Abstract: A method for configuring telemetry transmission includes receiving a transmitted telemetry signal at the earth's surface. The received signal is selected and processed to compute a measured power spectral density (PSD). A theoretical PSD is computed of the transmitted signal at the known carrier frequency and processed in combination with the measured PSD to compute an attenuation factor of the transmitted signal at a plurality of frequencies.

Abstract: To improve the accuracy of estimating when a fracture occurred in the metal material. In the estimation apparatus, the fracture temperature estimation unit performs estimate calculation of the fracture temperature of a fractured metal material from the hydrogen embrittlement rate of the fractured metal material, the fracture occurrence time narrowing unit performs calculation by using the temperature information of the installation environment in which the fractured metal material is installed and narrow down the period during which the fracture temperature is obtained as the fracture occurrence period of the metal material, and the fracture time estimation unit performs estimate calculation of the timing of when stress was applied to the fractured metal material during the fracture occurrence period as the timing of when a fracture occurred in the metal material.

Abstract: A machine-learning tool learns from sensors' data of a nuclear reactor at steady state and maps them to controls of the nuclear reactor. The tool learns all given ranges of normal operation and responses for corrective measures. The tool may train another learning tool (or the same tool) that forecasts the behavior of the reactor based on real-time changes (e.g., every 10 seconds). The tool implements an optimization technique for differing half-life materials to be placed in the reactor. The tool maximizes isotope production based on optimal controls of the reactor.

Abstract: A method for protecting a nuclear reactor includes reconstructing a maximum linear power density released among the fuel rods of the nuclear fuel assemblies of the core; calculating the thermomechanical state and the burnup fraction of the rods; calculating a mechanical stress or deformation energy density in the cladding of one of the rods by using the said reconstructed maximum linear power density, the calculated thermomechanical states and the calculated burnup fractions, by means of a meta-model of a thermomechanical code; comparing the calculated mechanical stress or the calculated deformation energy density with a respective threshold; and stopping the nuclear reactor if the calculated mechanical stress or the calculated deformation energy density exceeds the respective threshold.

Abstract: The invention relates to methods for protecting a nuclear reactor core, such as a boiling water reactor core, from fuel and cladding damage due to thermal hydraulic instability in extended operating power flow conditions and, in particular, when an extended power uprate is implemented. The methods employ existing licensed stability methodologies and incorporated minor changes, e.g., to the Average Power Range Monitor (APRM)-based trip system to preclude operation inside the stability vulnerable region of the power/flow map. The APRM-based trip system is modified to set down the APRM flow-biased scram line when core flow is less than a predetermined core flow to prevent the core from entering an unstable region of operation.

Abstract: A system configured to monitor the structural health of reactor vessel internals of a nuclear reactor is disclosed herein. The system includes a memory configured to store historical information associated with past performance of the nuclear reactor, and an anomaly detection subsystem including a control circuit configured to receive a signal from a sensor. The anomaly detection subsystem is configured to determine, via the control circuit, a characteristic of a vibrational response of the reactor vessel internals based, at least in part, on the signal; access, via the control circuit, the historical information stored in the memory; compare, via the control circuit, the determined characteristic to the historical information stored in the memory; and determine, via the control circuit, a condition of the reactor vessel internals based, at least in part, on the comparison of the determined characteristic and the historical information.

Abstract: A method for diagnosing faults includes receiving a system description of a thermal hydraulic system, the system description indicating a plurality of components and sensors. The method also includes constructing, based on physical conservations laws and using the system description, a plurality of physics-based models for the plurality of components, each of the plurality of physics-based models including unknown parameters. The method further includes receiving historical measurements and calibrating the physics-based models by calculating the unknown parameters of each of the physics-based models using the historical measurements to produce calibrated models. The method also includes receiving sensor measurements of the sensors, and calculating residuals corresponding to differences between measurements predicted by the plurality of calibrated models and the sensor measurements.

Abstract: A system including a sensor interface for determining a substitute frequency value via a sensor interface is provided. The system can include a first circuit receiving a frequency signal. The system can also include a sensor interface coupled to the first circuit and configured to determine a substitute frequency value based on the frequency signal. The system can also include a second circuit providing the substitute frequency value output from the sensor interface. The second circuit can provide the substitute frequency value in place of an analog input value by mimicking the behavior of an analog-to-digital converter. An apparatus including the sensor interface and methods of determining the substitute frequency value using a sensor interface are also provided.

