Abstract: A pressurized water reactor (PWR) is protected from departure from nucleate boiling and from hot leg boiling by generating a segmented delta temperature trip function having a first line segment which closely follows the departure from nucleate boiling core limit line and a second segment closely following the hot leg boiling core limit line. Each line segment is a function of the average core temperature and coolant pressure with the departure from nucleate boiling segment further adjusted for axial power distribution. The two set points are continuously compared to actual delta temperature in four independent channels. If either set point is exceeded in at least two channels, the reactor is tripped.

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

Abstract: An information system for a complex plant which clearly presents to the operator, in parallel as alarms, abnormal process changes which the operator can, at his discretion, manually respond to either to prevent conditions from deteriorating further to the point where automatic protective actuations will be initiated or where automatic protective actuations have not been completed because of a malfunction or were completed but have not been successful. Such alarms are displayed on a dedicated alarm display panel which spatially groups components and subsystems to show appropriate relationships, with the alarm messages shown in the associated area of the display so that possible disturbance propagation can be foreseen. Messages related to uncompleted automatic protective actuations may include a list of faults which prevented completion of those actuations.

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

Abstract: A plant condition display system for displaying items of information regarding a plant operation for an operator comprises a process signal input device into which process values are inputted from a plant and converted therein into process signals, a display data preparation device for preparing data for displaying plant conditions in response to the process signals from the process signal input device, and a plant condition information display device for representing process information of the plant in accordance with plant condition display data from the display data preparation device and then displaying the represented process information on a display screen. The plant condition information display device is a display device for displaying plant start up conditions or a decay heat removal operation condition during a plant shutdown.

Abstract: A computer-based power oscillation detection system and method for detecting, monitoring and indicating thermal-hydraulic stability margin in a nuclear reactor is provided. A group of neutron flux detectors is distributed throughout the reactor core, in contiguous relation to the reactor fuel assemblies, with each flux detector providing an output signal indicative of the power density of the portion of the core adjacent to the detector. A computer-based system processes the detector output signals utilizing a period based algorithm that employs an oscillation detection counting function wherein a count corresponding to the level of periodicity of the signal is determined. Since a representative oscillation count for each reactor stability state usually only occurs on the order of several minutes for a single detector signal, a unique simulated decay ratio algorithm reduces the time required to obtain a representative count by processing the group of detector signals spread throughout the core.

Abstract: An enhanced protection system for protecting against transient overpower in a boiling water nuclear reactor which automatically adjusts the reactor over-power protection trip setpoints to be a controlled margin above the operating power level, so that enhanced fuel and reactor protection is provided at all power levels.

Abstract: A control system for avoiding coupled neutronic-thermal hydraulic instabilities during maneuvering of a boiling water nuclear reactor that determines the average elevation in the reactor core of the coolant boiling boundary, and avoids instabilities by maintaining the core average coolant boiling boundary elevation above a target value, which is dependent on reactor design specifics but may be about four feet, in core power-flow operating regions which are subject to instabilities. The system may be implemented without physical modification of existing reactors or reactor instrumentation and control systems through use of the reactor core monitoring computer, which typically has available all core parameters necessary for the computations required in the invention.

Abstract: An uncoupling tool has a two-module design: one module being a driver assembly for developing the uncoupling movement and the other module being a probe assembly for verifying the uncoupled condition. The probe assembly attaches to the threaded end of the piston tube with its probe extending inside the piston tube. The driver assembly attaches to the CRD ring flange. The probe assembly detects the ring magnet in the drive piston using analog position sensors which operate on the principle of the Hall effect and solid state circuitry. A top sensor provides a signal indicating that the control rod is in its "full out" position; a bottom sensor provides a signal indicating that the control rod is uncoupled; and an intermediate sensor provides a signal indicating that the control rod is in a transition position between the "full out" and uncoupled positions. The probe incorporates spacers which focus the magnetic flux onto the sensors.

Abstract: A transient mitigation system and method of operation is provided for a nuclear reactor having a reactor core. A plurality of monitors monitor operation of the reactor. A plurality of independent feedback control systems control respective reactor operations. The transient mitigation system is integrated with the control systems for automatically adjusting a selected one thereof to mitigate abnormal behavior of the monitored operations. A data base contains predetermined membership functions of monitoring and control parameters, each membership function including a plurality of discrete adjacent ranges having respective values. An encoder identifies membership ranges and values for respective singular monitoring parameter values which are evaluated in predetermined fuzzy control rules to select a control parameter and membership range to mitigate the abnormal behavior. A decoder provides a single value mitigating signal for the selected control parameter range to mitigate the abnormal behavior.

