Abstract: Water testing apparatus 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 apparatus and comprises filtering and a preconditioning of selected sample specimens of the coolant water with a pH adjustment for inhibiting unbalancing an equilibrium in solution of water coolant with ion solutes.

Abstract: An automated system for examining incore housing welds in a nuclear reactor from above is disclosed. The probe uses both ultrasonic and eddy current non-destructive inspection coils to examine the weld for indications, or flaws. It is inserted into the incore housing tube from above because of the high radiation exposure workers experienced using the prior method of inserting a probe into the incore housing from below. This is conveniently done when the incore flux monitors are removed during standard maintenance, and is performed from the refueling bridge.

Abstract: The invention makes it possible to measure the degree of radioactive radiation emitted by large pieces. It includes two large gamma ray detectors (12A, 12B) placed opposite each other and whose spacing is adjustable. They are completed by a photoelectron multiplier (14A, 14B). The piece (2) to be measured is suspended so as to be brought between the two detectors (12A, 12B). The positioning of these detectors is obtained with the aid of a computer and a system (22) for recognizing the shape of the piece (2) functioning with the aid of optical clip-on lenses. The fact that the two detectors are position-adjustable makes it possible to take account of the shape of the piece to be measured. An application is the recovery of materials originating from nuclear installations with the view to re-using the materials.

Abstract: A container for electrochemical reactants for an electrode is provided, as well as electrodes employing the same. The electrodes can withstand the rigorous environment of a nuclear reactor core. The design of the container reduces stresses in the electrode and significantly improves its reliability. The container has a generally cylindrical member of an insulator. The member has a base region with a cut-out extending therethrough. Sidewall means extend from the base to define a cavity for containing electrochemical reactants, e.g., silver chloride. An insert of an insulator at one end mates with the cut-out and at another end mates with and is brazed to a cap for the container. The insert has a second sidewall means for holding the electrochemical reactants, and is shaped to fit within the first sidewall means. The base of the insert and the the cut-out are metallized for acceptance of a silver base.

Abstract: The system of the present invention uses the importance of a piece of malfunctioning equipment, the severity and the confidence level in a diagnosis to determine repair priority of the equipment by obtaining the product (46) of the confidence level (CF), importance (IMP) and severity (SEV). The these variables are determined by expert system rules. The severity is the reciprocal (88) of the time to failure. The importance (44) is the cost to repair the maximum damage when the malfunction continues. A malfunction can affect several pieces of equipment in combination, the severity and importance associated with each piece is combined (70) with the confidence and used to determine the repair priority. When the diagnosis of a malfunction is by malfunctioning sensors, the expected equipment life (40) and the availability (86) of sensors that provide a partial backup are considered in prioritizing sensor as well as equipment repair.

Abstract: Measurement values of conductivity of pH of water in the primary cooling water system of a boiling water atomic power plant are input through on-line and the measurement values of conductivity and pH at times of measurement when water quality changes are compared with values at the time of normal operation. Tendencies to increase or decrease the conductivity and pH are combined to prepare one diagnosis data, and causes of anomaly in progress in component apparatus members are discriminated and diagnosed by the diagnosis data. By using a correlation between logarithm of conductivity and pH in preparing the diagnosis data from combinations of conductivity and pH, causes of anomaly can be diagnosed more accurated.

Abstract: A device for non-destructive testing of an upright cylindrical nuclear reactor pressure vessel in which there is disposed a coaxially arranged likewise cylindrical nuclear fuel container. A first upper annular rail is fastened to an upper edge of the pressure vessel. A second lower annular rail is fastened to an upper edge of the nuclear fuel container. A bridge element is displaceably mounted on the first annular rail. A vertical, hollow mast projects into the pressure vessel and has an upper end mounted at the bridge and a lower end movably mounted to said second annular rail so that the mast is movable in a circumferential direction of the pressure vessel. A deflectable belt is disposed for longitudinal displacement through the hollow mast. The belt exits the hollow mast at a lower region of the mast and has an end extendable into an annular space between an inner wall surface of the pressure vessel and an outer wall surface of the nuclear fuel container.

