Abstract: A method and apparatus for testing characteristics of nuclear fuel pellet interaction with nuclear reactor fuel rod cladding whereby the temperature profile within a nuclear fuel rod is more closely modeled in the test device than in the prior art discloses the use of high frequency microwave radiation to heat the fuel pellets within the cladding and using the cladding itself as a waveguide. Sensors monitor various operating parameters and a water jacket cools the test cladding.

Abstract: A portable, microprocessor-based test instrument for the calibration, check testing, and checkout of nuclear power reactor systems under simulated operating conditions. The instrument includes a master microprocessor bus having direct communication to interchangeable plug-in printed circuit boards. Applications to which the invention may be put include rod control simulation and analysis, neutron monitoring system analysis, and rod block monitoring system simulation and analysis. The test instrument electronically interconnects at any of a plurality of points within a reactor control and monitoring system. The electrical network from the point of interconnection to the core rods is disabled and the test instrument, receives serial control command signals as would the disabled portion of the network. The test instrument also generates appropriate acknowledgements and response signals. Alternatively, a test instrument can generate faulty response signals to simulate improper command responses.

Abstract: An improved model steam generator for simulating the conditions within a nuclear steam generator in order to monitor the condition of the heat exchange tubes and tubesheet of the nuclear steam generator is disclosed herein. The improved model steam generator includes a highly effective separator assembly for separating water droplets entrained within the steam flowing out of the outlet of the secondary side of the generator formed from a plurality of separator grids, each of which includes an array of semi-cylindrical deflector members. The grids are vertically stacked with the deflector members transversely disposed to the flow of steam generated by the model steam generator. Each of the parallel arrays of deflector members in each grid is angularly disposed approximately 45.degree. to the deflector members in the grids above it and below in order to provide a tortuous path for the flow of steam ascending therethrough.

Abstract: An apparatus for inspecting fuel assemblies, includes a water container, a receptacle disposed in the container for securing bases of fuel assemblies vertically disposed in said container, a beam, a plurality of fingers engaged by the beam, measuring probes supported on the fingers, and a guiding device disposed in the container and accurately positioned relative to the receptacle for guiding the beam laterally toward the fuel assemblies for inserting the measuring probes in a given direction into spaces between cladding tubes of fuel rods of the fuel assemblies, each of the fingers being movable relative to the given insertion direction after striking an obstacle.

Abstract: An electrode (20) for electrochemical measurements in aqueous solutions at high temperatures, especially measurements in water in or from a primary or secondary circuit of a nuclear reactor in operation at a temperature of the water substantially prevailing at that time, comprises a conductor (21) of a platinum-group metal which for part of its extension is enclosed within a casing (22) making tight contact with the conductor and consisting of sintered-together particles of aluminum oxide, zirconium oxide or other electrically-insulating material which is inert to water. The conductor has one part (21a) exposed the aqueous solution and one connection part (21b) for connection to an electrical measuring member.

Abstract: Disclosed is an alpha monitor usable in an automated nuclear fuel pin loading and processing unit. Fuel pins or other elongated pins are fed laterally into the alpha monitor in a singular fashion and are translated by a first roller assembly into a weld flare machining and decontamination substation not forming a part of the invention. Pins return and are lifted upwardly and transferred across to a combined pin lifting and electrode operating means which lifts the pins upwardly into a clamshell electrode assembly which is spread open by a combined pin lifting and electrode operating means. Once inserted the clamshell type electrode arrangement closes around the fuel pins so that inspection can occur. Fuel pins are inspected by charging electrodes to a negative potential and measuring the change in charge occurring when positively charged alpha particles strike the negatively charged electrodes.

Abstract: Device for locating the position of a nuclear reactor control rod which is suspended from a shaft (12) sliding vertically inside an adaptor tube (10, 11) under the effect of a screw-nut mechanism which comprises a nut-forming rotor (14) mounted to rotate around the vertical axis of the tube (10) and provided with a threaded bore in which is engaged a corresponding threaded part (13) of the shaft (12).