Abstract: Excore detector measurements are used to generate on-line absolute reactor power in a pressurized water reactor (PWR) by calibrating detector current measurements to the reactor thermal power calculation made at a base time early in reactor cycle while the thermal reactor power measurement is still accurate. Measurements are also made at the base time of the three-dimensional core power distribution and the core inlet temperature. Present core power measurements are then made by measuring the present excore detector current, the most recent three-dimensional core power distribution and the present core inlet temperature. The present core power is then calculated as the ratio of the present detector current to the detector current at the base time multiplied by the reactor thermal power measurement at the base time.

Abstract: An online system for diagnosing operability of a rotating electrical apparatus includes sensors producing electrical variables corresponding to operating conditions of the apparatus, data converters for converting the electrical variables to digital values, a comparator for comparing the values to corresponding predetermined baseline values of the apparatus and producing a corresponding comparison value, and a signalling mechanism for outputting signals related to a period of predicted operability of the apparatus whenever the comparison value exceeds a corresponding predetermined deadband value. The operating conditions may be non-electrical operating conditions, such as a condition of a lubrication system or a bearing of the apparatus. Alternatively, the sensors may sense electrical insulation non-thermal parameters during operation of the apparatus, in order to produce signals related to the operability of an insulator of the apparatus.

Abstract: A method and a device for detecting oscillations in the core of a boiling-water nuclear reactor (BWR) comprising a plurality of neutron detectors, wherein instability is detected based on oscillations in the output signals of the neutron detectors. For each one of a number of selected neutron detectors, oscillations in the output signals of the neutron detectors are detected. The oscillations are detected on the basis of an oscillation signal (SS) which indicates that an oscillation criterion is fulfilled. An alarm signal for remaining oscillations Ka is generated if the oscillation signal (SS) remains during one delay interval (T3). A first alarm signal (LR) is generated if the oscillation signal (SS) indicates that the oscillation criterion is fulfilled at least once, during each of a predetermined number of consecutive alarm intervals (T1).

Abstract: A method and system for monitoring an industrial process and a sensor. The method and system include generating a first and second signal characteristic of an industrial process variable. One of the signals can be an artificial signal generated by an auto regressive moving average technique. After obtaining two signals associated with one physical variable, a difference function is obtained by determining the arithmetic difference between the two pairs of signals over time. A frequency domain transformation is made of the difference function to obtain Fourier modes describing a composite function. A residual function is obtained by subtracting the composite function from the difference function and the residual function (free of nonwhite noise) is analyzed by a statistical probability ratio test.

Abstract: A method and system for processing a set of data from an industrial process and/or a sensor. The method and system can include processing data from either real or calculated data related to an industrial process variable. One of the data sets can be an artificial signal data set generated by an autoregressive moving average technique. After obtaining two data sets associated with one physical variable, a difference function data set is obtained by determining the arithmetic difference between the two pairs of data sets over time. A frequency domain transformation is made of the difference function data set to obtain Fourier modes describing a composite function data set. A residual function data set is obtained by subtracting the composite function data set from the difference function data set and the residual function data set (free of nonwhite noise) is analyzed by a statistical probability ratio test to provide a validated data base.

Abstract: In a nuclear reactor the feedwater control system receives steam generator water level signals, including information to determine signal quality. The signal quality and additional signal validation determine how the water level signals are to be used by the feed-water control system. In the preferred embodiment, each water level signal is compared with a possible range and the difference or differential between all possible pairs of the signals is calculated and compared with an acceptable differential range. While alarms are sounded if any of these checks on the water level signals indicate an invalid signal, the water level signals are used to calculate a median level signal until there is less than two water level signals with good quality within the possible range or all differentials are outside the acceptable differential range.

Abstract: The use of alarm indication on the overview (IPSO) display to initiate diagnosis of challenges to critical functions or unavailability of success paths, and further alarm-based guidance toward ultimate diagnosis.

Abstract: Through a localized compact workstation (104), the operator is offered all the monitoring, alarming and control functions which have been previously dispersed over many control panels located throughout the control room (100). The information display hierarchy is composed of an IPSO overview display (122), video display units (108) preferably based on CRT technology and qualified video display units (116) preferably based on solid state flat panel display technology. Associated with each display unit is an individual, independent control module to allow process and component control. One control module (112) supports non-safety control functions and avoids separate controllers for components and process control. Another control module (118) is a qualified system which supports safety system control functions and avoids separate controllers for component and process control.