Abstract: A unique and optimal method for assessing the feasibility of employing chemical decontamination processes to remove crud from nuclear reactor primary systems without adversely affecting the integrity of the system components or piping during further reactor operation is disclosed. Materials used to construct such components and piping are exposed to several cycles of simulated decontamination processing with a test loop constructed to simulate a range of full system decontamination process conditions. The material specimens are then removed and tested to assure that the particular decontamination processing system has no adverse effects on the materials used.

Abstract: The distance between substantially parallel metal tubes forming a flat sheet and the thicknesses of oxide layers covering the tubes are simultaneously measured. The tubes may particularly be sheaths of a nuclear fuel assembly.

Abstract: In a system for detecting abnormal movement of a part of a gas-cooled nuclear reactor, for example abnormal tilting of a boiler unit caused by failure of a boiler support, a tracer material is released into the gas coolant path in response to the movement. The releasing of the material may be effected by causing a gas canister to be pierced as a result of the movement. The presence of the tracer material in the coolant is detected by the coolant monitoring system.

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 of 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: A thimble tube for use in a guide the in a nuclear reactor. The thimble tube is an elongated hollow tube having one end closed. The portion of the tube that extends into the reactor is formed from a plurality of sections attached to each other to form the tube in an alternating fashion along the width and length of the tube. The sections are formed from materials having different coefficients of thermal expansion. This causes the thimble tube to warp into supporting contact with the interior of the guide tube at normal reactor operating temperature.

Abstract: An improved apparatus and method for ultrasonic inspection of materials through barriers such as gaps in manufactured parts is disclosed. The improvement herein is directed to enabling such ultrasonic testing to bridge ambient gaps such as intentionally formed gaps in composite structures having a first structure for originally receiving and transmitting sound separated by the gap from another structure to be inspected. Preferably, the gap is flooded with a gas having a predictable and optimum speed of sound relative to the material of the first and second structures. Sound is propagated to the first structure in a wave packet that is transmitted through the couplant fluid. The sound is generated in a wave packet having a spatial width at least twice the dimension of the gap to be bridged. The wave packet has a contained frequency having a wavelength (relative to the speed of sound of the gas flooding the gap) to create a constructively interfering standing wave node within the gap.

Abstract: A container for electrochemical reactants for an electrode is provided, as well as electrodes employing the same. The electrodes can withstand the rigorous environment of a nuclear reactor core. The design of the container reduces stresses in the electrode and significantly improves its reliability. The container has a generally cylindrical member of an insulator. The member has a base region whose outer surface is a surface attachment region. The base region has an annular channel for a lead. The base region also has a generally conical cut-out. Sidewall means extend from said base to define a cavity for containing electrochemical reactants, e.g., silver chloride. A generally conical insert of an insulator mates with the cut-out. The outer surface of the insert, cut-out, and surface attachment region is metallized, and the insert is set in place by brazing. In the electrode the lead is in electrical contact with the apex of the insert.

Abstract: A vehicle of the invention comprises at least two transfer arms including means for causing expansible positioning fingers to penetrate into the tubes of a steam generator and for extracting said fingers from said tubes. The vehicle also includes at least two pivoting fastening heads which contain said expansible positioning fingers. The heads are interconnected by one of the telescopic transfer arms, which arms extend perpendicularly to said pivoting heads. The other telescopic transfer arm is fixed to one of the heads and it extends perpendicularly therefrom with its opposite end carrying a tool support. Such a vehicle is particularly suitable for inspecting and maintaining tubes in tube apparatuses, e.g. the steam generators of pressurized water nuclear reactors.

