Abstract: More accurate prediction of reactor water quality of a nuclear reactor is implemented. A device for prediction of reactor water quality of a nuclear reactor in a nuclear power plant is disclosed. The device stores a reactor water quality prediction model which is learned using learning data, and with which future reactor water quality is predicted. An explanatory variable of the reactor water quality prediction model includes a value in a predetermined period unit that is generated from data acquired in an operating nuclear power plant. The device generates the value in a predetermined period unit from data acquired in a target operating nuclear power plant, and acquires a predicted value of the reactor water quality in the target nuclear power plant based on the reactor water quality prediction model and the value in a predetermined period unit.

Abstract: A reactor control interface includes a home screen video display unit (VDU) displaying blocks representing functional components of a nuclear power plant and connecting arrows that connect blocks that are providing the current heat sinking path for the nuclear power plant. Directions of the connecting arrows represent the direction of heat flow along the current heat sinking path. If the current heat flow path of the plant changes, the connecting arrows are updated accordingly. Additional VDUs include: a mimic VDU displaying a mimic of a plant component; a procedures VDU displaying a stored procedure executable by the plant; a multi-trend VDU trending various plant data; and an alarms VDU displaying side-by-side alarms registries sorted by time and priority respectively. If a VDU fails, the displays are shifted to free up one VDU to present the display of the failed VDU, and one display is shifted to an additional VDU.

Abstract: Provided is an apparatus for evaluating a response time of a plant protection system. The apparatus includes: a classification module classifying a design process related to a response time requirement of each channel performing a safety function into four operations of a safety analysis operation, a system design operation, a response time analysis operation, and a response time test operation; and an integrated evaluation module determining whether a response time evaluation is proper based on a time t1 derived in the safety analysis operation, a time t2 derived in the system design operation, a time t3 derived in the response time analysis operation, and a time t4 derived in the response time test operation.

Abstract: A production method for producing a circuit optimized to be protected against radiation includes a preliminary characterization stage performed on a reference circuit. The preliminary characterization stage includes the steps of: irradiating the reference circuit a plurality of times; after each irradiation, if one or more reference elements of the reference circuit have failed, locating said reference element(s); and mapping the impact of the irradiations on the reference surface of the reference circuit. The production method further includes an optimization stage comprising the step of adapting the position of at least one optimized radiation-sensitive element on at least one optimized surface of the optimized circuit as a function of the mapping performed on the reference circuit.

Abstract: The present invention relates to a screen sharing system of a nuclear power plant main control room for improving operation safety of a power plant. According to an example of the present invention, a power plant control screen of each operator is shared in real-time among other operators by a simple manipulation such that operators can easily simultaneously check a power plant operation screen of other operators, whereby a mistake of another operator can be detected and manipulated in time, thus improving operation safety of a power plant.

Abstract: An apparatus and method for system safety analysis evaluation is provided, the apparatus including processing circuitry configured for generating a calculation matrix for a system, generating a plurality of models based on the calculation matrix, performing a benchmarking or convolution analysis of the plurality of models, identifying a design envelope based on the benchmarking or convolution analysis, deriving uncertainty models from the benchmarking or convolution analysis, deriving an assessment judgment based on the uncertainty models and acceptance criteria, defining one or more limiting scenarios based on the design envelope, and determining a safety margin in at least one figure-of-merit for the system based on the design envelope and the acceptance criteria.

Abstract: A work planning system includes an event-data acquisition unit that acquires event data indicating an event of a facility, a storage unit that stores therein a plurality of pieces of work data indicating contents of works that can be performed with respect to the facility, and a decision unit that decides each importance degree of a plurality of pieces of work data based on the event data.

Abstract: According to an example embodiment of the present invention, there is provided a method, comprising defining a task category information element, the task category information element being associated with at least one functional requirement and at least one design principle, associating the task category information element with at least one architecture definition information element, associating each of the at least one architecture definition information element with at least one system-level information element, and verifying the system described by the at least one architecture definition information element and associated system-level information elements is compliant with the at least one design principle.