Abstract: The alarm activation set point and priority for a given, spatially fixed alarm tile can vary depending in part on the mode of plant operation.

Abstract: This invention relates to an automation system for nuclear power plants. This system comprises operation plan making means, supervisory control means, and control means for controlling controlled objects. The operation plan making means makes a new operation plan to restore a normal operating condition which existed before the plant condition deviated from the normal operation range. The supervisory control means outputs control commands according to an operation plan under abnormal condition when the plant operating condition deviates from the normal operation range. Furthermore, the supervisory control means outputs control commands according a new operation plan prepared by the operation plan making means after an operation according to an operation plan under abnormal condition has been finished. The control means for controlling controlled objects controls the controlled objects according to control commands output from the supervisory control means.

Abstract: An integrated information system is provided for a plant with interactive processes running in functional equipment subsets, such as a nuclear power generation plant. Sensors are operatively coupled to monitor processes and equipment in the plant, collecting sample data for assessing operational conditions and for predicting maintenance requirements based on loading of the equipment. One or more processors access the sample data and compares present conditions to diagnostic specifications, technical specifications and historical data stored in memory and indexed to equipment subsets and functional operating groups. The processor(s) generate prioritized reports to alert users to potential operational and/or maintenance problems. In addition to the prioritized reports, the processor accesses and outputs to the users reports of the diagnostic and technical specifications applicable to the process parameters exhibiting the potential problems.

Abstract: An intelligent education and simulation system which is capable of executing an optimized follow-up reeducation to the learner reflecting his/her idiosyncrasy toward understanding in learning. The intelligent education and simulation system has an execution instruction to execute a curriculum comprising a plurality of instruction courses regarding the subject teaching of an educational object and its simulation-based instruction. According to this execution instruction, intelligent computer assisted instruction of the subject teaching and the simulation-based instruction are implemented. The degree of understanding of these instructions by the learner is evaluated, and according to the degree of understanding thus comprehended, a pertinent follow-up instruction course(s) is chosen for reeducation. In this way, a pertinent reeducation instruction course(s) optimized for each learner is capable of being selected, and the time required for reeducation of the learner is minimized.

Abstract: Decreasing the speed of a pump supplying cooling water to the core in the boiling water type nuclear reactor based on a trip signal that is generated when a pump supplying feed water is tripped. This performs a first operation to decrease the nuclear reactor power to a first fixed value. A second operation lowers the nuclear reactor power to a second fixed value that is lower than the first fixed value and that prevents run-out of other first pumps in operation. The first fixed value is lower than the reactor power at the moment when the first pump is tripped. The steam flow is decreased substantially simultaneously with the decrease of the feed water flow. Thereby, the decrease of the steam flow is carried out in an early stage when the feed water pump is tripped and the possibility of nuclear reactor scram is reduced.

Abstract: The invention provides integration of the automatic semiautomatic and manual modes on a single man-machine interface device (200,400), with the master control of a system parameter such as pressurizer level, which is responsive to at least two system components (charging pumps, letdown valves), each of which can adjust a different system variable. The setpoint for automatic master control, is selected and displayed (216) on the device, along with the current value (212) of the setpoint parameter (pressurizer level). In the semiautomatic master mode selectable by the operator on the device, either a setpoint (272) or an output demand (282) can be specified for one of the components (206), and the master controller will then adjust the other component (204) accordingly. When all subgroup controls are in the manual mode, the master controller is completely overridden by operator control of each component or subgroup.

Abstract: An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually.

Abstract: An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually.

Abstract: An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually.

Abstract: Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame.

Abstract: A system for a process facility monitors the execution of a procedure in which steps are defined with associated conditions. Each condition is one of an initial condition, sequential condition or constraining condition. While constraining conditions are checked for violations during several steps, sequential conditions are checked only during an associated step. The sequential conditions may be transformed into constraining conditions for subsequent steps by setting a transform flag either prior to execution of the procedure or during execution of the procedure in response to an operator request. Once a transformable sequential condition has been met or the step associated therewith has been completed, an enable flag is set making the sequential condition a transformed condition which is checked at the same time as the predefined constraining conditions.

Abstract: A man-machine system connected to a plant has a storage device for storing data of an ideal operator model and recognizing section for inputting voice information generated by the operator and information about the operator's line of sight. When information about a failure in the plant is input into the man-machine system, plant data or operation information provided at the corresponding locations on an operational panel recognized by the recognizing section according to voice information or information about the operator's line of sight is selected and shown to the operator, so that the operator can take remedial measures for the failure quickly.