Abstract: An abnormality degree calculation system includes: a feature amount vector extraction unit configured to generate and output a feature amount vector from an input signal originating from vibration of a target device; an encoding unit configured to receive as an input a set composed of the feature amount vector and a device type vector representing a type of the target device and output an encoding vector; a decoding unit configured receive as an input the encoding vector and the device type vector and output a decoding vector; a learning unit configured to learn parameters of the neural networks of the encoding unit and the decoding unit; and an abnormality degree calculation unit configured to calculate a degree of abnormality defined as a function of the feature amount vector from the feature amount vector extraction unit, the encoding vector from the encoding unit, and the decoding vector from the decoding unit.

Abstract: A detection system is provided. The detection system includes a sensing device and a controller. The sensing device is within an elevator and in communication with the controller. The sensing device continuously monitors sound waves within the elevator and communicates the sound waves as electrical signals to the controller. The controller operates the elevator and analyses the sound waves received as the electrical signals from the sensing device to detect an emergency condition within the elevator.

Abstract: A method for determining at least one threshold value of at least one operating parameter of a nuclear reactor is implemented by an electronic determination system and includes the steps of determining a first threshold value of a respective operating parameter for an operation of the reactor at a first power; and determining a second threshold value of said parameter for an operation of the reactor at a second power. The operation at the lower power of the first and second powers is an operation continued for a duration of at least 8 hours over a 24-hour sliding window. The method also includes determining a third threshold value of said parameter for an operation of the reactor at a third power between the first power and the second power.

Abstract: Embodiments are directed to providing a user interface (UI) that streamlines and simplifies the process of monitoring critical power-generation module (PGM) parameters after a PGM assembly is shutdown. The UI displays, in real-time, indicators corresponding to one or more post-shutdown PGM parameters. The UI provides indications of whether the post-shutdown PGM parameters meet post-shutdown criteria of the PGM assembly. When a post-shutdown PGM parameter does not meet the post-shutdown criteria, a user alert is provided to the user. A protocol may additionally be provided to the user. In some embodiments, the protocol may enable the user to return the PGM assembly to a condition that satisfies the post-shutdown criteria. The protocol may be a safety protocol and/or an asset protection protocol.

Abstract: Provided is a statistical overpower penalty calculation system for a generic thermal margin analysis model, the system including: a random number generating unit generating a plurality of random numbers; an power distribution generating unit generating power information of an axial direction and a radial direction for a core burnup; an operating condition generating unit extracting an arbitrary value for a plurality of operating conditions from the random number generated above; a POL calculating unit calculating a POL of a reload core thermal margin analysis model and a POL of a generic thermal margin analysis model and calculating a plurality of the overpower penalties through the POLs; and a statistics processing unit calculating tolerance limit values according to the core burnup by statistically analyzing a distribution formed of the plurality of the overpower penalties and selecting a smallest tolerance limit value as a representative value of the overpower penalties.

Abstract: A method, system, and apparatus for the thermal storage of nuclear reactor generated energy including diverting a selected portion of energy from a portion of a nuclear reactor system to an auxiliary thermal reservoir and, responsive to a shutdown event, supplying a portion of the diverted selected portion of energy to an energy conversion system of the nuclear reactor system.

Abstract: Illustrative embodiments provide for the operation and simulation of the operation of fission reactors, including the movement of materials within reactors. Illustrative embodiments and aspects include, without limitation, nuclear fission reactors and reactor modules, including modular nuclear fission reactors and reactor modules, nuclear fission deflagration wave reactors and reactor modules, modular nuclear fission deflagration wave reactors and modules, methods of operating nuclear reactors and modules including the aforementioned, methods of simulating operating nuclear reactors and modules including the aforementioned, and the like.