Abstract: Embodiments of the present disclosure relate to systems and methods for controlling power generation plant operations via a human-machine interface. In one embodiment, a method can provide: receiving a user selection of a power generation plant component in a first navigation menu of a human-machine interface (HMI); based at least in part on the user selection, generating a second navigation menu of the HMI; receiving a subsequent user selection of a subcomponent in the second navigation menu of the HMI, wherein the subcomponent is associated with the power generation plant component; generating a control region for the HMI, the control region operable to manipulate the subcomponent; receiving a user input for the control region; and based at least in part on the user input, facilitating manipulation of one or more operations of the subcomponent.

Abstract: A digital protection system includes a process protection system having at least two channels and a reactor protection system having at least two trains. The process protection system includes, in one channel, first and second comparative logic controllers of different types that are mutually independent of each other and that respectively receive process variables as inputs and each outputting comparison logic results. The reactor protection system includes, in one train, first and second concurrent logic controllers of different types that are mutually independent from each other and that respectively receive the comparison logic results as inputs and each outputting concurrent logic results. The reactor protection system includes initiation circuits, each circuit including a plurality of relays connected in series and a plurality of relays connected in parallel.

Abstract: A digital protection system is provided. The digital protection system may include: a process protection system including at least two channels, each channel including a first bistable logic controller and a second bistable logic controller which are independent and different from each other, the first bistable logic controller and the second bistable logic controller outputting bistable logic results; a reactor protection system including at least two trains, each train including a first coincidence logic controller and a second coincidence logic controller which are independent and different from each other, the first coincidence logic controller and the second coincidence logic controller outputting coincidence logic results by receiving the bistable logic results from the process protection system; and an initiation circuit normally operating or stopping a reactor according to the coincidence logic results received from the reactor protection system.

Abstract: Systems and methods to graphically display process control system information are disclosed. Some example methods include monitoring process variables in a process control system, determining a current state of a first one of the process variables, and determining a trend associated with the first process variable. Some such examples further include generating a first graphic representative of information associated with the first process variable, the information comprising the current state of the first process variable and the trend of the first process variable. Some example methods also include rendering the first graphic via a display.

Abstract: According to an embodiment, a neutron measurement apparatus has: a neutron detector; a preamplifier that amplifies an output signal of the neutron detector and outputs a neutron detection signal; a pulse counting unit that measures the neutron intensity by using pulse counting method by which the neutron detection signal is used to count the number of individual pulses; a Campbell measurement unit that measures the neutron intensity by using Campbell method by which a time average of squares of AC component of the neutron detection signal is calculated; a correction constant calculation unit that calculates a correction constant for correcting an output of the Campbell measurement unit by using an output of the pulse counting unit and an output of the Campbell measurement unit; and a correction value calculation unit that outputs, based on the output of the Campbell measurement unit, a corrected value by using the correction constant.

Abstract: A control system allows controlling a nuclear facility in an evacuation area. The control system includes a control device in ordinary use disposed in a non-evacuation area, an emergency control device—for emergency in the evacuation area, a plant control facility connectable to the control device—and the emergency control device, a signal switching unit that switches from a normal coupling to an emergency coupling based on an emergency switch signal, a first selector switch in the non-evacuation area, a second selector switch-in the evacuation area, an AND circuit configured to output the emergency switch signal to the signal switching unit in the case where the emergency switch signal is input from the first selector switch and the emergency switch signal is input from the second selector switch.

Abstract: In a control system for a plant controlling a plant such as a nuclear power plant 1 with use of a plurality of digital control devices 41, the plurality of digital control devices 41 include a plurality of control functions 46 to 56, and the plurality of control functions 46 to 56 are provided in the plurality of digital control devices 41 in a distributed manner so that the digital control devices 41 may not fall below safety standards preset by safety analyses. This provides a control system for a plant using the plurality of digital control devices 41 and configured to control a plant safely even when a digital control device is failed.

Abstract: Provided are a safety protection system facility 43 outputting a normal actuating signal S1 in a case where the safety protection system facility 43 controls actuation of a unit provided in a nuclear facility to a safe side based on an abnormality detecting signal output at the time of occurrence of an abnormality in the nuclear facility, and where this causes the unit to be actuated normally, and a CCF countermeasure facility 44 outputting a CCF-case actuating signal S2 actuating the unit to a safe side in a case where the CCF countermeasure facility 44 determines from output results of the abnormality detecting signal and the normal actuating signal S1 that the unit is not actuated normally at the time of occurrence of the abnormality in the nuclear facility.