Abstract: A system and method for determining the condition of one or more parameters of a valve, e.g., particularly a check valve in a fluid carrying pipe, by monitoring fluid turbulence downstream of the valve. An ultrasonic transmitter-receiver transducer pair provides a signal modulated by the fluid turbulence, which signal is demodulated and processed to obtain a signature or other signal indicative of the valve condition. The processes signature signal may additionally be compared with a measure of flow rate determined just upstream from the valve. By operating the ultrasonic system at high frequency and using appropriate detecting circuitry, low frequency interference from mechanical shocks and vibrations is avoided, yielding a reliable detection of the flow turbulence caused by the valve.

Abstract: An apparatus and method to facilitate the in situ detection of small leaks in fluid-filled, capacitive type pressure transducers by remotely and non-intrusively inducing a brief time varying transient signal, and then analyzing the frequency response of the transducer to such excitation in order to detect the presence and effects of a fluid leak. The adverse effect of this induced transient signal on the process control circuitry is insignificant, while providing for an accurate and sensitive method for detecting the extent of fluid loss in the transducer when evaluated according to the present invention. As a consequence of such transducer evaluation techniques, frequency responsive characteristics of such devices may be recorded and self-correlated over long periods of time.

Abstract: Disclosed is a device for measuring dimensional characteristics of an elongate component which comprises:(a) an elongate rigid support that bears (i) an upper platform upon which is mounted a drive; (ii) a clamp mechanism for securing said component about its proximal end in a reference position; (iii) a lower platform which supports a pedestal for securing said component about its distal end in a reference position; and(b) a sensor carriage movably mounted to said rigid support and operatively connected to said drive for movement along the elongate extent of said support and having an opening penetrating therethrough for receiving said elongate component when it is secured by and substantially parallel to said rigid support, said carrier bearing a plurality of pivotally-mounted bell cranks spaced about said opening, each bell crank having a wheel for riding on said component during movement of the carrier and being in contact with a sensor capable of providing a signal correlative to displacement of the bell

Abstract: The shutoff device consists of a housing (14) inserted in the conduit, in which a movable shutter (32) is mounted for oscillation about a horizontal axis (33) in the housing (14) and comprising a closing member (35) returned into a position of closing a channel (28) by a counterweight (36). A manometer (43) enables the pressure inside the housing (14) to be measured. The shutoff device, wholly passive in operation, can be employed in particular in a conduit for guiding a measurement probe employed in a pressurized water nuclear reactor.

Abstract: A method and apparatus for inspecting the change in a water quality in terms of the change in a corrosion rate by detecting the capacity change between opposed electrodes due to the corrosion of electrode surfaces. As a result, the water quality can be continuously monitored over a long time while leaving the electrodes in the water to be inspected. Thus, the method and apparatus are suited for controlling the water quality especially in an atomic reactor vessel.

Abstract: An apparatus is disposed below a tube plate in a chamber of a steam generator which is accessible through a manhole, for examining and/or repairing steam-generator tubes in the tube plate. A tool or inspection-instrument carrier of the apparatus includes a pneumatic swivel cylinder detachably connected to the apparatus. The swivel cylinder has a housing and an output shaft. A feed unit is supported on the output shaft. A flexible shaft is axially movable by the feed unit. A compressed-air motor is associated with the housing for driving the flexible shaft.

Abstract: Nuclear fuel cladding tube end plug welds are ultrasonically scanned to develop data indicative of weld thickness, weld penetration, and the existence and size of any included feature, such as pores. A computer analyzes the data to determine critical weld characteristics which are then tested against established quality assurance standards to see if the end plug weld is acceptable, all on an automated production line basis.

Abstract: A system for measuring the power distribution in a reactor core includes a plurality of gamma detectors disposed outside of the reactor vessel which houses the reactor core. The gamma detectors correspond in number to the number of instrumentation tubes in the fuel assemblies in the reactor core to be sampled. A first plurality of separate flow branches connect the instrumentation tubes with the gamma detectors for routing the flow of coolant from the instrumentation tubes to the gamma detectors so that the gamma detectors can measure the .sup.16 N decay gamma activity in the respective instrumentation tubes of the fuel assemblies. A second plurality of separate flow branches connect a cold leg of the reactor coolant system with the gamma detectors for routing the flow of coolant from the cold leg to the gamma detectors so that the gamma detectors can measure the background gamma activity in the respective instrumentation tubes of the fuel assemblies.