Abstract: An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually.

Abstract: An improved method for monitoring the parameters of a nuclear core reactor, analyzing the monitored parameters using modified two neutron group diffusion equations that have been subjected to space-time factorization such that the factorized shape functions are approximated, and determining the core neutronics in constant time steps, for example, about 0.25 seconds, thereby to provide analysis and simulations in full range of core neutronics in real-time continuously.

Abstract: An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

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

Abstract: In an electrical safety system primarily for controlling the insertion of absorber rods into a nuclear reactor core in response to a trip signal, two groups of switches are connected between a power supply and the rod magnets. One group of switches is controlled by signal converters which receive input signals from respective guardlines. The other group of switches is controlled by opto-isolator/photo-detector links which also receive the input signals from the guardlines. If fault conditions in the reactor are sensed by more than one of a group of fault sensors, the feeding of the input signals to the signal converters and the optical links is inhibited, and the switches open, thereby deenergizing the magnets.

Abstract: To detect core-wide and regional neutron flux oxcillations in a nuclear reactor core induced by thermal-hydraulic instabilities, local power range monitoring (LPRM) strings radially distributed throughout the core and having plural vertically spaced neutron flux detectors are locally assigned to individual oscillation power range monitoring (OPRM) cells radially distributed throughout the core. Groups of OPRM cells are assigned to different OPRM channels based on their geographical positions. Detector signals of the LPRM strings assigned to each OPRM cell are processed pursuant to a unique trip algorithm to detect neutron flux oscillations, and, upon meeting prescribed amplitude and frequency criteria, the assigned OPRM channel is tripped. Suppression of a thermal hydraulic instability is initiated when at least two OPRM channels assigned to geographically adjacent OPRM cells are tripped.

Abstract: A reactor core monitoring system and method wherein: by solving the heterogenous neutron diffusion equation in an infinite lattice consisting only of the fuel segment under consideration of the fuel segments arranged in a reactor core, the local power distribution in infinite lattice and the R factor in infinite lattice used in the critical quality calculation are found beforehand, the neutron diffusion equation is solved by homogenizing each of the fuel segments in a region defined in the core and consisting of the selected fuel segment and the fuel segments surrounding the selected fuel segment. Using this solution, the local power distributions are found by correcting the previously found local power distribution in an infinite lattice consisting only of the selected fuel segment. The R factor for the selected fuel segment is then corrected, and the critical power ratio is determined using this corrected R factor.

Abstract: A method of displaying data corresponding to a procedure is provided where all of the data required to follow the procedure is displayed together with abbreviated text corresponding to the procedure. Text and data corresponding to only a portion of the procedure is displayed at any given time. Additional information related to several consecutive steps of the procedure which appear on separate screens is displayed continuously to one side of the screen. Additional data may be selected to displayed prior to returning to the original screen.

Abstract: A control system for a nuclear reactor using an algorithm having a time cycle which is short compared with the delayed neutron lifetimes in the reactor. The use of a "stepping drive" or actuator to change an operating parameter by a fixed unit permits changes in the reactivity of the reactor. The short time cycle of the algorithm allows an evaluation of the reactor and actuation of the stepping drive or actuator during a very short period of time. The algorithm includes digital topology permitting the use of non-linear response functions in the determination of the required change in reactivity or period.

Abstract: A computed model of reactor power output distribution in space is read periodically to microprocessor based computer memory and retained in memory in a three dimensional matrix. This retention occurs between regular updates on the order of every two minutes. The reactor is conventionally monitored in groups of 16 fuel bundles each. Each 16 bundle group is monitored as to its thermal neutron flux by four vertical strings of local power range monitors, each string having one of four power monitors disposed at four different elevations extending the height of the active core. Each bundle group is controlled by four control rods and is assumed to be subject to uniform flow change with overall reactor flow change. The automated thermal limit monitor (ATLM) takes as inputs all power range monitor information from the BWR reactor core on a continuous basis to two channels one channel for determining operating limits the other channel for determining safety limits.

Abstract: Control clusters, each of which comprises a plurality of neutron-absorbing rods, are operable by means of electromagnetic mechanisms whose coils are powered by power bridges driven from a control desk (11). The control system includes a controlling hard-wired logic circuit (12) serving the desk and supervising a plurality of hard-wired logic interfaces each of which serves a subgroup (10) of electromagnetic mechanisms for a determined number of clusters. The static power assemblies (9) each including a hard-wired logic interface and the power bridges of the subgroup(s) (10) served thereby are successively and independently replaced by at least one electronic power assembly (23) constituted by a computer and the power bridges of a subgroup served thereby.