Abstract: A method includes partitioning alarms into two or more alarm groups. The alarms are associated with operation of an industrial process system performing a process, and each alarm has multiple states. The method also includes generating a graphical display for presentation to a user. The graphical display identifies at least one of the alarm groups and multiple alarm indicators. Each alarm indicator indicates a state transition of an associated alarm between two of the alarm's states. The alarm indicators are arranged in the graphical display based on times of their respective state transitions.

Abstract: The invention is related to a device for detecting insulation degradation in an inverter-driven load device, in particular a motor, the device including: zero-phase current measuring means for measuring a zero-phase current in power-feed lines, provided in the power-feed lines between an inverter device and the motor; and command control means for putting rotation of the motor on standby; wherein the zero-phase current measuring means measures the total of phase currents fed into respective phases so as not to rotate a shaft even when an external force is applied to the shaft during the rotation being on standby, whereby allowing regular detection of insulation degradation without switching over the power-feed lines connected with the load device.

Abstract: In an operation monitoring apparatus of a nuclear power plant, a first operation console (12) that can be operated by an operator and includes displays (31 to 37), an operation command console (14) that can be operated by a shift supervisor and includes displays (41 to 44) , a transfer device (51) that can display an image displayed on the displays (31 to 37) of the first operation console (12) on the displays (41 to 44) of the operation command console (14), and a switching device (54) that switches an image on the displays (41 to 44) of the operation command console (14) to the image on the displays (31 to 37) of the first operation console transferred by the transfer device (51) are provided. With this configuration, it is possible to improve communications between the supervisor and the operation in operation of the nuclear power plant.

Abstract: A system and method to influence routing of a call by a sequenced application from among a plurality of sequenced applications is disclosed. The method comprises receiving a header for the call wherein the header comprises at least one directive from each of the plurality of sequenced applications. Further the method comprises arbitrating conflicts from among the one directive in the header in order to determine a set of sequenced application headers to at least partially execute. Finally, the determined set of sequenced application headers is at least partially executed.

Abstract: Provided are an operation support device and an operation support method for a nuclear power plant which are capable of accurately grasping the condition of a nuclear power plant and capable of offering support for instantly performing a predetermined operation by an operator when an unusual situation occurs. The operation support device for a nuclear power plant is provided with an operation state check display unit (20) for displaying whether or not a system or equipment necessary for a safe operation of the entire nuclear power plant operates on the basis of a command signal to the system or equipment and a bypass and operation impossibility state display unit (21) for displaying whether the system or equipment is operable or not.

Abstract: Provide a nuclear reactor power monitoring technology for enhancing the monitoring accuracy and reliability in nuclear thermal hydraulic stability.

Abstract: An ex-core nuclear instrumentation system in which the width of measurable neutron detector current can be accurately widened is obtained. In order to output the condition of neutron flux in operation by performing arithmetic processing of a current value measured by a neutron detector by using a detector signal processing circuit, the detector signal processing circuit includes a current/voltage conversion unit which converts the current value converted by the neutron detector into a voltage value corresponding to the current value; and a variable gain amplification unit which has an operational amplifier having a resistance circuit for corresponding to current levels, the resistance circuit being capable of selecting a gain, and a D/A converter that adjusts the gain, and amplifies the voltage value converted by the current/voltage conversion unit.

Abstract: Systems and methods of monitoring rod control systems of nuclear power plants, including: measuring a particular signal applied to at least one coil of a rod movement mechanism at one or more predetermined times relative to an initial energizing of a mechanism coil; comparing the measured signal to a reference signal parameter, and determining if the measured signal deviates from the reference signal parameter by a predetermined amount to indicate degradation of the rod control system.

Abstract: According to one embodiment, a reactor oscillation power range monitor includes: a receiving unit configured to receive LPRM signals issued by LPRM detectors differing in vertical position out of a plurality of LPRM detectors placed in a reactor core; a specification unit configured to specify any exclusion signal which meets an exclusion condition out of the received LPRM signals; an estimation unit configured to estimate an alternative signal for the exclusion signal based on a regular signal which does not meet the exclusion condition out of the received LPRM signals; an arithmetic averaging unit configured to output an arithmetically averaged signal obtained by arithmetically averaging the regular signal and the alternative signal; a time averaging unit configured to output a time-averaged signal obtained by time-averaging the arithmetically averaged signal; and a calculation unit configured to output a standard value obtained by dividing the arithmetically averaged signal by the time-averaged signal.