Abstract: A method of determining a core design parameter of a nuclear reactor, includes: calibrating an isolated voltage output from a NIS cabinet associated with the nuclear reactor using a calibrated signal source as an input to the NIS cabinet; recording values of the calibrated signal source used in the calibrating and corresponding values of the output voltage from the calibrating in an as-left cabinet calibration data table; using a computing device connected to the isolated voltage output from the NIS cabinet, converting the voltage output signal to a converted detector signal using at least some of the values in the as-left cabinet calibration data table in an improved signal conversion equation; and using the computing device, employing the converted detector signal to determine the core design parameter.

Abstract: An electronic device (18) is for managing the display of data to control a nuclear power plant. The data comes from a plurality of electronic control units (16A, 16B, 16C). Each control unit is configured to perform at least one action from among acquiring a value measured by a sensor (12A, 12B, 12C) and controlling an actuator (14A, 14B, 14C), the control units, sensor(s) and/or actuator(s) being according to several different nuclear safety classes. This electronic device (18) is able to be connected to the control units, and includes a set (25) of electronic module(s) (26A, 26B, 26C) for creating overlay(s) (28A, 28B, 28C). Each overlay contains information associated with one or several control units and according to a respective safety class; and a module (30) for generating a page (32) to be displayed, by superposition of several separate overlays.

Abstract: A correlation tolerance limit setting system using repetitive cross-validation includes: a variable extraction unit randomly classifying data of an initial DB set into training set data and validation set data at a specific rate and then extracting variables for determining a DNBR limit by optimizing coefficients of a selected correlation; a normality test unit testing normality for a variable extraction result; a DNBR limit unit determining whether data sets have a same population or not depending on normality result and determining DNBR limit from a distribution of 95/95 DNBR; and a controller.

Abstract: Method and user interface for controlling an apparatus are provided. In accordance with one embodiment, the method comprises: displaying a dynamic bar on a display of a mobile electronic device, the dynamic bar having an appearance that updates dynamically to reflect a change to information managed by an executable running on the mobile electronic device, wherein the dynamic bar includes a label portion and an icon menu comprising a plurality of icons, each icon in the plurality of icons being associated with a respective application or function; and displaying a dialog for configuring options associated with a particular application or function, the dialog displayed in response to selection of a respective icon of the plurality of icons in the icon menu associated with the particular application or function.

Abstract: A simulation method run on a computer simulating the characteristics of a real controlled device including a heating apparatus that changes a heating value in accordance with a first manipulated value, includes creating a controlled-device model representing the real controlled device where a first manipulated value is an input and a process value for the real controlled device is an output, acquiring a first time-related characteristic as input to the controlled-device model, and calculating a second time-related characteristic from the output from the controlled device model with respect to the input of the first time-related characteristic. The controlled-device model includes a heating component corresponding to the heating apparatus for increasing the process value in accordance with the size of a first manipulated value.

Abstract: Control rod drives include all-digital monitoring, powering, and controlling systems for operating the drives. Each controlling system includes distinct microprocessor-driven channels that independently monitor and handle control rod drive position information reported from multiple position sensors per drive. Controlling systems function as rod control and information systems with top-level hardware interfaced with nuclear plant operators other plant systems. The top-level hardware can receive operator instructions and report control rod position, as well as report errors detected using redundant data from the multiple sensors. Positional data received from each drive is multiplexed across plural, redundant channels to allow verification of the system using independent position data as well as operation of the system should a single channel or detector fail.

Abstract: A method for controlling a pressurized water nuclear reactor is provided, including core producing thermal power, sensors for acquiring the mean temperature of the primary coolant and for calculating the thermal power, actuators for controlling the axial distribution of power, the control method including: a first control phase for controlling the reactor during normal operation by controlling the mean temperature of the primary coolant so as to make it correspond to a reference temperature profile (Pref) dependent on the thermal power of the reactor; and a second control phase, referred to as stretchout, that occurs after normal operation of the reactor in order to control the reactor in stretchout by controlling the axial distribution of power, the mean temperature varying freely in a temperature range delimited by an upper limit and a lower limit.

Abstract: Simulation of a complex system involves representing the complex system in a plurality of simulation building blocks, each simulation building block is associated with a subsystem of the complex system. Each simulation building block includes at least one client-side simulation component and a server-side simulation component. During the simulation a user interaction is detected by a client-side simulation component, the complex system is simulated in server-side simulation components on the basis of the user interaction. Simulation data for the subsystem is generated by the server-side simulation components, and the subsystem is represented by the corresponding client-side components based on the simulation data.

