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: A moderator temperature coefficient measuring device include a control rod driving device; a reactivity measurement processing unit capable of measuring the reactivity of the reactor; an inflow temperature sensor for sensing the inflow temperature of a reactor coolant flowing into the reactor; an outflow temperature sensor for sensing the outflow temperature of the reactor coolant flowing out of the reactor; an average temperature measurement processing unit for measuring the average temperature from the sensed inflow and outflow temperatures; an power control unit for providing control so that the reactor delivers constant thermal power; and a temperature coefficient calculating unit for calculating the temperature coefficient of the reactor coolant based on the amount of change in reactivity between before and after the insertion of the control rod and the amount of change in average temperature between before and after the insertion of the control rod under the constant thermal power.

Abstract: The present invention provides a method of operating a Boiling Water Reactor, having the steps of analyzing LPRM signals for oscilliatory behavior indicative of neutron-flux-coupled density wave oscillations, determining if oscilliatory behavior is present in the signals; initiating a reactor protective corrective action if the oscilliatory behavior is determined, and in addition, initiating corrective actions if neutron uncoupled oscillations are possible. Detecting the later is performed through analytically determined exclusion zone on the power flow map or by on-line stability calculations for several high power channels.

Abstract: The present invention relates to a method of controlling a nuclear reactor during a transient period. The method includes actuating the steam dump system in response to a temperature error signal and a power mismatch signal.

Abstract: In order to provide a reactor power control apparatus which can maintain a stable water level when reactivity control based on control rod operation and water level adjustment is performed, this invention comprises a turbine control which calculates the load set error signal from the error between the set target generator power value and the generator power that was fed back and outputs to the turbine control apparatus; control rod control in the natural circulation boiling water reactor which calculates the control rod operation signal and outputs it to the control rod drive control apparatus; water level control inside the natural circulation boiling water reactor which calculates the water level set signal and outputs it to the feed water control apparatus; and the switch determining device which selectively outputs one of the control rod control and water level control based on the generator power that was fed back, the reactor power and water level, as well as switching rules and determination values.

Abstract: The present invention relates to a method of controlling a nuclear reactor during a transient period. The method includes actuating the steam dump system in response to a temperature error signal and a power mismatch signal.

Abstract: A method determines a volumetric power distribution of a core of a nuclear reactor using a set of detectors for measuring a neutronic flow provided outside the reactor vessel, and a set of probes for measuring the temperature of the coolant at the outlet of the fuel assemblies. The method comprises the step of determining a first volumetric power distribution using a neutronic calculation code instantaneously solving the diffusion equation and updating the isotopic balance of the core during the fuel depletion based on values of the core operation parameters, and the step of determining a new volumetric power distribution by adjusting the first volumetric power distribution using the measures provided by the sensors for measuring the neutronic flow and provided outside the reactor vessel and by the temperature measuring probes.

Abstract: The method of improving nuclear reactor performance involves generating an operational solution for a nuclear reactor based on a constraint accounting for a problem with operation of the nuclear reactor. The generated operational solution can then be implemented at the nuclear reactor.

Abstract: A system and method is provided for a continual updating of the optimization of multiple operational control-variables during the operation of a nuclear reactor over a plurality of fuel cycles. A networked computer system includes one or more hosts programmed to execute an optimization process to identify and make changes in quantitative values of operational control-variables that result in improved efficiency and operational flexibility. Optimization and updating of operational control-variables may proceed selectively under manual control for inputting specific optimization constraints and reactor state-point information or may proceed autonomously through a repetitive performing of the optimization process based upon a predetermined user-defined strategy stored on the network.

Abstract: In the present invention, a reactor power control apparatus of a natural circulation reactor comprises a reactor pressure vessel which circulates cooling water using the density difference of the coolant inside, a feed water pipe which is connected to the reactor pressure vessel and supplies cooling water, a power control section which controls the reactor power using a control rod. The feed water pipe has an ultrasonic thermometer sensor. Driving of the control rod is controlled by the power control section based on the feed water temperature detected by the feed water thermometer. The reactor power control apparatus can detect the temperature of the feed water and perform drive control of the control rod preferentially, and obtain stable reactor power.