Abstract: A device for measuring and correcting a parallelism error of an upper plug end of a nuclear fuel rod comprising mechanisms for measuring a parallelism error and for correcting said error, and a mechanism for positioning said device on the fuel rod and cooperating with a rack on which the fuel rod is stored, said correction means being arranged opposite the measuring mechanism relative to the location of the fuel rod, in order to allow a measurement of the parallelism error during correction of said error.

Abstract: A reactor in-core instrument handling system in which the signal leads are routed from the instrument sensors through an outer sheath through the upper reactor internals and out of and around the sheath in a substantially tightly wound spiral before exiting the reactor vessel.

Abstract: Various embodiments enable wireless monitoring of a radioactive environment and related operating conditions. Structures of increasing complexity and formed at least in part from a semiconductor material based on crystalline actinide oxide-based material enable monitoring at least one operating conditions of the radioactive environment. An exemplary embodiment is a device comprising one or more crystal oscillator units that can generate a first oscillating signal, and an actinide oxide-based unit functionally coupled to at least one of the one or more crystal oscillator units, and configured to receive the first oscillating signal. The actinide oxide-based unit can supply a second oscillating signal to an antenna that delivers the second oscillating signal wirelessly, wherein the second oscillating signal is based on the first oscillating signal and is indicative of an operating condition of the radioactive environment. The antenna is part of the radioactive environment.

Abstract: A reactor core coolant temperature measuring apparatus for measuring the temperature of the coolant flowing in the reactor core of a nuclear reactor comprises a ?-ray thermometer having a temperature measuring section arranged below the bottom of the reactor core and a ?-ray heat generation detecting section arranged between the bottom and the top of the reactor core, a cable for transmitting the signal output from the temperature measuring section and an output signal processing means for receiving the signal and computationally determining the local temperature of the coolant at the position of placement of the temperature measuring section.

Abstract: Systems and methods of monitoring position information of a control rod in a nuclear reactor, including detecting an output signal induced by movement of a control rod with respect to at least one detection coil of the control rod, the output signal having identifiable characteristics representative of a plurality of positions of the control rod within the at least one detection coil, and processing the identifiable characteristics to derive the positions of the control rod within the at least one detection coil.

Abstract: A device for detecting steel is provided. The steel is used in a nuclear plant. The device has a detecting circuit of electrochemical corrosion potential (ECP) and alternative current (AC) impedance. An environment in a boiling water reactor (BWR) is simulated under hydrogen water chemistry (HWC). The environment is used for detecting intergranular stress corrosion cracking (IGSCC) of components coated with different precious metals. Thus, effect of coating different precious metals on steels can be evaluated for HWC.

Abstract: A nuclear reactor vibration surveillance system has a first ultrasonic transducer for transmission, an ultrasonic transmitter, a second ultrasonic transducer for reception, an ultrasonic receiver, a signal processor, and a display unit. The first ultrasonic transducer for transmission is arranged on the outer surface of a reactor pressure vessel and is configured to convert a transmission signal into an ultrasonic pulse signal and allow the ultrasonic pulse to be transmitted to a reactor internal component. The second ultrasonic transducer for reception is arranged on the outer surface of the reactor pressure vessel and is configured to receive a reflected ultrasonic pulse reflected by the reactor internal component and convert the received reflected ultrasonic pulse into a reception signal.

Abstract: An object of the invention is to embody a small and inexpensive passage selector, which can be applied by commonly using an index device even if the number of detector passages is changed, and which is easy to make inspection and maintenance. The passage selector of a reactor in-core nuclear-measuring apparatus of the invention includes: a drive motor; an index device that is driven by the drive motor and that makes a rotary output of a predetermined index number; a central rotating shaft that is driven to rotate by the index device and that causes a passage selecting guide tube to be located in opposition to any detector passage; and a speed-increasing and decreasing device that is interposed between an output shaft of the index device and the central rotating shaft, and that adjusts the index number of the central rotating shaft.

Abstract: A system for monitoring a condition of a nuclear reactor pressure vessel disposed in a radioactive environment includes an instrument structured to monitor a condition of the nuclear reactor pressure vessel; a powered wireless transmitting modem disposed in the radioactive environment, the wireless transmitting modem being electrically coupled to the instrument; a receiving modem disposed in the line of sight of the transmitting modem, the receiving modem being in wireless communication with the transmitting modem; and a signal processing unit electrically coupled to the receiving modem, the signal processing unit being structured to determine the condition of the nuclear reactor pressure vessel from the instrument. The transmitting modem is powered by a thermocouple disposed in or on the reactor pressure vessel.

