Abstract: A controller for producing a nuclear reactor shutdown system trip signal in response to at least one detector signal. The controller includes a signal conditioning module receiving the at least one detector signal and outputting a measured flux signal. A rate module generates a rate signal from the measured flux signal. A comparator circuit compares the rate signal to a trip setpoint and generates a first trip signal.

Abstract: A spring mounting device in a cladding tube for nuclear fuel, comprising a spring distributor, a spring loader in the cladding tube, the distributor supplying the springs to the loader, said loader comprising a longitudinal slide to receive the spring, a pusher in order to set the spring into place in the cladding tube and is able to move in the slide, means of displacement of said pusher, said device comprising mechanical means in order to associate the actuation of the means of displacement with that of the spring distributor.

Abstract: In one embodiment, the method includes determining the safety limit minimum critical power ratio using the operating limit minimum critical power ratio, a change-in-critical-power-ratio distribution bias and a change-in-critical-power-ratio distribution standard deviation.

Abstract: According to the method, a reactor core is shut down during the operation cycle. One or more fuel bundles of the reactor core are then moved to new positions within the reactor core to increase a total energy, output of the reactor core as compared to continuing operation of the reactor core without the shutting down and moving steps.

Abstract: With this method it is possible to evaluate opposing friction forces occurring after a certain operating time on a mobile object in a guide. The method consists of measuring the variations in the speed of movement of the mobile object, of calculating the theoretical speed of movement without friction, of combining these two results and thereby deducing outside opposing forces through calculation of the acceleration of the mobile object. Particular application to the evaluation of additional friction forces acting on a mobile control cluster assembly in the core of a pressurized water nuclear reactor.

Abstract: A fuel assembly for a pressurized water reactor having control rod guide thimbles (5) each having a dashpot (12) for protecting against flexural deformation which may impair insatiability of a control rod. The fuel assembly includes, a plurality of control rod guide thimbles (5) having bottom and top end portions fixedly secured to a lower nozzle (2) and an upper nozzle (4), respectively, disposed in opposition to each other. The dashpot (12) of each control rod guide thimble (5) includes a small diameter section (13b) having an outer diameter smaller than that of the control rod guide thimble (5) formed at an upper portion of the dashpot (12), and a large diameter section (13a) having an outer diameter substantially equal to that of the control rod guide thimble (5) formed at a lower portion of the dashpot (12). With the length of the control rod guide thimble (5) represented by L, the effective length (S) of the small diameter section (13b) is selected to lie within a range of from 0.03 L to 0.1 L.

Abstract: In order to provide a boiling water type nuclear reactor use control rod with a guide use roller which improves corrosive environment at a clearance in the guide use roller and suppresses a generation of stress corrosion cracking while maintaining sliding function of the control rod by the guide use roller, at least one of a handle 5, a lower portion supporting plate 2a and a dropping speed limiter 2 is provided with the guide use roller, and a space which causes water flow in a clearance between a pin 9 and a pin hole 12 in the guide use roller is provided adjacent the clearance.

Abstract: A method of controlling a nuclear power plant includes determining a first maximum linear heat generation rate and a first minimum critical power ratio by a core monitoring system, and determining a second maximum linear heat generation rate and a second minimum critical power ratio by utilizing the first maximum linear heat generation rate and the first minimum critical power ratio determined by the core monitoring system and plant data. The second maximum linear heat generation rate and the second minimum critical power ratio are compared with predetermined thermal limit values and a control signal outputted from an automatic power regulator system to a re-circulation flow control system and to a control rod control system is held when at least one of the second maximum linear heat generation rate and the second minimum critical power ratio exceeds the predetermined thermal limit values.

Abstract: A control system, and a control method, of a nuclear power plant capable of easily executing automatic output regulation by an automatic power regulator system even when a lower-accuracy maximum linear heat generation rate and a lower-accuracy minimum critical power ratio are determined in a short cycle by using brief calculation. While a core monitoring system does not calculate a maximum linear heat generation rate and a minimum critical power ratio, they are determined by utilizing the maximum linear heat generation rate and the minimum critical power ratio calculated by the core monitoring system as well as plant data, and the lower-accuracy maximum linear heat generation rate and the lower-accuracy minimum critical power ratio so determined are compared with predetermined thermal limit values.

Abstract: A fuel assembly for a pressurized water reactor having control rod guide thimbles (5) each having a dashpot (12) for protecting against flexural deformation which may impair insertability of a control rod The fuel assembly includes, a plurality of control rod guide thimbles (5) having bottom and top end portions fixedly secured to a lower nozzle (2) and an upper nozzle (4), respectively, disposed in opposition to each other. The dashpot (12) of each control rod guide thimble (5) includes a small diameter section (13b) having an outer diameter smaller than that of the control rod guide thimble (5) formed at an upper portion of the dashpot (12), and a large diameter section (13a) having an outer diameter substantially equal to that of the control rod guide thimble (5) formed at a lower portion of the dashpot (12). With the length of the control rod guide thimble (5) represented by L, the effective length (S) of the small diameter section (13b) is selected to lie within a range of from 0.03 L to 0.1 L.