Abstract: Systems and methods for a method for determining a critical effective k at an off-rated core state of a nuclear power plant includes determining, for the off-rated core state a control rod density, a percent core power, a gadolinium reactivity worth, a doppler reactivity worth, and a xenon reactivity worth responsive to a control rod pattern, a reactor power plan including the off-rated core state, and a reference effective k, calculating a change in an effective k from the reference effective k at the off-rated core state responsive to two or more parameters selected from the group consisting of the control rod density, the percent core power, the gadolinium reactivity worth, the doppler reactivity worth, and the xenon reactivity worth, and generating the critical effective k for the off-rated core state responsive to the change in the effective k from the reference effective k.

Abstract: In a method of determining an operating margin to a given operating limit in a nuclear reactor, operational plant data from an on-line nuclear reactor plant is accessed, and reactor operation is simulated off-line using the operational plant data to generate predicted dependent variable data representative of the given operating limit. The predicted dependent variable data is normalized for evaluation with normalized historical dependent variable data from stored operating cycles of plants having a similar plant configuration to the on-line plant. A time-dependent average bias and a time-dependent uncertainty value for the predicted dependent variable data are determined using the normalized historical dependent variable data, and a risk-tolerance level for the on-line plant is obtained.

Abstract: In order to provide a reactor power control apparatus which can maintain a stable water level when reactivity control based on control rod operation and water level adjustment is performed, this invention comprises a turbine control which calculates the load set error signal from the error between the set target generator power value and the generator power that was fed back and outputs to the turbine control apparatus; control rod control in the natural circulation boiling water reactor which calculates the control rod operation signal and outputs it to the control rod drive control apparatus; water level control inside the natural circulation boiling water reactor which calculates the water level set signal and outputs it to the feed water control apparatus; and the switch determining device which selectively outputs one of the control rod control and water level control based on the generator power that was fed back, the reactor power and water level, as well as switching rules and determination values.

Abstract: A method of operating a nuclear power plant includes determining and licensing a maximum power level at which the power plant can be operated subsequent to the beginning of a fuel cycle, and with the power plant being operated at less than its maximum power rating at certain times such as at the beginning of a fuel cycle. The maximum power level is greater than the power level that would be calculated based upon an assumption that the heat flux peaking factor (FQ) and enthalpy rise peaking factor (F?H) remain at their maximum level throughout an entire fuel cycle. The maximum power rating takes advantage of factors such as known reductions in FQ and F?H at certain points in the fuel cycle, the marginal additional capacity of the Balance Of Plant, and the occasional optimization of process parameters such as ultimate heat sink temperature and atmospheric conditions.

Abstract: A method for operating a system having a plurality of modes and interlocks between the modes is provided. The method includes operating the system in a first mode and switching the system to a second mode without going to a standby mode.

Abstract: A diagnostic method and system for a technical installation establish the cause of a fault event. A reliable, flexible and simple possible way of diagnosing an installation is provided by placing state variables that characterize an operating state of the technical installation in a dependency tree and establishing a fault path in the dependency tree by linking change directions of diagnostic parameters contained in the dependency tree.

Abstract: A method of evaluating a containment building design includes the steps of assembling a first database of test results; selecting models (algorithms) for the design to be evaluated; comparing a first set of test results of the model to test results in the first database of test results; establishing uncertainty boundaries for the first set of test results of the model; assembling a second database of test results; determining whether the test results of the second database are within the uncertainty boundaries of the model; and evaluating an actual or proposed containment building design based upon the model when test results of the second database are within the uncertainty boundary.

Abstract: A branch amplifier card for a nuclear reactor control rod drive control system is provided. The control system includes a control processor, a plurality of transponder cards arranged in clusters with each cluster under the control of a branch amplifier card. The branch amplifier card is configured to receive commands from the control processor, send the converted commands to transponder cards under the control of the branch amplifier card and to a downstream branch amplifier card, receive an acknowledge word from transponder cards under the control of the branch amplifier card, add AC voltage threshold level information about the transponder cards under the control of the branch amplifier card to the acknowledge word, permit transponder trouble information attached to the acknowledge word to remain in the acknowledge word, and resend the acknowledge word including the transponder trouble information to an upstream branch amplifier card.