Abstract: A non-destructive method for determination of neutron exposure and constituent concentrations in an object, such as reactor pressure vessel, is based on the observation of characteristic gamma-rays emitted by activation products in the object by using a unique continuous gamma-ray spectrometer. The spectrometer views the object through appropriate collimators to determine the absolute emission rate of these characteristic gamma-rays, thereby ascertaining the absolute activity of given activation products in the object. These data can then be used to deduce the spatial and angular dependence of neutron exposure or the spatial constituent concentration at regions of interest within the object.

Abstract: A heat pipe stablized specimen container for irradiation of specimens at substantially constant temperature within a liquid metal cooled fast reactor comprises a heat pipe containing a vaporizable substance such as sodium. The container is of double-walled construction with the gap filled with argon and at top of the container a volume of argon is trapped within a cavity of the liquid metal level within the container in such a way that retention of argon in this zone is not dependent on sealing welds in the structure of the container, the argon blanket in this zone affording thermal insulation at the top of the container and also around part of the heat pipe, viz an adiabatic section of the latter.

Abstract: A system for monitoring a limited number of operating conditions in a nuclear reactor and for graphically displaying same on a display device is disclosed. The system parameters selected are those which typically change as a result of a reactor trip. As long as these system parameters remain within a "post-trip window" on the graphic display, the power plant is operating normally. If, however, any of the foregoing system parameters falls outside this "window", this is an immediate indication to the power plant operator to institute the necessary procedures to remedy same. A sequential controller (26) is provided to continuously sequentially sample the foregoing operating parameters and to transmit data regarding same to the display device.

Abstract: Compression loading on pin 14 is determined by placing hydraulic jack 20 adjacent thereto. A voltage 40 is imposed between the base member 12 and the load member 10. Current flowing through the pin is detected 42. Hydraulic pressure in the jack is increased 28 until current flow through the pin stops. The pressure level at this point is determined 34.

Abstract: Method of monitoring a nuclear power station by gathering information, called parameters, relating to the operation and state of the power station and its auxiliary equipment, comparing these parameters with reference data corresponding to normal operation, and signaling the differences, in order to permit an operator to evaluate the state of the power station very quickly, and to undertake any needed steps promptly.

Abstract: A device for positioning a member so as to face each of a plurality of perforations arranged in a given grid in a plate, especially for the inspection of the tubes of a steam generator in a nuclear power station. It comprises a body having at least two attachment members for fastening the device in the perforations. The device is rotatable around one or the other of these attachment members so as to displace it on the plate, and has a support carrying the member and pivotable with respect to the body in order to bring the member into a position facing several perforations without displacing the device.

Abstract: In a device for the determination of the vibrations occurring at the internals of a reactor pressure vessel, it is possible to perform the maintenance of the vibration detector located in a housing from outside the reactor pressure vessel. One end of a guide tube is fed through a pipe stub to a receptacle cone connected to the core internals. The end of the guide tube facing away from the reactor pressure vessel is cup-shaped and is sealed with a cover. A flexible rod is connected to the housing and its other end extends into the cup-shaped end of the guide tube. The flexible rod and with it the housing connected to it can be inserted into the receptacle cone, providing good transmission of the vibrations prevailing at the internals, which reach an evaluation unit through the cable.

Abstract: The counting system includes a portable gamma ray detector having a detecting portion electrically connected to instrumentation for analyzing the radioactivity of nuclear reactor fluids, a radiation shielded panel including a plurality of openings for receiving the detecting portion of the detector, and sample vessels for receiving a radioactive fluid located behind the panel. Samples of radioactive fluids such as reactor coolant, reactor coolant off-gas and containment air, are selectively introduced into the sample vessels. The reactor coolant is cooled, de-pressurized and de-gassed and the off-gas and the liquid are routed to separate sampling vessels. The radioactivity of the radioactive fluid contained in each sample vessel is determined by simply inserting the detecting portion of the detector into a corresponding panel opening. Radiation shielded plugs are inserted into each of unused panel openings to shield the operator against radiation.

