Abstract: The invention concerns a position detector without mechanical contact. A permanent magnet attached to a mobile component co-operate with a sensor governed by the local action of the magnet. The sensor is constituted by an electromagnetic delay line having a saturable ferrite core onto which a winding propagating high frequency pulses is coiled. A reflectometer delivers the position indicator signal. The detector may be used to detect the position of a nuclear reactor control rod.

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: 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: The present invention relates to a device making it possible to detect variations in the height of the free level of a liquid in an enclosure.The device comprises a vertical short tube issuing into the upper part of the vessel and whose upper end issues into a first condensation pot, a long tube whose long end issues into the vessel at a level below that at the end of the short tube, the end of the long tube being provided with a hydraulic seal, while the upper end of the long tube issues into a second condensation pot, a differential manometer indicating any pressure differences between the two condensation pots.Application to the measurement of the level of the water contained in the vessel of a pressurized water nuclear reactor.

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: A fuel pin transfer tool has a latching device of the collet type attached to a first member movable vertically through a long work stroke enabling a fuel pin in an under water assembly to be engaged and withdrawn therefrom or placed therein and released. The latching device has a collet provided with a plurality of resilient fingers having cam portions normally spaced apart to receive the upper end of a fuel pin between them and a second member, movable vertically through a short stroke relative to the first member is provided with cam portions engageable with those of the fingers and is yieldably and resiliently held in a raised position in which its cam portions engage those of the fingers and force the fingers into their pin-gripping positions.

Abstract: A liquid level detector, or detector array, for insertion into the pressure vessel of a nuclear reactor, or into any liquid-containing vessel, for that matter. A coaxial cable supports the detector, which includes a sensing element having a positive temperature-coefficient of electrical resistivity. The detector is driven by a constant current source of electricity and the presence of liquid is determined by making a resistance measurement.

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 method of nondestructively detecting the presence of free liquid within a sealed enclosure containing solidified waste by measuring the levels of waste at two diametrically opposite locations while slowly tilting the enclosure toward one of said locations. When the measured level remains constant at the other location, the measured level at said one location is noted and any measured difference of levels indicates the presence of liquid on the surface of the solidified waste. The absence of liquid in the enclosure is verified when the measured levels at both locations are equal.

Abstract: Method and apparatus for monitoring the movement of the control rods of a nuclear reactor.The number of steps of movement in either direction of the rod from which the control rod is suspended is counted. According to the height of the step, an indication of the position of the suspension rod and of the control rod. The apparatus comprises devices for measuring the speed of movement of the control rod, for logging variations in speed higher than a given value, and for counting such variations according to their sign. The invention is particularly useful in pressurized water nuclear reactors.

Abstract: A fuel pin transfer tool has a tubular housing of sufficient length to enable it to be suspended in a body of water in which a fuel pin assembly is submerged with the upper end of the housing above the water and with the tool in vertical alignment with a selected fuel pin from which the lower end of the tool is closely spaced. The housing has a cylinder spaced from its ends in which there is a piston having a tubular piston rod extending slidably through both ends of the cylinder. An actuator, the control for which has first and second positions, is fixed on the upper end of the piston rod and has a member extending freely downwardly therethrough which on short downward travel, resulting when the control is in its second position, effects the pin-releasing position of a latching device secured to the lower end of the piston rod.

Abstract: This invention relates to a system for sensing the liquid level and coolant properties in a fluid containing vessel by insertion into the vessel of a plurality of matched spaced sensor probes comprising a first and second temperature sensor, the first sensor essentially measuring the media temperature, a heater, an arrangement for applying differential heat between the sensors, the heater and second sensors being thermally connected, all positioned to permit the sensing at the cooling quality of the media present, the heater located directly adjacent the second sensor and adjusted so that when the cooling quality of the normal fluid is sensed the heat generated by the heater and passing into the fluid is virtually completely dissipated and only a small finite differential temperature exists between the sensors and thereby indicates a sub-cooled liquid is present.

Abstract: A liquid level sensing apparatus (32) senses the level of liquid surrounding the apparatus. A plurality of axially spaced sensors (36) are enclosed in a separator tube (86). The separator tube (86) tends to collapse the level of a two-phase fluid within the separator tube into essentially a liquid phase and a gaseous phase where the collapsed level bears a relationship to the coolant inventory outside the separator tube. The level of the liquid phase is sensed by level sensing apparatus (32). The separator tube (86) contains inlet-outlet ports (90,92) near the top and bottom thereof to equalize the liquid level inside and outside the separator tube (86) when the level fluctuates or the water within the separator tube (86) flashes to steam. Each sensor (36) is comprised of a heater (64), a heated thermocouple junction (68) and an unheated thermocouple junction (66) within an elongated heat conductive housing (62).

Abstract: In a nucler reactor containment structure, a fuel transfer arrangement including a bridge, trolley, and grapple, each of the latter having a position sensor mounted thereon, and control means for determining the position of bridge, trolley, and grapple in response to signals from said position sensors, whereby fuel assemblies may be transferred between locations in the containment structure.