Abstract: A method of operating a nuclear power plant includes determining and licensing a maximum power level at which the power plant can be operated subsequent to the beginning of a fuel cycle, and with the power plant being operated at less than its maximum power rating at certain times such as at the beginning of a fuel cycle. The maximum power level is greater than the power level that would be calculated based upon an assumption that the heat flux peaking factor (FQ) and enthalpy rise peaking factor (F&Dgr;H) remain at their maximum level throughout an entire fuel cycle. The maximum power rating takes advantage of factors such as known reductions in FQ and F&Dgr;H at certain points in the fuel cycle, the marginal additional capacity of the Balance Of Plant, and the occasional optimization of process parameters such as ultimate heat sink temperature and atmospheric conditions.

Abstract: A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. Based on a predetermined generic transient bias and uncertainty distribution in the change in critical power ratio (&Dgr;CPR/ICPR), shifted histograms of fuel rod critical power ratio (CPR) are generated. Ultimately, the operating limit minimum critical power ratio (OLMCPR) of the reactor is directly evaluated from a shifted histogram of fuel rod CPRs for a particular set of initial conditions which result in a probability calculation representing the number of fuel rods subject to a boiling transition (NRSBT) during the transient condition being equal to a predetermined value.

Abstract: A method of operating a nuclear power plant includes determining and licensing a maximum power level at which the power plant can be operated subsequent to the beginning of a fuel cycle, and with the power plant being operated at less than its maximum power rating at certain times such as at the beginning of a fuel cycle. The maximum power level is greater than the power level that would be calculated based upon an assumption that the heat flux peaking factor (FQ) and enthalpy rise peaking factor (F&Dgr;H) remain at their maximum level throughout an entire fuel cycle. The maximum power rating takes advantage of factors such as known reductions in FQ and F&Dgr;H at certain points in the fuel cycle, the marginal additional capacity of the Balance Of Plant, and the occasional optimization of process parameters such as ultimate heat sink temperature and atmospheric conditions.

Abstract: A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. A data processing system is used to execute specific program routines that simultaneously simulate the thermal operating characteristics of fuel rods within the reactor during a transient operational condition. The method employs a multi-dimensional approach for the simulation of postulated operational events or an anticipated operational occurrence (AOO) which produces a transient condition in the reactor—such as might be caused by single operator error or equipment malfunction. Based on a generic transient bias and uncertainty in the change in critical power ratio (&Dgr;CPR/ICPR), histograms of fuel rod critical power ratio (CPR) are generated.

Abstract: A method and a device are provided for monitoring a power rise during startup of a nuclear reactor (diversitary excursion monitoring). In order to be able to shut down the reactor during startup in the event of prompt critical states, a power band situated in a lower measuring range is prescribed for a signal of power range channels which detects the reactor power in the power range. Startup is continued only when the signal does not exceed an upper limit within a minimum time, after the last occasion of exceeding a lower limit of the power band.

Abstract: In a boiling water reactor controlling each control rod hydraulically by driving a solenoid valve, the control system comprises an operation control means 41 having a duplicated data processing unit for generating sequence patterns based on the timing of the driving sequence based on the control information provided manually, a transmission control means 42 for creating a command word corresponding to each control rod being controlled based on said sequence pattern, and mutually communicating each command word between duplicated data processing units and computing the AND logic of said data within a predetermined time difference, and when the computed result coincide, transmitting the selected command word serially; a transmission unit 32 for receiving said command word and performing protocol conversion thereto, and transmitting the same to a plurality of transmission branch units positioned downstream as the control command; and a solenoid valve drive circuit 31 for driving the control rod drive unit corres

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 method and apparatus for controlling components in a component control system in a nuclear power plant includes an image processing system and display processor (IPS-DP) for issuing an encrypted command for a selected component in a component command system (CCS) in the nuclear power plant; a soft controller for receiving the encrypted command and matching the command with the selected component in the CCS for issuing a control command to the selected component; and control channel gateway for receiving the encrypted command from the IPS-DP and soft controller, deciphering the encrypted command, and, if authenticated, issuing a control command for said selected component in the CCS. Security and isolation, while complying with applicable codes and regulations, are assured.

