detector segments (42-52) spatially congruent in the same assembly. The vanadium detectors calibrate the platinum detector signals to remove the flux contributions of fission products. A full length vanadium detector (62) and a full length platinum detector (90) can be substituted and used to determine the compensation for the platinum detector segments (30-40). The compensated platinum detector signals, wherein compensation consists of isolating that portion of the total platinum detector signal that is directly proportional to the current local heat deposition rate in the reactor fuel, can be used for reactor core protection purposes.

Abstract: An eddy current probe, especially useful for inspecting roll plugs in steam generator tubes, has a detector assembly which is automatically extended and retracted in a radial recess in an elongated body to prevent damage to the detector assembly during insertion and retraction of the probe, yet permits the detector assembly to bear against the interior wall of the roll plug, including the roll zone. In one embodiment, a mechanical actuator in the form of a pair of finger members spring biased radially outward operate through camming surfaces to rotate a pivot member which retracts the detector assembly; however, when the probe enters the bore of a roll plug the fingers are compressed to permit extension of the detector assembly. In another embodiment, a linear actuator such as a solenoid rotates the pivot member to extend and retract the detector assembly.

Abstract: To inspect the integrity of pressurization holes in nuclear fuel rod final end plugs, quality assurance apparatus includes an adapter providing a test chamber into which an end plug is inserted. Instrument gas is introduced into the test chamber through an inlet port for flow to the atmosphere through the pressurization hole. A mass flow meter measures the gas flow through the pressurization hole to determine if it is capable of supporting requisite pressurization of a nuclear fuel rod when the end plug is assembled thereto.

Abstract: A computerized dynamic displacement control display system which involves a method of systematically simulating and displaying the cyclical components of the behavior of the movement of a curve represented by a plurality of points relating two parameters to one another, in which the following are obtained: a moving average of the curve; an oscillator, a series of dynamic moving averages; and pairs of inner and outer envelopes. This derived information is used as a basis to calculate the probable future movement of the curve. An industrial application of this method controls the flow of coolant liquid through a chamber. A temperature sensor supplies information to a computer which is in communication with the use through a keyboard and control unit. Sets of lines relating to the above derived information are displayed on a V.D.U.

Abstract: The fuel element is inserted into a cell, under water, in a pool. Compressed air is injected into the cell so as to drain it of water. A pressure lower than atmospheric pressure is established in the cell and then the gases contained in the cell are subjected to a stream of scavenging air and are collected in a fission product detecting unit. Water is introduced into the cell, the water in the cell is sampled, fission products in the water sample are detected and the fuel element is extracted from the cell if the possible presence of a leak from this fuel element is determined. The detection device comprises a cell having an elongate body delimiting a central housing for a fuel element and end parts. One of the end parts, traversed by a channel in the extension of the housing, receives a cap for opening and closing the channel associated with a remotely controllable opening and closing arrangement.

Abstract: Coolant fluid is sampled from a substantially horizontal part of the hot leg (16) of the primary circuit at at least three points (19, 20, 21) distributed at the periphery of a straight section of the hot leg (16), in such a way that at least one of the sampling points, or lower sampling point, is situated beneath the axis (17) of the hot leg (16). The temperature of the coolant water sampled at each of the sampling points (19, 20, 21) is measured at its outlet from the hot leg (16) and the coolant fluid is reintroduced into the hot leg (16) at a point situated in a position substantially diametrically opposite, with respect to one of the lower sampling points (20, 21), on the straight section of the hot leg (16). The sampling devices (19, 20, 21) connected by pipes (24, 25, 26) to a reintroduction element (22) are arranged at the periphery of the hot leg (16).

Abstract: An improved method for monitoring the parameters of a nuclear core reactor, analyzing the monitored parameters using modified two neutron group diffusion equations that have been subjected to space-time factorization such that the factorized shape functions are approximated, and determining the core neutronics in constant time steps, for example, about 0.25 seconds, thereby to provide analysis and simulations in full range of core neutronics in real-time continuously.