Abstract: A double-cylinder selector valve for detecting and locating failed nuclear fuel assemblies including a fixed outer cylinder provided on the circumferential wall surface thereof with a plurality of outer sampling holes arranged in a grid in both the longitudinal and circumferential directions, and a movable inner cylinder disposed rotatably within the fixed outer cylinder and provided on the circumferential wall surface thereof with a plurality of inner sampling holes. Each of coolant sampling pipes extending from the coolant outlets of each of a plurality of nuclear fuel assemblies is connected to one of the outer sampling holes. The inner sampling holes communicate with some of the plurality of outer sampling holes in the fixed outer cylinder at one position of the movable inner cylinder and with others of the outer sampling holes at other positions of the movable inner cylinder.

Abstract: An apparatus and method for supporting plural radioactive sources in operative gauging relationship at different elevations in a vessel containing material and density and/or level of which is to be monitored, and for transporting the radioactive sources to a radiation-proof source holder for storage exteriorly of the vessel when the vessel is being cleaned or the like.In one embodiment the sources are independently supported on a continuous basis on separate cables and transported between the vessel and the source holder by separate winches associated with the source holder, with each winch independently controlling the transport of its respectively associated source.

Abstract: A system for monitoring and displaying those operating parameters of a nuclear reactor which pertain to the core power distribution is disclosed. An incore system of detectors (24) provides power peaking data to a power peaking margin display system (16) which determines the operational limit line for the reactor core. The positions of the axial power shaping rods and the regulating rods are determined and are shown graphically by means of a cursor on a visual display (18) on which is superimposed the operational limit line determined by the power peaking margin system (16).

Abstract: Two parallel heat exchange tubes having that part of a support structure between them to be interrogated for defects. A pair of radiation sources are positioned in the one tube to straddle the interrogated support portion, and a radiographic film is positioned in the tube behind the interrogated portion.

Abstract: A method and system for quality control of nuclear fuel materials, said system including amplifier/integrator circuitry, signal processors, and drift control feedback circuitry. The invention permits the qualitative and quantitative measurement of paramagnetic additives such as gadolinium for example in a paramagnetic base nuclear fuel material including ferromagnetic impurities. Drift problems relating to temperature and other effects are solved electrically and mechanically. Analysis can be conducted irrespective of the form or composition of the nuclear fuel.

Abstract: A method of and apparatus for heating test specimens to desired elevated temperatures for irradiation by a high energy neutron source. A furnace assembly is provided for heating two separate groups of specimens to substantially different, elevated, isothermal temperatures in a high vacuum environment while positioning the two specimen groups symmetrically at equivalent neutron irradiating positions.

Abstract: The invention relates to a method of protection of a pressurized-water nuclear reactor, which enables the crisis of boiling and the melting of fuel to be avoided. An emergency shutdown is triggered when a parameter representative of the crisis of boiling or the power released per unit length of the fuel elements exceeds a limiting value. For the determination of these parameters, the most exact representation possible of the distribution of the power in the core of the reactor is employed. This distribution is obtained by measurement of the axial power by ionization chambers, and determination of the radial power by measurement of the positions of the control rods.

Abstract: The operation of a nuclear fueled, electric power generating unit is monitored by utilizing decision tree analysis of selected system parameters to generate a representation of the real time system status. Current system status is presented to the operator either in the form of conclusory indications, such as verbal statements, representative of system status, or in the form of a visual display of the entire decision tree including indications of the path through the tree representative of current conditions. In either case, indications of off-normal conditions are accompanied by directions as to appropriate action to be taken, either by the operator or the automatic control system, to steer the system toward more acceptable conditions. In addition, the status indications are prioritized to indicate the seriousness of the off-normal conditions and to direct the sequence of corrective action to be taken when multiple off-normal conditions exist.

Abstract: The radioactivity of radioactive fluid is monitored, in the presence of background radiation, without interruption of radioactivity detection with two successive volumes of radioactive fluid, the background being eliminated by subtraction. Typically the intrinsic responses are in the ratio of two to one for such successive measurements, when the proper volumes have been selected.