Abstract: A component actuation circuit for an engineered safety features actuation system (ESFAS) includes a plurality of coincidence logic contacts, and a plurality of parallel-connected ESFAS component relays. Each of the coincidence logic contacts closes in response to a command signal. The plurality of parallel-connected ESFAS component relays are connected in series between at least two of the plurality of coincidence logic contacts. Closure of at least two of the coincidence logic contacts energizes the parallel-connected ESFAS component relays, thereby energizing a plurality of ESFAS components.

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 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 maximum linear heat generation rate and a 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 maximum linear heat generation rate and the minimum critical power ratio so determined are compared with thermal limit values set in advance for them. The automatic power regulator system holds a control signal when at least one of the maximum linear heat generation rate and the minimum critical power ratio exceeds the thermal limit values.

Abstract: A control rod coupling assembly for coupling a nuclear reactor control rod to a control rod drive mechanism is disclosed. The control rod drive mechanism includes an index tube and a bayonet head. The control rod includes blades and a tube at the intersection of the blades. The control rod coupling assembly includes a bayonet socket configured to receive the bayonet head, a shaft extending from the bayonet socket through the control rod, and a handle extending from the shaft, the handle movable to rotate the bayonet socket without rotation of the control rod.

Abstract: An interface between a Plant Protection System and Engineered Safety Features in a nuclear power plant is disclosed for continuously monitoring the plant protection system initiation circuit for each remotely actuated Engineered Safety Feature system to effect remedial action in the event that the Plant Protection System generates a ‘trip’ signal. By using actuation inputs from the Plant Protection System and manual, operator implemented inputs, controls are provided for remote equipment components, such as solenoid valves, motor operated valves, pumps, fans and dampers.

Abstract: A method and apparatus for monitoring a source of data for determining an operating state of a working system. The method includes determining a sensor (or source of data) arrangement associated with monitoring the source of data for a system, activating a method for performing a sequential probability ratio test if the data source includes a single data (sensor) source, activating a second method for performing a regression sequential possibility ratio testing procedure if the arrangement includes a pair of sensors (data sources) with signals which are linearly or non-linearly related; activating a third method for performing a bounded angle ratio test procedure if the sensor arrangement includes multiple sensors and utilizing at least one of the first, second and third methods to accumulate sensor signals and determining the operating state of the system.

Abstract: A method of controlling the system pressure in a power generating system, having a turbine-generator and a BWR, that modulates the core thermal power of the reactor while maintaining the main turbine control valves in a constant steady position is described. The core thermal power may be adjusted by adjusting the control rod density within the reactor core or by adjusting the flow rate through the reactor which may be accomplished by modulating the speed of variable frequency recirculation pumps or by modulating recirculation flow control valves. The method includes transferring the power generation system from normal turbine control valve modulation pressure control to core thermal power modulation pressure control. Additionally the method includes modifying the bypass valve closure bias and the power control bias to accommodate the variances from core power modulation pressure control over normal pressure control.

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 temperature measurement system includes, for each of two nuclear steam supply system (NSSS) loops, four temperature sensors each of which provides a value of the temperature of reactor coolant flowing in the NSSS hot leg of the loop; a mechanism for determining, for the hot leg, a first bias value associated with a first pair of the temperature sensors and a second bias value associated with a second pair of the temperature sensors from the value of the temperature of the reactor coolant of each of the temperature sensors; a mechanism for averaging the first bias value and the values of the first pair of the temperature sensors to determine a first average temperature therefrom; and a mechanism for averaging the second bias value and the values of the second pair of the temperature sensors to determine a second average temperature therefrom, whereby two separate average temperatures of the temperature of the reactor coolant are determined for each of the two loops.