Abstract: The duplex tube (1) comprises a tubular core (2) and a cladding or covering layer (3) made from an alloy, the base metal of which is identical to the base metal of the alloy constituting the tubular core (2).

Abstract: An electrode probe suited for employment as an electrical potential reference electrode in an aqueous, high pressure, high temperature, and high radiation field environment such as 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 base which is brazed to a kovar/stainless steel housing, welded in turn, to a coaxial cable assembly for signal transfer. The base 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: Using the method of this invention, the performance of a heat exchanger is determined by measuring the heat transfer capabilities of an individual tube. A relatively small reservoir of service fluid is connected to the inlet and outlet ports of a tube. The reservoir is provided with a heater or chiller and the service fluid is circulated through the tube. When a steady state is reached, the heat transfer characteristics of the tube are measured using known mathematical relationships.

Abstract: To measure fuel rod-centering forces exerted by double-acting springs assembled with different pairs of ferrules in a nuclear fuel bundle spacer, a gauge is provided to include an alignment rod and a probe carried by a handle for insertions into the ferrules of a ferrule pair. The alignment rod loads the side of the spring acting in its ferrule, while the other spring side exerts its fuel rod-centering force on a plunger mounted in the probe. The plunger is mechanically linked to a load cell which provides an electrical readout of the spring force.

Abstract: An instrument thimble tube shroud is disclosed which has a central cup surrounding an instrument thimble tube to prevent fluid flow about the thimble tube from causing excessive vibration and wear. A plurality of arms extend radially from the cup and have locking pads disposed at the end of each arm, springs extending from the pads engaging recesses in an adjacent fuel assembly nozzle to lock the shroud in position and prevent movement of the shroud or rotation due to aggressive fluid flow. Utilizing the inventive shroud prevents vibration and damage to the instrument thimble tube passing therethrough, enhancing life of the instrumentation and allowing continued observation of dynamics within a nuclear reactor.

Abstract: A method and apparatus for eliminating erroneous test readings during ultrasonic testing of nuclear fuel rods is disclosed. The method and apparatus employ the use of an alignment guide to effect temporary alignment of the fuel rods during testing in order to overcome mechanical deviations of the rods. The alignment guide includes elongated, parallel guide bars for insertion into an array prior to ultrasonic testing. The bars are mounted on a base which, in turn, is movably mounted on a support element. A fixed guide maintains the relative position of the guide bars with respect to one another. A hydraulic element causes the guide bars to be inserted and removed from the fuel array. In a preferred arrangement, the alignment guide may be removably mounted on a frame containing the ultrasonic testing equipment.

Abstract: A fuel pellet surface defect inspection apparatus has an infeed conveyor, a discharge conveyor, and a slide and inspection assembly between the conveyors. The assembly includes a slide defining an inclined track having exit and entry ends adjacent the respective discharge and infeed conveyors. The entry end is at a higher elevation than the exit end. The assembly also includes an inspection station located along the track between its entry and exit ends. The station has lower and upper sound reflectors configured to define an annular inspection chamber through which a pellet moves as the pellet slides down the inclined track. The chamber completely encloses the cylindrical surface of the pellet as the pellet moves through the chamber. An ultrasonic inspection head is mounted at the station and transmits and receives sound energy to and from a pellet as it moves through the chamber such that the sound energy completely surrounds the moving pellet being inspected within the chamber.

Abstract: Water testing means for monitoring circulating water coolant in service within a water cooled nuclear reactor system is disclosed. The invention is an improvement in such water testing means and comprises a preconditioning of selected sample specimens of water with a pH adjustment for inhibiting unbalancing an equilibrium in solution of water coolant with ion solutes.

Abstract: An improved tube inspection probe is provided of the type that includes a probe head assembly formed from a body member and a sensor holding mechanism for resiliently engaging a sensor against the inner wall of a tube, and a flexible cable connected to the body member for helically moving the probe head assembly through the interior of a tube. The improved probe comprises at least one centering device mounted around the cable for maintaining the axis of rotation of the cable in alignment with the center line of the tube, which includes a plurality of resilient finger members mounted around and extending away from the periphery of an annular member that surrounds the cable, wherein each of the fingers is canted at an acute angle with respect to a line tangent to the periphery of the annular member.