Abstract: Constant or transient forces and displacements between a motion snubber and its associated load are detected and converted to an electrical signal. Electrical signals from the detection devices are transmitted to a receiving device which is electrically connected to a recording means. The signals from a plurality of detecting devices are encoded to identify the snubber having the detected force and displacement.

Abstract: An apparatus for measuring the concentration of radioactivity in a radioactive fluid discharged from a nuclear power plant. The apparatus comprises a radioactivity concentration measurement pipe arranged to extend vertically, fluid supplying means for supplying the fluid into the measurement pipe in such a manner that the fluid flows down through the measurement pipe along the axis of the measurement pipe without contacting the inner surface of the latter, radioactivity concentration measuring means disposed at the outside of the measurement pipe for measuring the radioactivity concentration in the fluid flowing in the measurement pipe, and means for discharging the radioactive fluid to the outside of the measurement pipe. The apparatus further comprises a gas supply pipe opening at its one end to the atmosphere and communicated at its other end with the inside of the measurement pipe, and a check valve disposed in the gas supply pipe and operated at a pressure higher than the atmospheric pressure.

Abstract: A sampling system in which carriers (1) containing sample bottles (4) are transported by pneumatic conveyor means (3) between an examination station, for example a laboratory, and a selected sampling station at a plant. On arrival at a sampling station (2) a bottle (4) is moved through a predetermined sequence of movements by a transfer arm (5) to cause the bottle to be removed from its carrier and placed over a sample dispenser (7) to introduce sample into the bottle. The bottle containing sample is then returned to the carrier for return to the examination station.

Abstract: The tubes of tube and shell heat exchangers destructively vibrate at certain velocities with which fluid flows over their outside surfaces. An acoustic generator is positioned to respond to the fluid vibrations generated by the tubes. The structure of the acoustic generator is sized and arranged to produce a modulated tone representative of tube vibration. A detector mounted external the heat exchanger shell responds to the tone to initiate corrective action which will reduce the tube vibration.

Abstract: Means and method are disclosed for determining the fluid content within a large pressure vessel (12), without penetrating the vessel. Strain gauges (60) are located at the weight-bearing support structures (24) of the vessel, whereby calibrated differences in the weight of the empty vessel and the vessel as measured during operation gives a direct indication of fluid content. Preferably, the strain gauges (60) are associated with strain amplification elements (80) or shims, which are interposed between mating surfaces of the vessel support (24) and its base (44). The stress amplification element (80) has a smaller area of contact with the support member (24) than with the base (44), providing an effective higher coefficient of elasticity than that of the support member. In one embodiment, the strain gauge (60) is connected to a bolt (64) which tightly secures the vessel support (24) to its base (44).

Abstract: A rail apparatus for running an automatic ultrasonic flaw detector around a reactor pressure vessel for detecting any flaw in the weld beads on the reactor pressure vessel. At least two circumferential rails and vertical rails are connected through turn tables at their crossing points to form a rail unit. The rail unit has vertical rails extending upward from the uppermost circumferential rail. The vertical rails are secured to a gamma shield, so that the rail unit as a whole is suspended from the gamma shield by means of these vertical rails. The rail unit is provided with swing prevention devices for preventing swing thereof. In installation of the rail apparatus, the rail unit is temporarily mounted on the reactor pressure vessel before mounting the latter in a containment vessel. Then, after accurately placing the reactor pressure vessel together with the rail unit in a right place within the containment vessel, the rail unit is supported suitably by the gamma shield.

Abstract: A self-contained modular nuclear reactor which can be prefabricated at a factory location, nuclear-certified at the factory, transported to a field location for final assembly and connection to a large-scale electric-power generating facility. The modular reactor includes a prefabricated nuclear heat supply module and a plurality of shell segments which can be assembled about the heat supply module and which provide a form for the pouring and curing of a cementatious biological shield about the heat supply module. The modular reactor includes passive shutdown heat removal systems sufficient to render the reactor safe in an emergency. A large-scale power plant arrangement is disclosed which incorporates a plurality of the modular reactors.