Abstract: A condensate drain apparatus including a valve for controlling discharge of condensate from the condensate drain apparatus; a collection chamber configured to receive a multiphase fluid flow upstream of the valve including gaseous and condensate phases, the collection chamber defining a collection volume for collecting the condensate to be discharged through the valve; sensor equipment for monitoring a parameter relating to a thermodynamic property of the fluid upstream of the valve; a controller configured to monitor the condensate collected in the collection chamber based on the monitored parameter, and to control the opening and closing of the valve to regulate the condensate collected in the collection chamber upstream of the valve; wherein the controller is further configured to determine the quantity of condensate discharged from the collection volume using a flow rate calculation for a choked flow of vaporizing liquid.

Abstract: A method, system, and apparatus for the thermal storage of nuclear reactor generated energy including diverting a selected portion of energy from a portion of a nuclear reactor system to an auxiliary thermal reservoir and, responsive to a shutdown event, supplying a portion of the diverted selected portion of energy to an energy conversion system of the nuclear reactor system.

Abstract: A monitoring control system includes: a safety control device that includes a state monitoring unit that monitors a state of a plant after emergency shut-down of the plant, and a work progress monitoring unit that monitors a work progress after the emergency shut-down of the plant; and a display device that displays, in a case where the emergency shut-down of the plant is performed by the safety control device, an emergency shut-down screen including a monitoring result of the state monitoring unit and a monitoring result of the work progress monitoring unit.

Abstract: A simulation system in particular for a control system which controls a process running in a technical system is provided. The control system has at least one first process environment embodied as a container and which is also designed to simulate the automatic process to be run in the system and includes corresponding interfaces to the guidance system. The simulation system includes, in addition to the first process environment, a second process environment embodied as a container for simulating the hardware of the periphery of the guidance system and a third process environment embodied as a container for simulating the process to be run in the technical system. In another embodiment, all process environments can also be combined to form one process environment. In both variations, the interfaces of the first process environment and the interfaces of the second process environment are practically identical to the interfaces of the third process environment.

Abstract: This invention relates to a control method for a pressurized water nuclear reactor, which comprises a core generating thermal power and means of acquiring magnitudes representative of core operating conditions. The method comprises a step to regulate the temperature of the primary coolant, if the temperature of the primary coolant for a given thermal power is outside a predefined set temperature interval (?TREF) depending on the reactor power. The set temperature interval (?TREF) is characterized by variable amplitude (?T) on a thermal power range between N % and 100% nominal power, where N is between 0 and 100 and comprises a zero amplitude at 100% nominal power, a zero amplitude at N % nominal power.

Abstract: The invention relates to a method for producing a radioisotope, which method comprises irradiating a volume of radioisotope-precursor fluid contained in a sealed cell of a target using a beam of particles of a given current, which beam is produced by a particle accelerator. The target is cooled and the internal pressure in the sealed cell is measured. During the irradiation, the internal pressure (P) in the sealed cell is allowed to vary freely. The irradiation is interrupted or its intensity is reduced when the internal pressure (P) in the sealed cell departs from a first tolerated range defined depending on various parameters that influence the variation in the internal pressure in the sealed cell during the irradiation. These parameters for example comprise, for a given target, particle beam and radioisotope-precursor fluid: the degree of filling of the hermetic cell, the cooling power used to cool the given target, and the beam current (I).

Abstract: The present disclosure describes system and methods for inferring energy usage at multiple levels of granularity. One embodiment describes an industrial automation system including a first industrial automation component, a first sensor coupled to the first industrial automation component, in which the first sensor measures a first amount of power supplied to the first industrial automation component, a second industrial automation component that couples to the first industrial automation component, and an industrial control system that infers energy usage by the first industrial automation component and the second industrial automation component based at least in part on the first amount of power supplied to the first industrial automation component.

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 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 nuclear power control system includes: a safety protection apparatus which outputs a first safety mode operating signal while outputting a first unsafety mode operating signal; a CCF apparatus that outputs a second safety mode operating signal; and a signal input/output circuit that is connected to the safety protection apparatus and the CCF apparatus. The signal input/output circuit includes an OR circuit that outputs a third safety mode operating signal based on the presence or absence of an input of the first safety mode operating signal and the second safety mode operating signal; a NOT circuit that is connected to the output side of the OR circuit; and an AND circuit that outputs a third unsafety mode operating signal, and the presence and absence of an input of a first unsafety mode operating signal.