Abstract: A robotic device for remotely inspecting pressurizer heater wells is provided which has the advantages of quickly, precisely, and reliably acquiring data at reasonable cost while also reducing radiation exposure of an operator. The device comprises a prober assembly including a probe which enters a heater well, gathers data regarding the condition of the heater well and transmits a signal carrying that data; a mounting device for mounting the probe assembly at the opening of the heater well so that the probe can enter the heater well; a first motor mounted on the mounting device for providing movement of the probe assembly in an axial direction; and a second motor mounted on the mounting device for providing rotation of the probe assembly. This arrangement enables full inspection of the heater well to be carried out.

Abstract: The present invention is addressed to an electrode for evaluating electrochemical potentials which has a robust structure particularly suiting it for employment within the rigorous environment of the reactor core of a nuclear power facility. The electrode of the present invention is comprised of five major segments: a metal cap electrode, an elongate annular metal jacket, an alumina retainer (i.e. an insulator), an annular metal transition sleeve, and a positioning and signal transfer assembly. The metal cap electrode has a tip and annulus extending therefrom formed from sidewalls having an interior surface. The elongate annular metal jacket has an annular retainer securing mouth at one end and an outlet at its opposite end. The alumina retainer has a recessed cap securing portion nestably secured to a portion of said cap electrode interior sidewall surface, and an oppositely-disposed recessed jacket securing portion nestably secured to the interior of said annular jacket retainer securing mouth.

Abstract: There is disclosed an inspection apparatus for peripheral surfaces of nuclear fuel pellets which includes a handling unit, an image pick-up device, a judging device and a sorting unit. The handling unit holds a prescribed number of nuclear fuel pellets in a line and rotates the same on their axes. The image pick-up device is disposed adjacent to the handling unit, and picks up image data as to the peripheral surfaces of the nuclear fuel pellets. The judging device is operably connected to the image pick-up device, and analyzes the image data outputted from the image pick-up device to output judging signals. The sorting unit is operably connected to the judging device and separates defective pellets from non-defective pellets based on the judging signals. The sorting unit includes a plurality of sorting members and actuators. The sorting members are disposed adjacent to the handling unit so as to correspond to the nuclear fuel pellets, respectively.

Abstract: Gamma radiation leaving the surface of a radioactive materials storage vessel is measured and monitored using a plurality of gamma ray absorbing calorimeters. The gamma ray energy absorbed is converted into heat and such heat is measured or indicated for the purpose of monitoring expected radioactive decay processes of the stored radioactive materials. Material leakage or removal from storage is detected and defined by deviations from the normal expected decay in gamma radiation. Such deviation or lack of deviation, as the case may be, provide the desired accountability of the radioactive materials. Since there can be several bands of gamma ray radiation, both flux and its energy level may be determined by using different gamma absorbing materials in the calorimeters selected for the known possible energy levels of the decaying radioactive material.

Abstract: In combination with a steam generator having a plurality of generator tube support plates, each generator tube support plate having a plurality of openings and a plurality of generator tubes, each tube passing through aligned openings in the support plates, a corrosion monitoring system is provided including a mockup probe, comprising a probe tube support plate having an upper side and a lower side and having substantially the same thickness and being constructed of substantially the same material as the generator tube support plates, having at least one opening of substantially the same size and shape as the openings of the generator tube support plates; at least one probe tube having an upper end and a lower end and having substantially the same diameter as the generator tubes and being constructed of substantially the same material as the generator tubes, the probe tube passing through the opening of the probe tube support plate; and wherein the mockup probe is adapted such that it may be inserted and seal

Abstract: A pellet turning apparatus for facilitating surface inspection of nuclear fuel pellets includes a pellet turner assembly and a pellet supply tray interface assembly. The pellet turner assembly includes a pellet turner deck and a swivel mechanism supporting the deck. The pellet turner deck is composed of a base frame mounted on the swivel mechanism, a plurality of elongated laterally-spaced pellet channel-defining members extending between and mounted at the opposite ends of the frame, a pellet rolling plate supported on the frame below the pellet channel-defining members, and an actuating mechanism coupled to the pellet rolling plate for moving the plate in oscillatory direction substantially perpendicular to the pellet channels for producing in situ turning of the pellets. The interface assembly is mounted to a pellet loading end of the pellet turner deck.