Abstract: The invention concerns a method comprising measurement on a fuel channel (14) of fuel assemblies (8) for nuclear boiling water reactors. The method comprises that: the measurement is done by the use of a non-destructive inductive eddy current measurement method, the measurement is done on a fuel channel (14) which has been used at least a certain time during operation in the core of a nuclear boiling water reactor, the measurement is done when the fuel channel (14) is located in water, the measurement is done on different places on the fuel channel (14), wherein through the method at least the hydride content of the fuel channel (14) at said places is determined. The method can be used in order to find out how shadow corrosion from a neighboring control rod influences the properties of the fuel channel (14).

Abstract: The invention relates to an apparatus and methods for operation in relatively high radiation fields to remotely switch signal devices through a shared single main umbilical signal cable. The invention is particularly suitable for use in a nuclear reactor, such as a boiling water reactor, and in difficult to access areas in the reactor pressure vessel. One or more main umbilical cables connect a control station to an enclosure housing a signal switching device. The signal switching device allows several signal generating/receiving devices, such as cameras and ultrasonic probes, to be controlled by the one or more main umbilical cables.

Abstract: Provided is an automated periodic surveillance testing method and apparatus in a digital reactor protection system, which unites a passive testing means such as a self-diagnosis or an on-line state comparison with an active testing means such as an automatic logic test and the like in the digital reactor protection system. Accordingly, a device self-diagnosis, an on-line state diagnosis, and an automatic logic test are used as the automated periodic surveillance testing method in the digital reactor protection system. Thus, the time consumed in a periodic surveillance test can be minimized, thereby saving reactor operating costs, and excessive manpower can be avoided, thereby decreasing potential human errors. Also, a channel detour time at which one of multiple channels is detoured can be minimized based on the automated periodic surveillance test in the digital reactor protection system, thereby enhancing stability.

Abstract: To provide a rod block monitor in which irrespective of the selection or non-selection of a control rod, such a process is repeatedly executed that gain adjustment is performed at fixed intervals at all times to match a local area average value of signals of neutron detectors surrounding the control rod with a nuclear reactor average power and a block level appropriate to the local area average value of the neutron detector signals after the gain adjustment is set.

Abstract: A motor operated valve diagnostic data acquisition system including: a motor operated valve assembly; a transmitter, the transmitter being permanently connected in a housing of a motor operated valve assembly; a receiver, the transmitter transmitting signals to the receiver via motor operated valve cables, the cables being existing motor operated valve cables; and electronic circuitry connected to the receiver, the electronic circuitry processing the signals received by the receiver.

Abstract: An ultrasonic sensor has a piezo-electric element attached at an end surface outside a reactor pressure vessel (RPV) of a sensor leading edge portion. The sensor leading edge portion passes through a bottom head of the RPV and is installed on the bottom head. Ultrasonic waves generated by the piezo-electric element are propagated to the sensor leading edge portion and are propagated to reactor water in the RPV from the sensor leading edge portion. When water surface of the reactor water in the RPV exists below a core support plate, the ultrasonic waves propagated inside the reactor water are reflected on the water surface. Ultrasonic waves reflected on the water surface are propagated into the reactor water, enter the sensor leading edge portion, and are received by the piezo-electric element. Using the ultrasonic waves received by the piezo-electric element, the water level in the RPV is obtained.

Abstract: A method of operating a nuclear reactor is disclosed. The reactor (1) encloses a core having a plurality of fuel rods (9). Each fuel rod (9) includes a cladding and fuel pellets of a nuclear fuel. The fuel pellets are arranged in an inner space of the cladding leaving a free volume comprising an upper plenum, a lower plenum and a pellet-cladding gap. The reactor is operated at a normal power and a normal inlet sub-cooling during a normal state. The reactor is monitored for detecting a defect on the cladding of any of the fuel rods. The operation of the reactor is changed to a particular state after detecting such a defect. The particular state permits an increase of the free volume in the defect fuel rod. The reactor is operated at the particular state during a limited time period, after which the reactor is operated at the normal state.