Abstract: A transiently unstable state of a boiling-water reactor is damped by measuring the oscillating neutron flux and, after a first threshold value is exceeded over a plurality of oscillation periods, the rate of increase of the oscillation is determined. Depending on the slope of the increase, and after a further threshold value (in particular one dependent on the rate of increase) for the oscillating flux is exceeded, a selection is made which one of various stabilization strategies for damping the oscillation should be triggered before a SCRAM becomes necessary. A hierarchy of stabilization strategies is available: blocking a power increase, for instance, in the control system, controlled slow reduction of the power, or rapid power reduction by a "partial SCRAM". The monitoring of the unstable state is effected with a system of sensors strategically distributed throughout the core and which redundantly measure the flux in one region of the core.

Abstract: A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation).

Abstract: A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors.

Abstract: In a drive device of control rod drive mechanisms, a large number of control rod drive mechanisms are divided into a plurality of groups, and there are provided, for each group, a control rod changeover device, and an inverter power source and inverter controller constituting the drive power source. The drive device comprises a control unit that receives control rod position signals from each of the control rod drive mechanisms and that outputs control signals to the control rod changeover device and inverter controller, and a man-machine device constituting an interface with the operator, that outputs control rod drive information to this control unit.

Abstract: A control rod drive having a magnetic coupling device which eliminates the need for packing-type seal assemblies for preventing leakage from the reactor pressure vessel adjacent the spindle. The magnetic coupling device includes a driving rotor and a driven rotor separated by a pressure barrier which forms part of the reactor pressure boundary. The driven rotor and the driving rotor are coupled magnetically so that the driven rotor rotates in response to rotation of said driving rotor, whereby the control rod displaces in response to operation of the drive motor. The driven and driving rotors are each constructed by stacking a plurality of circular magnetic rings concentric with a centerline axis. Each circular ring has multiple poles circumferentially distributed thereon. An anti-rotation device is incorporated to prevent free rotation of the spindle when the drive motor is removed during maintenance operations. The anti-rotation device includes a push rod actuator which penetrates the pressure barrier.

Abstract: A safety system grade dropped rod detection system for a pressurized water reactor (PWR) utilizes core exit thermocouples arranged in multiple trains and hot and cold leg RTDs to generate a safety system grade rod stop signal. The system generates from the temperature signals a relative power deviation (RD) and a curvature index (CI), which is the spatial second derivative of RD for each fuel assembly. The CI signatures not only provide rapid, reliable detection of dropped control rods, but also clearly identify failed and failing thermocouples.

Abstract: A temperature regulation system of a pressurized water nuclear reactor is inhibited when a combination of the nuclear and the axial power difference exceeds a predetermined inhibit threshold.

Abstract: A nuclear reactor includes a pressure vessel enclosing a plenum, a core in the pressure vessel, a pressure-tight encapsulation in the pressure vessel communicating with the plenum, and control members passing through the encapsulation and having first and second control member segments. The second control member segment has an absorber part to be at least partly inserted into and retracted out of the core for controlling reactivity. The first control member segment has an outer drive part. A control drive is associated with the outer drive part for adjusting the control member along the longitudinal direction thereof. An antieject device for restraining ejection of the control member out of the pressure vessel due to a leak at the encapsulation includes a rod coupling having a disengageable coupling catch interconnecting the first and second control member segments and remaining in engagement during normal control motion of the control member and means for unlocking the coupling catch.

Abstract: A brake assembly is disclosed for selectively preventing rotation of a shaft, such as a shaft used in a control rod drive for a nuclear reactor. The brake assembly includes a stationary base, a rotor disc fixedly connected to the shaft for rotation therewith, and a brake member disposed adjacent to the rotor disc. The rotor disc includes at least one rotor tooth and the brake member includes at least one braking tooth. The member is selectively positionable in a deployed position for allowing the braking tooth to contact the rotor tooth for preventing rotation of the shaft in a first direction while permitting roation in the opposite direction, and in a retracted position for allowing the rotor disc and shaft to rotate in either direction without restraint from the brake member.

Abstract: A radial brake assembly is disclosed for selectively preventing rotation of a shaft, such as a shaft used in a control rod drive for a nuclear reactor. The brake assembly includes a stationary housing, a rotor disc fixedly connected to the shaft for rotation therewith, and a brake member disposed adjacent to the perimeter of the rotor disc. The rotor disc includes at least one rotor tooth and the brake member includes at least one braking tooth. The brake member is selectively positioned in a deployed position for allowing the braking tooth to contact the rotor tooth for preventing rotation of the shaft in a first direction, and in a retracted position for allowing the rotor disc and shaft to rotate without restraint from the brake member.