Abstract: A method for detecting changes in the rate of heat transfer through a tube wall due to the deposition or removal of deposits formed thereon. The method involves obtaining a segment of tubing, placing a highly conductive material in contact with one of the inner surface and outer surface of the tubing wall, contacting the other of the inner and outer surface of the tubing wall with a heating or cooling medium, and measuring the time required for the temperature of the thermally conductive material to change between two predetermined temperature measurement points. Subsequently, the wall of the tubing segment is cleaned or the tubing segment is replaced with a different tubing segment, and the rate of temperature change of the tube walls is tested again in a similar manner. By comparing the times measured in the first and second trials, the difference in the rate of heat transfer before and after cleaning of the tubing segment can be determined.

Abstract: An apparatus and method for monitoring corrosion to members within the core of a nuclear reactor, particularly fuel rod cladding. A sensor means is submerged inside the core of a nuclear reactor near the member or fuel rods. The sensor means is comprised of a generally cylindrical section having an outer surface that is subject to corrosion and radiation, and has a cross-sectional area A.sub.1. The sensor means additionally has a reference section subjected to radiation but not to corrosion, and having a cross-sectional area A.sub.2. At least one pair of first probes, separated by a length L.sub.1, is placed in electrical contact with the cylindrical section. At least one pair of second probes separated by a length L.sub.2, is placed in electrical contact with the reference section. A current is passed throughout the sensor means to produce a potential gradient in the cylindrical section and reference section.

Abstract: An apparatus for inspection of a reactor vessel includes trolley supporting an axillay extending support mast and extension mast. The support mast has a tilting device to adjust the angle between the support mast and vessel wall to allow the extension mast to be accessible to lower portions of the reactor vessel. The lower part of the extension mast is connected to a rack having a horizontal trolley capable of tangential movement relative to the reactor vessel wall. The horizontal trolley also is connected to a probe position trolley enabling a probe holder to move tangentially relative to the horizontal trolley. The apparatus enables the probes to move radially and axially for greater access to positions of the reactor vessel wall.

Abstract: An apparatus and method are disclosed for determining the integrity of coated fuel particles via acoustic characterization. Particle supply means (4) drops the nuclear fuel particles (2) individually onto the face (6) of a transducer (8). The generated signal which is indicative of the integrity of the coating on the dropped particle is analyzed with analyzing means (26, 28, 30) for discriminating between a flawed and an unflawed particle.

Abstract: A foreign matter recovering apparatus includes a body for approaching a fuel assembly, a body fixing section for fixing the body to the fuel assembly for positioning, a moving mechanism section movable relative to the body, a recovering working unit adapted to be moved by the moving mechanism section to gain access to clearance of the fuel assembly to allow foreign matter to be recovered thereby, and a remote control section for remotely controlling the working unit on the basis of an image representing a working state of the working unit to operate it properly.

Abstract: A robot arm apparatus is provided for inspecting and/or maintaining an interior of a steam generator which has an outside wall and a port for accessing the interior of the steam generator. The robot arm apparatus includes a flexible movable conduit for conveying inspection and/or maintenance apparatus from outside the steam generator to the interior of the steam generator. The flexible conduit has a terminal working end which is translated into and around the interior of the steam generator. Three motors located outside the steam generator are employed for moving the terminal working end inside the steam generator in "x", "y", and "z" directions, respectively. Commonly conducted inspection and maintenance operations include visual inspection for damaged areas, water jet lancing for cleaning sludge deposits, core boring for obtaining sludge deposits, and scrubbing of internal parts.