Abstract: Disclosed is a system for driving inspection probes through steam generator tubes of a nuclear power plant, particularly eddy current probes. The system utilizes a single conduit that is connectable to the tube sheet of a steam generator and comprises a multiple-unit probe pusher having a housing containing a plurality of independently actuable drive paths, a multiple-unit probe storage device including a plurality of independently movable take-up reels, and a guide or deflector device for directing any one probe that is driven from any one of the drive paths of the probe pusher to the single conduit. Preferably, the entire system is also remotely controllable. The system enables one probe to be in use for tube inspection while providing readily available spare probes to replace a defective probe with a minimum of downtime and worker radiation exposure.

Abstract: System for inspecting the interior of tube plugs, particularly tube plugs of the type having expander elements disposed therein. The system includes a probe assembly connectable to the tube plug, the probe assembly having a rotatable and translatable probe carrier housing for housing a sensor probe capable of inspecting the tube plug. The probe assembly also includes a leaf spring extending through the probe carrier housing and attached to the sensor probe for radially outwardly biasing the sensor probe into sensing contact with the interior of the tube plug, particularly the interior of the tube plug located between the top of the tube plug and the expander element. A tangle-free flexible hose is connected at one end thereof to the probe carrier housing for transversely rotating and longitudinally translating the probe carrier housing. The flexible hose is also connected at the other end thereof to a probe driver assembly that rotates the flexible hose which in turn rotates the probe carrier housing.

Abstract: A light water nuclear reactor test assembly for emulating the performance of a nuclear boiling water reactor fuel bundle and measuring the subchannel void fraction in the two-phase region of the emulated fuel bundle by gamma ray attenuation is disclosed. The emulated fuel bundle has individual emulated nuclear fuel rods typically heated by individual electrical currents instead of nuclear reaction and are cylindrically hollow. A gamma-emitting source is placed on a probe and mounted for vertical excursion inside a selected emulated hollow fuel rod. A detector, typically a Geiger-Muller counter, is placed for corresponding vertical excursion inside another and preferably adjacent fuel rod. Gamma radiation from the source to the detector through the walls of the emulated fuel rods is measured. Preferably, both detector and source are collimated so that the detector does not receive gamma ray scattering from the interior of the test assembly.

Abstract: A method and apparatus for providing a permanent identification of reactor fuel assemblies with an in-reactor remote readout capability. A bar code type code assemblage is permanently coupled with the upper portion of the fuel assembly which is comprised of alternating components of magnetic and non-magnetic metal such as stainless steel. An active sensor including a magnetic core, excitation winding, and sensor or pick-up winding is employed to interact with the individual code components of the bar code assembly. The latter assembly may be formed as a sequence of rings through which the sensor is passed during a grappling procedure or as a sequence of vertically-oriented bar segments positioned outwardly of the fuel assembly and readable by rotating the fuel assembly before the active sensor.

Abstract: A detection unit (20) is moved under water inside the fuel storage pond (15), over the entire height and in the vicinity of a wall of the cell (2). The detection unit (20) comprises at least one probe for monitoring the thickness of the material. The probe readings are collected from a region (22) situated above the pond (15) and these readings are compared with reference measurements. The monitoring process makes it possible to verify the presence of the neutron-absorbing material in the side walls (10) of the cell (2).