Abstract: A drive for moving a control rod in a pressure vessel containing a pressurized fluid includes a spindle disposed solely within the pressure vessel, which spindle is effective for moving the control rod upon rotation thereof. A driven rotor is fixedly joined to the spindle within the pressure vessel for rotating the spindle, and the driven rotor is selectively rotated in first and second opposite directions without the need for packing-type seal assemblies for preventing leakage of the pressurized fluid from the pressure vessel adjacent to the spindle.

Abstract: A latch assembly is disclosed for selectively preventing rotation of a shaft, such as a shaft used in a control rod drive for a nuclear reactor. The latch assembly includes a stationary housing for receiving the shaft, and a gear fixedly joined to the shaft. A latch arm is pivotally joined to the housing and includes at least one latch tooth facing gear teeth extending from the gear. The latch arm is selectively positionable in an engaged position to abut the latch tooth against a first one of the gear teeth for preventing rotation of the shaft in a first direction, and in a disengaged position to space the latch tooth away from the gear teeth for allowing the shaft to rotate without obstruction between the gear teeth and the latch tooth.

Abstract: A lock assembly is disclosed for selectively preventing rotation of a shaft, such as a shaft used in a control rod drive for a nuclear reactor. The lock assembly includes a stationary housing for receiving the shaft, and a gear fixedly joined to the shaft. The gear includes a plurality of circumferentially spaced gear teeth. An elongate key is slidably joined to the housing and has at least one locking tooth facing the gear teeth. The key is selectively positionable in an engaged position wherein the locking tooth and gear teeth prevent rotation of the shaft in a first direction, and in a disengaged position for allowing the shaft to rotate without obstruction between the gear teeth and the locking tooth.

Abstract: A rotary lock is disclosed for preventing rotation of a shaft, such as a shaft used in a control rod drive for a nuclear reactor. The lock includes a stationary housing for receiving the shaft, and an arm extending outwardly from the shaft. At least one pin is joined to the housing for circumferential restraint and is positionable in a withdrawn position away from the arm for allowing the arm to rotate with the shaft without obstruction from the pin. The pin is also positionable in a deployed position contacting the arm for preventing rotation of the arm and shaft. Means are provided for selectivity positioning the pin in the withdrawn and deployed positions.

Abstract: A nuclear fuel assembly has mechanical elements which constrain the flux thimble tube within the instrumentation tube to maintain physical contact with the interior surface of the instrumentation tube at a plurality of points staggered on a single-diametral plane and spaced substantially throughout the length of the instrumentation tube. The staggered mechanical elements induce a controlled elastic sinuous deflection of the inner thimble tube. The mechanical elements can take several different forms, for example, dimples or cantilevered spring fingers formed in the wall of the instrumentation tube and projecting radially inwardly therefrom.

Abstract: A process for automatic regulation of the soluble boron content of the cooling water of a pressurized water nuclear reactor in which operating regions (2, 3) of the means of boration and of the means of dilution respectively are determined a priori, corresponding to pairs of values of two control parameters relating to the position of the reactor regulating group in the core and to the deviation of the axial power imbalance relative to the reference axial imbalance. The momentary value of the control parameters is determined continuously during the operation of the reactor, and the stopping or the triggering of the means for boration or for dilution is commanded when the operating point crosses a boundary of an operating region (2, 3).The invention applies in particular to a nuclear reactor controlled in G mode.

Abstract: An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Abstract: A protection system such for a nuclear power plant shuts down the reactor in response to the detection of a negative flux rate characteristic of a dropped rod, but inhibits the trip if the specified negative flux rate is preceded within a selected time interval by a positive flux rate characteristic of a net load rejection when the moderation temperature coefficient is highly negative.

Abstract: The setpoints for control systems in a pressurized water reactor are adjusted by an amount corresponding to the expected change in the controlled process variable resulting from rapid fluctuations in load to reduce the duty time of the control system components such as the rod drive mechanisms in the rod control system and the spray and heater units in the pressurizer pressure control system. In addition, the deadband in the response of the rod control system is continuously varied as a function of the variance of the magnitude of the rapid fluctuations in load to further reduce wear while maintaining good response during load following.

Abstract: In a pressurized water reactor in which a dropped control rod is detected by a negative rate of change of neutron flux which exceeds a preset negative limit, the reactor is only shutdown if the axial offset exceeds a preselected limit, so that it is possible to maintain power, and in many cases full power, despite the presence of a dropped rod. Where part length control rods are used during power changes, separate limiting values on axial offset are used when these rods are fully inserted and when they are fully retracted.

Abstract: A velocity limiting device providing greater resistance to motion through a fluid in one direction than in the opposite direction. The device includes a toroidally shaped control member having a smooth generally conical surface on one side for low resistance movement through the fluid in the one direction. On its opposite side the control member is formed with at least one groove providing a concave surface spaced from which is an arrangement of vanes forming a jet for directing a stream of fluid into the groove upon movement in the opposite direction whereby the resultant increase in fluid turbulence increases resistance to rapid movement in this opposite direction.