Abstract: A rail apparatus positioned around the surface of a reactor pressure vessel surfaces (1) for carrying equipment such as a flaw detector to permit the equipment to move along the surface of a reactor pressure vessel in the vertical and circumferential directions. The rail apparatus has a plurality of vertical rails (81) and at least one circumferential rail (82) connected to the vertical rails through turn tables (83) at respective points of intersection. Each vertical rail (81) is supported by a supporting member wall fixed at its one end to a gamma-ray shielding wall (2) surrounding the reactor pressure vessel. The vertical and circumferential rails as a unit are retained by the gamma-ray shielding member (84) through oscillation prevention means (86).

Abstract: Machine for monitoring internal equipment of a nuclear reactor, when stored outside the reactor vessel in the swimming pool of the reactor, comprising a first longitudinal carriage (16) movable on rails (14) on the bottom of the swimming pool, a transverse beam (46) movable on the first carriage (16), a second transverse carriage (61) bearing monitoring instruments (65) and movable on the beam (46), and an arrangement for moving the beam (46) in predetermined steps on the first carriage (16), and the second carriage (61) on the beam (46).

Abstract: Given are a method and apparatus for measuring nondestructively and non-invasively (i.e., using no internal probing) the burnup, reactivity, or fissile content of any material which emits neutrons and which has fissionable components. No external neutron-emitting interrogation source or fissile material is used and no scanning is required, although if a profile is desired scanning can be used. As in active assays, here both reactivity and content of fissionable material can be measured. The assay is accomplished by altering the return flux of neutrons into the fuel assembly. The return flux is altered by changing the reflecting material. The existing passive neutron emissions in the material being assayed are used as the source of interrogating neutrons.

Abstract: Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If the strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint.

Abstract: A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am.

Abstract: Apparatus for removing deposit from fuel elements of a nuclear reactor, by scraping the elements C by means of a remote-controlled scraping device entering the pool A of the reactor. The scraped sample is sucked, together with a certain amount of liquid, into an intermediate reservoir (20), the liquid removed with the sample is passed into a detachable filtration chamber (3), which retains the sample, the liquid being led into a buffer reservoir (21), and the filtration chamber (3) is introduced into a storage container (4), all the operations being carried out under the protection of shielded walls.

Abstract: A water level indicator for a pressure vessel is provided which employs a plurality of vertically spaced cantilevered mounted spring members. The free end of each spring member is connected to a depending float member. The float member has sufficient mass and density to depress its spring member when the float member is in a gaseous medium. The float member has sufficient buoyancy to elevate its spring member when the float member is in a water medium. One or more strain gauges is secured to each spring member to indicate the instantaneous position of the spring member and thereby provide an indication of the instantaneous water level.

Abstract: A nuclear steam generator including a vessel, means to introduce vaporizable fluid into the bottom portion of the vessel, an outlet near the top through which vapor is discharged, a horizontal tube sheet extending across the vessel, a plurality of U-shaped tubes, having each end secured to and extending through the tube sheet, means for introducing heating fluid to one end of each of the U-shaped tubes, means for removing heating fluid from the other end of each of the U-shaped tubes, tube support means positioned within the vessel for preventing tube vibration, the tube support means including horizontally positioned means closely surrounding, but slightly spaced from each tube, means through which access can be had to the vessel interior beneath the tube sheet when the steam generator is not in operation, and testing means for determining whether or not a tube is locked into a tube support means including a longitudinal member, with a first end located inside the tube to be tested, and a second end located

Abstract: Bolts of a liquid metal fast breeder reactor, each bolt provided with an internal chamber filled with a specific, unique radioactive tag gas. Detection of the tag gas is indicative of a crack in an identifiable bolt.

Abstract: An apparatus for remotely positioning sensing devices within a reactor. The apparatus comprises a pressure-tight housing located adjacent to the reactor wherein the housing has a port connected by means of a passageway to an inlet of the reactor. Valve means are provided within the passageway for isolating the housing from the reactor. A plurality of cylinders are rotatably mounted within the housing about an axis thereof such that each of the cylinders may be aligned with the port in the housing. Each of the cylinders have a piston and a ram. The ram includes a sensing device such as a thermocouple. Means are provided for moving the piston within the cylinder such that when a cylinder is registered with the housing's port, the sensing device may be positioned in the reactor.