Abstract: A digital plant protection system for use in nuclear power plants of the pressurized water type includes cross-connected sensed-parameter processing channels that provide a suitably conditioned digital value to a digital comparator that tests the conditional digital value against a predetermined value to determine whether or not the sensed parameter has been exceeded. A comparator is associated with each of the plural channels and receives a separate measurement of the sensed parameter for each channel. If a sensed-parameter is determined to be out-of-specification by 2 of 4 or more channels, a `trip` signal is generated to effect remedial action.

Abstract: A feedwater control system and method for a pressurized water reactor steam generating system having first and second output signals. The first output signal is determined by first and second input signals, and, when combined with a third input signal automatically controls at least one feedwater pump and first designated valves which regulate water flow from the one or more feedwater pumps to a steam generator when a steam generator steam load and reactor are operating at a first predetermined power level. The first input signal is determined by a downcomer feedwater flow differential pressure. The second input signal is determined by a reactor power level. The third input signal is determined by a steam generator level. The second output signal, determined by a steam generator water level, automatically controls at least one startup feedwater control valve when the steam generator steam load and the reactor are operating at a second predetermined power level.

Abstract: A method and apparatus for monitoring a source of data for determining an operating state of a working system. The method includes determining a sensor (or source of data) arrangement associated with monitoring the source of data for a system, activating a method for performing a sequential probability ratio test if the data source includes a single data (sensor) source, activating a second method for performing a regression sequential possibility ratio testing procedure if the arrangement includes a pair of sensors (data sources) with signals which are linearly or non-linearly related; activating a third method for performing a bounded angle ratio test procedure if the sensor arrangement includes multiple sensors and utilizing at least one of the first, second and third methods to accumulate sensor signals and determining the operating state of the system.

Abstract: A safety or protection system includes a plurality of divisions of first signals; a plurality of safety or protection subsystems each of which receives a corresponding one of the divisions of first signals and produces a corresponding plurality of second signals therefrom; and a reflective memory subsystem for communicating the second signals between the safety or protection subsystems.

Abstract: The present invention is an electronic awareness system for evaluating, monitoring, and controlling, in real-time, an outside environment having entities and occurrences with characteristics. The electronic monitoring system includes input sensors for receiving real-time data produced by the entities, a computer workstation for storing the input data in a database, a computer program having a graphical user interface for interfacing a user with the workstation and the computer program. The graphical user interface comprises a map frame chart function, a worksheet function, a utility function, and a priority navigator function for monitoring and evaluating the input data graphically, tabularly, statistically, and comparatively with a predetermined relative priority scale, respectively, for producing output data. The electronic awareness system also includes output reactors for receiving the output data for controlling certain entities and occurrences of the outside environment.

Abstract: A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles.

Abstract: A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. A data processing system is used to execute specific program routines that simultaneously simulate the thermal operating characteristics of fuel rods (or plural groups of fuel rods) within the reactor during a transient operational condition. The data processing system is also used for compilation of a transient response histogram that incorporates the effect of inherent "uncertainties" in various parameters of interest. In an initial phase, the method employs a multi-dimensional approach for the simulation of postulated operational events or an anticipated operational occurrence (AOO) which produces a transient condition in the reactor--such as might be caused by single operator error or equipment malfunction.

Abstract: A Supervisory Sequential Controller Interface (SSCI) provides for automatic and manual sequencing of the sequential steps of multiple procedures simultaneously in a complex process facility while preserving operator supervisory control over each procedure. The SSCI includes an Executive Interface which gives the operator global control over all of the procedures and a Procedure Interface which is arranged such that the operator may ascertain the relevant information about a particular procedure, including the mode (auto/manual), the current step being executed and the relevant states of plant components and parameters for the current step. Multiple Procedure Interface screens may be displayed to monitor/execute multiple procedures running simultaneously.