Abstract: The ex-core nuclear instrumentation system includes a key hole, and it is configured such that in the state where a key lock switch is inserted into the key hole, writing to a electrically rewritable nonvolatile memory is made valid by a general-purpose logic, and when writing to a electrically rewritable nonvolatile memory is in valid and the operation panel and the FPGA perform a serial communication and in a case where the data order of the serial communication is the predetermined data order, the CPU controls the electrically rewritable nonvolatile memory, and rewriting data, which is outputted from the operation panel, to the electrically rewritable nonvolatile memory is transmitted from the FPGA to the electrically rewritable nonvolatile memory so as to rewrite the data in the electrically rewritable nonvolatile memory.

Abstract: A control system for monitoring a pressure vessel includes a pressure sensor coupled to a first pressure vessel component. The system also includes a level sensor coupled to a second pressure vessel component. The system additionally includes at least one computing device including at least one input channel configured to receive data from the pressure sensor and the level sensor and a processor coupled to the at least one input channel. The processor is programmed to populate a level and pressure dynamics model associated with the pressure vessel with data received from the pressure sensor and the level sensor, the processor further programmed to be capable of controlling operation of the pressure vessel based on the level and pressure dynamics model.

Abstract: This document's object is to provide an axial power distribution control method in which only the control of an axial power distribution in a nuclear reactor with a simple operation with a clear operational target keeps the control of a xenon oscillation, thereby suppressing the xenon oscillation to an extremely small magnitude in advance at the same time.

Abstract: An axial power distribution control device includes an axial offset calculation unit 52, a parameter calculation unit 53, and an axial offset determining unit 55. The axial offset determining unit 55 predicts whether a core axial offset of the power distribution is increased or decreased after a current time, based on a major axis of an ellipse drawn by the xenon parameter and the iodine parameter calculated by the parameter calculation unit 53 and the xenon parameter and the iodine parameter at the current time. This makes it possible to predict a change of the axial offset of the power distribution of a reactor for suppressing a xenon oscillation in the reactor.

Abstract: A central control device of a nuclear power plant monitors and controls the nuclear power plant. The central control device includes monitoring operation devices (for example, an alarm VDU, a common system VDU, and a safety system VDU of an operation console) each having both a function as a display unit that displays information on the nuclear power plant (for example, operation information, safety information, and the like) and a function as an operation unit that operates the nuclear power plant, and a control unit that controls these monitoring operation devices. The control unit selects an operation procedure document corresponding to an input operation to a monitoring operation device and displays the operation procedure document on the monitoring operation device.

Abstract: To predict xenon oscillation at the present time and later. For this purpose, as an axial offset of a power distribution of a reactor is represented by AOp, an axial offset of a power distribution based on xenon distribution is represented by AOx, and an axial offset of a power distribution based on an iodine distribution is represented by AOi, a parameter DAOpx(=AOp?AOx) and a parameter DAOix(=AOi?AOx) are described by a relational expression of a trigonometric function and an exponential function using an angular frequency of xenon oscillation. Next, phases with respect to initial values of the parameters DAOpx and DAOix are obtained (Step S101). The parameter DAOpx and the parameter DAOix expressed by the obtained phase and a coefficient of the relational expression obtained from this phase are plotted on an X coordinate and a Y coordinate, respectively.

Abstract: A computer implemented method for modelizing a nuclear reactor core, including the steps of: partitioning the core in cubes to constitute nodes of a grid for computer implemented calculation, calculating neutron flux by using an iterative solving procedure of at least one eigensystem, the components of an iterant of the eigensystem corresponding either to a neutron flux, to a neutron outcurrent or to a neutron incurrent, for a respective cube to be calculated. A control parameter is varied to impact a neutron eigenvalue ? through a perturbed interface current equation and drive the neutron eigenvalue ? towards 1.