Abstract: A detector path insertion verification system for use with flux mapping of a pressurized water nuclear reactor uses a transfer insertion switch at the input of each multiple-path selector to detect insertion of a detector into the multiple-path selector. No insertion switches are used at the output of the multiple-path selectors; instead software is used to verify the position of the multiple-path selectors. The sensed actual position of each multiple-path selector is monitored as the multiple-path selector is rotated to verify that the sensed actual position passes through the expected sequence of positions in an expected amount of time. The multiple-path selectors are rotated until a sensed actual position is equal to a desired position and rotation is stopped. After the passage of a short period of time, the actual and desired positions are again compared and if equal, insertion is begun.

Abstract: This system measures tritium on-line and collects tritium from a flowing inert gas stream. It separates the tritium from other non-hydrogen isotope contaminating gases, whether radioactive or not. The collecting portion of the system is constructed of various zirconium alloys called getters. These alloys adsorb tritium in any of its forms at one temperature and at a higher temperature release it as a gas. The system consists of four on-line getters and heaters, two ion chamber detectors, two collection getters, and two guard getters. When the incoming gas stream is valved through the on-line getters, 99.9% of it is adsorbed and the remainder continues to the guard getter where traces of tritium not collected earlier are adsorbed. The inert gas stream then exits the system to the decay chamber.Once the on-line getter has collected tritium for a predetermined time, it is valved off and the next on-line getter is valved on.

Abstract: Method and apparatus for carrying out the in situ inspection of the shaft of a pump. A spacer coupling or link is manually removed from a limited access area intermediate the pump drive and top of the pump shaft. Inspection then is carried out through a previously formed inspection bore extending concentrically along the interior of the shaft. A probe supported by an extension assembly is lowered into the inspection bore until characterizable portions of the shaft such as a labyrinth seal region are identified. A discrete small region of interest which will initially evidence any fault condition then is surveyed with a raster scan movement of the inspection probe. Lower components of the region of interest are also identified by a next adjacent characterizable region such as a fillet welding for a journal assembly. Because of the limited longitudinal extent of the region of interest, the raster type movement of the probe becomes available.

Abstract: A preferred embodiment of the present invention provides an incore instrumentation system for detecting conditions, e.g., neutron flux density, within the core of a pressurized water nuclear reactor (PWR). The instrumentation system includes a plurality of long straight thimbles (52), containing fixed incore detectors, which are inserted through the top closure head (18) of the PWR. The invention meets today's principal utility, EPRI, DOE and NRC requirements and preferences, including no bottom RV head penetrations and the use of fixed incore instrumentation. In addition, an instrumentation system according to the invention is mechanically simple with a straight core insertion/removal path that does not conflict with or add significant complexity to the reactor internals.

Abstract: A flux mapping system for a nuclear reactor uses a detector having an outer diameter approximately fifteen percent smaller than the inner diameter of distribution tubing through which it passes and a length less than eight percent of the minimum bend radius of the tubing. Frictionless limit switches are used in the flux mapping system to further reduce friction against cables which drive the detectors through the distribution tubing. In combination, these dimensions and switch changes result in a fifty percent improvement in driving force at the thimbles into the core of the nuclear reactor.

Abstract: An inspection system has a pellet diameter inspecting station and an apparatus for transferring pellets past the station. The apparatus includes a stationary support track for aligning and guiding pellets in a row along a linear path, and a transfer mechanism operable through forward and reverse strokes. The transfer mechanism has first and second grippers actuatable between pellet gripping and releasing positions and operable to grip respective rearmost pellets in a first row of pellets to be advanced past the inspecting station and in a second row of pellets previously advanced past the inspecting station. The transfer mechanism is operable to move from an initial position through the forward stroke to advance the pellets at a constant velocity along the linear path.

Abstract: A system for monitoring an operating condition of a control rod drive mechanism (CRDM) having a pressure housing erectly mounted on a reactor vessel head closure, a vertically movable control rod drive shaft accommodated in the housing and an electromagnetically motive latch mechanism includes an accelerometer fitted to a housing upper end and signal recording means for receiving output signals from the accelerometer through signal processing means and recording the processed output signals.