Abstract: An in-core nuclear instrumentation apparatus is provided with a neutron flux current detection circuit which converts a neutron flux current signal fed from a neutron detector into a voltage signal and amplifies the voltage signal with a capability to continuously vary amplifier gain. The neutron flux current detection circuit includes a current sensing resistor for converting the neutron flux current signal into the voltage signal, an amplifier for amplifying the voltage signal, wherein an equivalent resistance value of a feedback circuit of the amplifier can be controllably varied, an equivalent resistance control circuit for varying the equivalent resistance value of the feedback circuit, and an output circuit which outputs an output terminal voltage signal of the amplifier.

Abstract: Described is the use of a transponder in a commercial installation, particularly a power plant, which allows problems and dangerous situations existing in this context to be avoided and the attainable occupational safety in a commercial installation, particularly a power plant, to be increased. The transponder includes a reception element, a transmission element, a visual display unit and a memory element and also a control element which is operatively connected to these, in a commercial installation, particularly a power plant, in an arrangement or positioning on or in proximity to a component or at a switching or measuring point associated with the component for the purpose of visually displaying an operating state for the component or the switching or measuring point associated therewith using the visual display unit of the transponder.

Abstract: In an operation monitoring apparatus of a nuclear power plant, a first operation console (12) that can be operated by an operator and includes displays (31 to 37), an operation command console (14) that can be operated by a shift supervisor and includes displays (41 to 44) , a transfer device (51) that can display an image displayed on the displays (31 to 37) of the first operation console (12) on the displays (41 to 44) of the operation command console (14), and a switching device (54) that switches an image on the displays (41 to 44) of the operation command console (14) to the image on the displays (31 to 37) of the first operation console transferred by the transfer device (51) are provided. With this configuration, it is possible to improve communications between the supervisor and the operation in operation of the nuclear power plant.

Abstract: An apparatus for inspecting and testing a startup range neutron monitoring system for a nuclear reactor. The apparatus includes: a neutron-flux detector; a preamplifier that amplifies an electric signal output from the neutron-flux detector; a pulse measurement unit that counts times when electric signal output from the preamplifier exceeds a discrimination voltage; a discrimination-voltage setting unit that applies the discrimination voltage to the pulse measurement unit; a voltage-setting unit that applies a voltage to the neutron-flux detector; an arithmetic processing unit that calculates an output power of the reactor based upon an output signal of the pulse measurement unit; an output unit that outputs data representing the output power of the reactor, calculated by the arithmetic processing unit; and an inspecting/testing unit that sets the discrimination voltage and the voltage to be applied by the voltage-setting unit.

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: An upper detector and a lower detector that face at least one side of a fuel assembly, on which neutrons are irradiated in a nuclear reactor, and detect radiation are set at a predetermined interval in an axial direction of the fuel assembly. Distributions of radiation signals are measured by the upper detector and the lower detector while the fuel assembly and the upper detector and the lower detectors are relatively moved along the axial direction of the fuel assembly. Soundness of radiation signals measured by the upper detector and the lower detector is determined in every measurement by comparing radiation signal distributions obtained by measuring the same portion in the axial direction of the fuel assembly in a multiplexed manner with the upper detector and the lower detector. Thereafter, relative burn-up is calculated by utilizing the measured radiation signals to measure a burn-up profile.

Abstract: The present invention certifies control modules of engineered safety feature instruments for a power plant automatically. The control modules can be tested before storing or operating. The test is done with enhanced testing speed and saved cost. Thus, safety of the control modules is confirmed.

Abstract: An advanced digital rod position indication (ADRPI) system having rod drop monitoring capability. When a reactor trip occurs, the ADRPI system derives the rod drop time from the signals induced in the detector coils by the falling drive rods. The ADRPI system is linked to the reactor trip breaker to provide a reference point to allow accurate timing of the rod drop times.

Abstract: According to an embodiment, a digital process type monitor device includes a plurality of modules and a mother board connected to each of the modules. Each module includes: a base board connected to a connector and having an FPGA for main control and an IPGA for sub board control mounted thereon; and a sub board for a main-machine I/F process, having an FPGA for an I/F process mounted hereon. Each sub board has storage devices for storing man-machine I/F information on the sub board. Each of the FPGA writes transmission data into a predetermined region of a transmission area and has a common transmission protocol to share the transmission data between the respective modules.