Abstract: A system for sampling material from a remote location which selectively collects a predetermined quantity of material in a valve disposed inside a container. This measured amount of material is transferred to a movable receiving carrier while the receiving carrier is in a sealed relationship with the valve to prevent leakage of the sample material outside the receiving carrier. The movable receiving carrier is transported via an elongated tubular structure to a location remote from the container.

Abstract: Probe-holding apparatus for holding a probe for checking steam generator tubes particularly in a nuclear reactor installation. The apparatus comprises a telescopic arm supported via a ball and socket joint from a support mounted in or near an access aperture in a chamber at one end of the steam generator. A probe guide is carried by a carriage pivotally mounted at the other end of the telescopic arm. The carriage includes an endless belt having a series of spaced projections which engage into the ends of the tubes, the projections being spaced by a distance equal to the tube pitch or a multiple thereof. The belt is driven by a stepping motor in order to move the carriage and place the probe guide opposite different ones of the tubes.

Abstract: An apparatus for detecting or otherwise monitoring, e.g. counting, each passage of a succession of conductive objects (especially graphite-sheathed ball-shaped nuclear fuel or breeder elements for the core of a nuclear reactor) through a pipe or tube having a conductive wall juxtaposed with the tip of an electrode. A high-voltage direct-current (d-c) generator is connected across the wall and the electrode in series with a high-ohmic resistor. A voltage divider is also connected across this gap and has a low-voltage portion across which a voltage drop of the order of up to 100 volts is detected as the signal which is applied to the monitoring circuit. Thus the electronic detector and counter circuit are connected across the low-voltage resistor of the voltage divider and this resistor is connected to the side of the electrode-pipe system opposite that to which the high-ohmic resistor is connected.

Abstract: The baffle maintenance apparatus comprises a peening apparatus mounted on a movable carriage for reducing the width of a gap between two adjacent baffle plates in a nuclear reactor. The peening apparatus comprises an hydraulic hammer mounted on a pivotable plate that is capable of being positioned in proper relationship to the baffle plate for reducing the gap therebetween. The baffle maintenance apparatus may also comprise a gauging mechanism mounted on the carriage for determining the width of the gap between the baffle plates.

Abstract: A mounting for power conductor and instrument lead penetrations in nuclear reactors provides a device that is simple, compact and at the same time easily installed, repaired or replaced. The mounting comprises a cylinder like, stepped housing which is to pass through the pressure vessel wall of the nuclear reactor and house at least one and preferably a plurality of electrical conductors.

Abstract: An amphibious end plug gauging device and method for determining the axial location of nuclear fuel element end plugs within or without the upper tie plate of a fuel bundle. The gauging device is maneuvered toward an underwater fuel bundle in a reactor service pool and its tip inserted in a preselected tie plate shank hole to engage the end of a fuel element in the bundle. The position of the fuel element relative the upper surface of the tie plate is registered on a dial indicator.

Abstract: This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the dealy time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector.

Abstract: A control system for coupling redundant logic channels while maintaining electrical and physical isolation between channels. The separate redundant logic channels each include an electro-optic modulator, which are connected by fiber optic links, with system condition sensors generating signals which are applied to the electro-optic modulators to control the light transmission characteristics of the electro-optic medium in response to a predetermined system condition. A light source is connected by input fiber optics to the logic channels, and output fiber optics connect the logic channels to a light detector.

Abstract: An ultrasonic transducer apparatus and method for examining nuclear reactor jet pump beams for cracking. Examination is conducted in situ. An operator lowers a carriage portion of the apparatus into the reactor vessel with a pole. Ultrasonic signals are transmitted through suitable wiring from an external source to the transducer apparatus, which may employ a pitch-catch or pulse-echo mode of ultrasonic examination to test the beams. The carriage holds oppositely disposed pairs of ultrasonic transducers and positions them suitably near the beam to be examined in a proper orientation thereto. The mode of examination is selected by a switching mechanism. The apparatus includes a signal generator, receiver, and visual display.