Abstract: A method and device for monitoring the power output of a nuclear reactor in the low power range, by detecting neutrons produced by the reactor and producing a representation of the number of neutrons detected during each of a succession of equal measuring intervals; and producing a power output level indication having a value associated with each measuring interval by a digital operation in which the indication value associated with a preceding time interval is altered according to a function of the representation produced during the detecting step for the present measuring interval.

Abstract: An apparatus and method for scanning the outer profile of control rods in a control rod assembly includes a plurality of rotatably mounted ultrasonic transducers which are mounted for rotation about the axis of at least some of the control rods in the assembly. The assembly is moved in a linear path parallel to the axis of the control rods, past the rotating transducers for scanning the surfaces of the control rods in helical paths. By distributing the transducers in a selected configuration, all control rods in the assembly can eventually be scanned.

Abstract: A sensor assembly comprising a rod-shaped thermally conductive element oriented radially with respect to a flow pipe and two or more thermocouples installed on the thermally conductive element. A first end of the thermally conductive element is arranged in contact with or in thermal communication with the outer periphery of the pipe, while a second end of the thermally conductive element is spaced from the pipe. As a result of heat conduction through the pipe to the first end, a temperature gradient is provided across the thermally conductive element. The temperature gradient exhibits a relatively high temperature near the first end and a relatively low temperature near the second end. The thermocouples are installed along the length of the thermally conductive element. One thermocouple is arranged nearer to the pipe than the other thermocouple. Temperature dependent signals obtained from the thermocouples are transmitted to suitable evaluation electronics.

Abstract: An apparatus is provided for determining the thermal history of an environment for purposes of monitoring thermal aging in equipment or the extent to which a process has proceeded. The apparatus comprises a plurality of solid state track recorders, and the range of activation energy values among the solid state track recorders encompasses the activation energy of the thermal aging process of the equipment to be monitored. Using Arrhenius functions, data obtained from the apparatus can be used to determine an equivalent average temperature which is combined with the equipment life data to determine the extent of thermal aging.

Abstract: A simple and accurate method for determining the enthalpy of wet steam using pressure and moisture content values. Moisture content is determined by measuring sonic velocity of the steam, which is dependent upon enthalpy. A small sample of the wet steam is superheated by pressure reduction to a single-phase state prior to measuring its sonic velocity in order to achieve greater accuracy of this measurement. From its sonic velocity, enthalpy of the sample is calculated. Since the process takes place at constant enthalpy, the enthalpy value of the sample is the same as that of the fluid from which it was drawn.

Abstract: A process is provided for the long term shutdown of a high temperature nuclear reactor comprised of a pile of spherical fuel elements in a core by means of a neutron absorbing absorber material wherein the absorber material consisting of spherical absorber elements is introduced in the core of spherical fuel elements in the form of a column, but wherein the intermixing of the absorber material with the fuel elements is prevented. An apparatus for practicing the process is also provided comprised of a graphite side reflector concentrically surrounding a circular cylindrical core filled with a pile of spherical fuel elements of a high temperature nuclear reactor, into which at least two nose shaped projections distributed uniformly about the circumference radially project, with each of the projections comprising a vertical cavity to contain the absorber material. The vertical cavity is located in the vicinity of the core in the area of the frontal side facing the core of each projection.

Abstract: The device comprises a flexible pipe (15) having one of its ends in a zone (20) situated outside the water box (6) and its other end joined to a tubular probe-guide (16) supported by a handling device (13, 14) inside the water box (6). A sheath (21) is slideably mounted inside the flexible pipe (15) and has mounted at one of its ends a stop and a bush for guiding the ultrasonic probe (18). A Bowden cable (22) is mounted rotatably about its axis inside the sheath (21), and is integral at one of its ends with a profiled rotational driving part and joined at its other end, outside the water box (6), to a motor (24). The support for the probe (18) has a threaded part engaged in a tapped part of the guide bush and can be driven in rotation by a profiled part integral with the end of the Bowden cable (22). The support for the probe (18) is thus driven in rotation and axial translation so as to perform a helical movement.