Abstract: A control strategy for a pressurized water nuclear reactor that employs separate, independent control rod banks for respectively controlling Tavg and axial offset within corresponding deadbands. The strategy does not permit the control banks controlling reactor core power and the control banks controlling axial offset to move together, but normally gives preference to the control banks controlling the Tavg except when a demand signal is received simultaneously by both independent control rod banks to move in a same direction, in which case, the control bank compensating for the axial offset is given preference.

Abstract: The invention relates to a method for regulation of operational parameters of the core of a pressurised water nuclear reactor comprising: a step of acquisition of values (FH, FB, TBC, TBF, Q) which are representative of the conditions of operating of the core of the reactor; a step of evaluation of the actual values (Tmoye, AOe, {circumflex over (P)} maxe) of the operational parameters at least according to the values acquired (FH, FB, TBC, TBF, Q); a step of selection of a control law for the concentration of a neutron-absorbent component ([B]) and for the positions of insertion (Z1 to Z5) of the groups (P1 to P5) of rods selected from at least the first and second control laws which are different from one another; and a step of regulation of the operational parameters by means of the control law selected, according to set points (Tmoyc, AOc, {circumflex over (P)} maxc) relative to the said parameters and to the actual values (Tmoye, AOe, {circumflex over (P)} maxe) evaluated.

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 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: The invention is principally directed to a reduced order model, XEDOR, facilitating the prediction of and the diagnostics of pellet-clad interaction stress-corrosion-cracking failure of nuclear fuel rods. The invention more particularly relates to assessment of susceptibility to PCI failure for guidance in the design of fuel loading in nuclear reactors. The invention additionally relates to the protection against PCI failure by providing operational information to operators of a nuclear reactor during power maneuvering, including predictive calculations prior to executing power maneuvers. Additionally, the invention relates to the diagnostics of an event suggesting a possible PCI cladding failure.

Abstract: In the present invention, a reactor power control apparatus of a natural circulation reactor comprises a power control apparatus for controlling reactor power and a pressure control apparatus for controlling reactor pressure. A power adjustment demand signal S4 is input from the power control apparatus into the pressure control apparatus. The pressure control apparatus controls degree of opening of a inlet port steam control valve provided in a moisture separation heater that is connected to a low-pressure turbine, based on the power adjustment demand signal S4. A reactor power control apparatus of the natural circulation reactor can suppress reactor power variation and obtain stable power.

Abstract: Provide the technique for controlling nuclear power plant in which the thermal limit can be brought close to the full limit of operation restrictions by the automatic control, using the simple information outputted shorter cycle than that of the thermal limit.

Abstract: According to one embodiment of a reactor core monitoring system, includes: an information retention portion for retaining a regular cycle and a short cycle as calculation information of reactor core performance data; a signal processing portion for creating heat balance data based on a process signal; a data acquisition portion for acquiring, in a timing of the regular cycle, the heat balance data and reactor core performance data which was calculated in a previous timing of the regular cycle, while acquiring, in a timing of the short cycle asynchronous to the regular cycle, the heat balance data and reactor core performance data which was calculated most recently; and a data calculation portion for calculating new reactor core performance data based on the acquired reactor core performance data and the heat balance data.

Abstract: A nuclear power plant control system (3) is provided with detection units (30a to 30d) which detect phenomena that occurs in a nuclear power plant for each of four systems, a trip control device (20) which starts, in the case where a signal that indicates an occurrence of the phenomenon is input from at least a predetermined number of signal lines out of signal lines of two systems, processing corresponding to the phenomenon, and majority circuits (50a and 50b) which are provided for each signal line of the two systems and each output, in the case where the phenomenon is detected by N or more detection units out of the detection units (30a to 30d), a signal that indicates an occurrence of the phenomenon to a corresponding signal line.

Abstract: An operation monitoring apparatus is provided with a function integration VDU (21) including a normal use system VDU (32) which operates and monitors a nuclear power plant under a normal state, and a safety protection system VDU (31) which shuts down the nuclear power plant under an abnormal state, and also a safety protection system VDU (22) which shuts down the nuclear power plant under the abnormal state, and further, a block device which inhibits an actuation of the safety protection system VDU (31) when the function integration VDU (21) is under an abnormal state, thereby capable of simplifying an operation manipulation in the nuclear power plant, and ensuring a high level of safety of an operation control.