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: The faster acting junction of a difference junction thermocouple associated with a gamma sensor is located vertically above the other junction to ordinarily monitor local power generation in the fuel core of a nuclear reactor and in response to a reversal in the signal voltage polarity indicate a drop in level of coolant to thereby also function as a coolant monitor. An internal electrical heater enhances the coolant monitoring capability of the sensor, which may also be extended into the dome of the reactor vessel for monitoring conditions therein.

Abstract: An apparatus for detecting nominal phase conditions of coolant in a reactor vessel comprising one or more lengths of tubing each leading from a location being monitored to a closed outer end exterior of the vessel. Temperature is sensed at the open end of each length of tubing. Pressure within the tubing is also sensed. Both measurements are directed to an analyzer which compares the measured temperature to the known saturated temperature of the coolant at the measured pressure. In this manner, the nominal phase conditions of the coolant are constantly monitored.

Abstract: A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

Abstract: A water level gauge for a nuclear reactor includes heat conduction means for absorbing and conducting heat to the outside of the drywell, the heat being supplied to reference water head water inside a reference water head tube defining a reference water level inside a reference water level container, and to water inside a lower part pressure pipe connecting a nuclear reactor pressure vessel with a differential pressure detector. The water level gauge further includes heat exchanger for removing the heat. With this device, the water level of the nuclear reactor may be correctly detected even in the case of leakage accident of the coolant of the nuclear reactor.

Abstract: This system incorporates a uniquely coded longitudinally extending absolute distance standard or code chart. The code chart includes bars and segments of different light reflectivity characteristics which uniquely define every different distance increment along the chart. Preferably the chart is binary coded. A transducer optically detects the code at each different distance measurement increment and converts the code into electrical signals. The electrical signals operatively control a manipulator device which moves along a movement track parallel to the stationarily positioned chart. The transducer includes means for detecting the code at or near the center of each distance measurement increment, rather than at the edges of the distance measurement increments where the code on the chart undergoes transitions. A sample bar extending along the chart defines the distance measurement increments and the points at which the code at or near the center of the distance measurement increments is to be sampled.

Abstract: A length of metal sheathed metal oxide cable is perforated to permit liquid access to the insulation about a pair of conductors spaced close to one another. Changes in resistance across the conductors will be a function of liquid level, since the wetted insulation will have greater electrical conductivity than that of the dry insulation above the liquid elevation.

Abstract: A nuclear reactor control rod position system wherein the control rods are connected to drive rods movable within a group of detector coils. Bilateral switches are connected to the respective coils of the detectors and a control circuit sequentially provides an enabling control signal to the various groups of bilateral switches connected to the respective detectors so that the coils of each detector in turn are connected to a single signal processing circuit which may be located inside or outside the containment structure of the reactor.

Abstract: A monitor circuit for a control rod drive mechanism (CRDM) that is adapted to detect erroneous stator phase sequences and generate a visible and audible alarm. The monitor circuit has particular application for use in connection with the control element assembly of a nuclear reactor. The monitor circuit includes an interface circuit for each stator phase winding of the drive motor, a plurality of logic gates to detect the predetermined erroneous phase sequences, a plurality of flip-flops to maintain the fault condition signal, visible and audible alarm circuits, and circuitry for resetting the monitor circuit after a fault condition. The monitor circuit also includes a ratchet trip circuit for disengaging the clutch mechanism from the control rods when the rods are in motion in the time a fault condition is detected.

Abstract: A nuclear reactor of the type including a reactor vessel and a core assembly to be maintained in a fixed position within the vessel is disclosed herein along with a structural arrangement also located in the vessel, for supporting the core assembly in its fixed position. The structural arrangement disclosed utilizes a plurality of components including a grillage of I-beams interconnected to one another by welded joints so as to define a unitary structure capable of supporting the core assembly within the vessel. These components including the I-beams are also mechanically interlocked such that a total failure of the welded joints will result in a limited but readily detectable downward deflection of the unitary structure, thereby indicating such a failure while, at the same time, retaining sufficient structural integrity to maintain the core assembly in a supported position.

Abstract: A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

Abstract: A nuclear reactor system is described in which a position indicator is provided for detecting and indicating the position of a movable element inside a pressure vessel. The movable element may be a valve element or similar device which moves about an axis. Light from a light source is transmitted from a source outside the pressure vessel to a first region inside the pressure vessel in alignment with the axis of the movable element. The light is redirected by a reflector prism to a second region displaced radially from the first region. The reflector prism moves in response to movement of the movable element about its axis such that the second region moves arcuately with respect to the first region. Sensors are arrayed in an arc corresponding to the arc of movement of the second region and signals are transmitted from the sensors to the exterior of the reactor vessel to provide indication of the position of the movable element.

Abstract: In a system, such as a water-cooled nuclear reactor, water level sensors provide signals to a control channel for control of feedwater flow and hence the water level in the pressure vessel. Rapid and fine level control is provided by comparing fluid outflow and inflow. Means are provided to block automatically the flow comparison signal from the control circuit in response to a signal characteristic of a faulted flow comparison signal. At least one redundant control channel is provided and level control is transferred thereto automatically upon excusion of the water level beyond predetermined normal upper and lower limits or in response to a rapid change in the water level control signal to avoid unnecessary shutdown in the event of a faulted channel.