Abstract: An apparatus for the inspection and/or repair of tubes discharging into a chamber of a heat exchanger includes a tube sensor. A manipulator provides remote-controlled positioning of the tube sensor. A plurality of mouthpieces disposed on the manipulator introduces the sensor into tubes of a heat exchanger. Tube connectors communicate with the mouthpieces and have two ends. One of the ends of the tube connectors is coaxially attachable to orifices of the tubes of the heat exchanger. The other of the ends of the tube connectors are bundled into a circular group at an adaptor. A distributor plate has orifices into each of which a respective one of the tube connectors discharge. A parallel turntable is supported on the distributor plate. A guide hose is introduced into the turntable for guiding the tube sensor. Incremental rotation of the turntable selectively aligns the axis of the guide hose with the axis of one of the orifices in the distributor plate.

Abstract: A system for helically driving an inspection probe within a tube of a nuclear steam generator is provided, wherein the system is particularly useful in driving a combination eddy-current and ultrasonic probe. The system generally includes a head assembly insertable with the tube for inspection, a drive frame assembly to be located external of the steam generator, and a conduit system connecting the head assembly and drive frame assembly. Particularly, the head assembly includes a portion fixable to the tube by a pressure bladder and a movable portion helically movable with a probe to the fixable portion. The drive motor and electrical pick-up are slidably movable on the drive frame assembly to move linearly with the probe during inspection so as not to limit the stroke of the probe, wherein the drive motor and pick-up are connected with the head assembly by way of two flexible coaxial conduits.

Abstract: A drive unit for gripping and rotating a fuel rod during inspection of the fuel rod. A housing has a drive motor mounted at its first end with a spindle rotatably mounted in the housing and directly coupled to the drive motor. A recess at one end of the spindle adjacent the second end of the housing receives a substantially donut shaped rubber bladder. A flanged port that extends radially from the bladder and is in fluid communication with the interior of the bladder provides a means for pressurizing the bladder. This causes radial inward expansion of the bladder for gripping a fuel rod to be inspected.

Abstract: A communications interface and system is configured to allow circumvention recovery of control after radiation) of a microprocessor portion thereof. The communications interface portion, including a hard memory interface device serves as a communications link to the microprocessor. The system is configured so that a transient upset from prompt gamma radiation is detected and a circuit freeze of the stored information in the system is carried out until the transient subsides. The system then restores the microprocessor to the preirradiation condition and resumes operation. Accordingly, the hard memory interface device of the communications interface, controls a READ operation and CLEAR and WRITE operations in addition to generating signals to control the "handshaking" between the hard memory of the communications interface and the aforementioned microprocessor.

Abstract: By providing a yoke (36) to the shoulder (15) of a robot for inspecting and repairing the tubes of a nuclear boiler, which yoke can receive an expandable anchoring piece (35) of an auxiliary tool (23) connected to the wrist (18) of the robot, said auxiliary tool (23) being provided with additional span fingers (37, 38), it is no longer a time-consuming and therefore expensive procedure to also inspect the tubes that lead to the tube sheet passages (9) into which the usual span fingers (8) of the robot were originally fixed.

Abstract: A switch for reacting quickly to a neutron emission. A rod consisting of fissionable material is located inside a vacuum tight body. An adjustable contact is located coaxially at an adjustable distance from one end of the rod. Electrical leads are connected to the rod and to the adjustable contact. With a vacuum drawn inside the body, a neutron bombardment striking the rod causes it to heat and expand longitudinally until it comes into contact with the adjustable contact. This circuit closing occurs within a period of a few microseconds.

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: A non-invasive system and method is provided for the inspection of check valves which combines acoustic and magnetic techniques. The acoustic and magnetic techniques are used to measure check valve disk motion and internal impacts or vibrations resulting from the disk motion.