Abstract: A functional approach to an information display system that enables a fast and easy understanding of the status of a complex process. The display screen is organized by process goals. Each process goal is then tied to a number of process objectives which directly relate individually to the goals. The objectives are in turn tied to the process functions which must be substantially maintained in order to achieve the respective objectives. A display of control functions related to each process function is mapped onto the regional displays. The individual display elements use a combination of coded background color, border color and other indicia of information, e.g., mimics, meters, etc. to present a significant amount of information about the status of the process and its components in a manner that is readily absorbed and easily understood.

Abstract: An electrical circuit backfits into different instrumentation and control (I&C) system hardware configurations as a replacement or upgrade. The circuit has a multiple signal path processing circuit including a multi-functional ASIC. A fixedly configured, multiple signal path connector such as a hardwired multi-pin package removably connected with a pin socket in the processing circuit connects the ASIC with the I&C system for coupling inputs to and outputs from the ASIC and for selecting operative signal paths of the ASIC.

Abstract: An electrical circuit backfits into different instrumentation and control (I&C) system hardware configurations as a replacement or upgrade. The circuit has a multiple signal path processing circuit including a multi-functional ASIC. A fixedly configured, multiple signal path connector such as a hardwired multi-pin package removably connected with a pin socket in the processing circuit connects the ASIC with the I&C system for coupling inputs to and outputs from the ASIC and for connecting operative signal paths of the ASIC.

Abstract: The present invention relates to an operator work station for nuclear power plants which can provide plant information in an effective way during abnormal/emergency conditions so that the operator can take necessary actions to recover the nuclear power plants. The operator work station for nuclear power plants of the present invention comprises dynamic alarm console; system information console; computerized operating procedure console; system-alarm connecting system which connects the alarm console with the system information console; and, system-procedure connecting system which connects the system information console with the computerized operating procedure console. According to the operator work station, efforts and time required for the operator to judge for himself, and possibility of mistakes of judgement are properly excused, since the operator can easily recognize the system where abnormality occurs, based on the generated alarms.

Abstract: A diagnosis system having sensors for supervising a plant, signal processors for processing signals of the sensors, and a state quantity arithmetic operation unit. The state quantity arithmetic operation unit calculates a quantity of state expressing the environment of the plant on the basis of supervisory information inputted through the signal processors. A state quantity prediction unit predictively calculates a quantity of state after a predetermined time on the basis of the quantity of state, supervisory information and a time change of the quantity of state inputted through the state quantity arithmetic operation unit. A future event prediction unit predicts a future event on the basis of predicted information of the quantity of state inputted through the state quantity prediction unit. An image information processor converts the quantity of state given by the state quantity arithmetic operation unit into image information and indicates the image information on a display unit.

Abstract: Apparatus that provides an integrated approach for all management activities for hazardous substances used or generated at a facility, including form generation and compliance with the reporting requirements. The apparatus includes eight functional groupings and a database schema or coordinator that integrates these groupings and allows them to share and exchange relevant information. The functional groupings include a hazardous materials index grouping (chemical profiles, waste profiles, Material Safety Data Sheets, etc.), a hazardous materials management grouping (MSDS generation, process definition, materials transfers, etc.), a human resource management grouping (training, exposure limits, etc.), a hazardous commitment management grouping (compliance requirements and deadlines), an emergency management grouping (planning for and response to unscheduled releases), and a facility management grouping that can be used by or for response to a regulatory agency.

Abstract: The present invention relates to a technique for estimating core performance which includes the steps of: dividing a reactor core into a plurality of core portions, each consisting of at least one fuel assembly, then further dividing each of the core portions into segments by planes perpendicular to the axis thereof and allocating a parallel processing apparatus having a plurality of arithmetic processing units, a memory for storing processing data of the arithmetic processing units, and communications circuitry for transferring data with respect to the arithmetic processing units, in a manner such that the arithmetic processing units correspond to the segments of the core. Data on the power distribution for the entire core portion to which a corresponding segment belongs is transferred to each arithmetic processing unit, before each of the arithmetic processing units performs thermal hydraulic calculations.