Abstract: Method for determining a evaluating the capacity of a pressurized water nuclear reactor to return to power rapidly, consisting in calculating a pick-up power by applying the general formula:Prip=Prel+.DELTA.Pin which Prip is the maximum pick-up power obtainable by acting on the control bars, Prel is the relative power delivered by the core, determined by the in-line measurement of the neutron flow emitted by the core and .DELTA.P is the additional power resulting from the potential reactivity of the assembly of control bars of the reactor when this assembly passes from the observed insertion to the minimum insertion, P being calculated from the measured position of the power regulation clusters, involving the axial power distribution measured (C1 to C6) and the previously defined exhaustion of the core.

Abstract: The emergency shutdown margin is determined and evaluated by calculating a shutdown margin with the general formulaMar=.DELTA.Pg+.DELTA.Pp-.DELTA.Pr,where Mar is the shutdown margin being sought, .DELTA.Pg is the negative reactivity attributed to the full set of the reactor's control rods when that set moves from withdrawn to fully inserted position, .DELTA.Pp is the reactivity brought about by the change from nominal power to zero power and .DELTA.Pr is the negative reactivity already consumed to bring the reactor into the situation it is in at the instant under consideration and is calculated from the negative reactivity derived from the measured position of the power control clusters (20), taking into account the axial power distribution as measured (C1 to C6) and the depletion of the core as defined in advance, from a term evaluating the reactivity contributed by the drop in power from nominal power to the current situation of the reactor and from a scale constant.

Abstract: A safety system controlling extremely unlikely disturbances in high temperature reactors in addition to and independent of the normal reactor protection systems. Accident instrumentation located in the high temperature reactor monitors certain characteristic process parameters (hot gas temperature, cold gas temperature, cooling gas pressure), for values of which clearly exceed limiting values of the reactor pressure system. The measured data is evaluated electronically and power supply to the cooling gas blowers, feed water pumps and absorber rod holding devices is interrupted if predetermined limiting values are exceeded. The power supply interruption is further actuated by temperature and pressure sensitive devices located in the nuclear reactor. A manually operated emergency switch is provided to turn off the aforementioned sensibilities.

Abstract: The present invention is a system that updates a pressurized water reactor core model, or data file describing the core, periodically using measured values determinable from core instrumentation. The system starts with a calibrated model calibrated from a flux map and a boron concentration measurement both determined monthly. Reactor system measurements are performed and the model is updated from the measurements using a standard depletion calculation. Some of the measured values are also used to determine measured actual values of axial offset and pinch. Standard neutronics equations are used to calculate analytical values of axial offset and pinch from the model, and the measured and calculated values are compared. If a deviation beyond a certain magnitude is detected, buckling coefficients in the model are adjusted to compensate for model drift. The updated model is stored and a time-out period is entered until it is time for another drift compensation cycle to start.

Abstract: A plant malfunction diagnostic method is characterized by determining by simulation a change in a plant state variable, forming a pattern among plant state variables obtained by autoregressive analysis of the change in plant state variable, inserting the formed pattern among the plant state variables in a neural network, performing learning until a preset precision is obtained, and identifying the cause of the malfunction by inserting, in the neural network, a pattern which indicates the pattern among plant state variables formed by data gathered from the plant. This makes possible early identification of the cause of a malfunction. Plant rate of operation and safety are improved by allowing the operator to perform the appropriate recovery operation with a sufficient time margin.

Abstract: An expert system is used to perform surveillance, diagnosis and prognosis of a plant such as a nuclear reactor used in electrical power generation. The expert system uses one of current actual plant data, previously stored actual plant data or simulated plant data as plant condition data to perform an evaluation of operating conditions in the plant. When current actual plant data is used, a data acquisition system supplies the plant condition data. The expert system accesses the plant condition data, regardless of its source, periodically as a surveillance frame is triggered by the rules in the root frame controlling the expert system. Other frames in the expert system perform diagnosis and prognosis of the plant condition.

Abstract: A control rod drop monitoring method and apparatus for use in a nuclear reactor including dedicated timers and displays to measure and display the drop time for each control rod when a reactor system trip occurs. The timers and displays may be made from CMOS technology, and may have millisecond resolution, if desired. The timers are initiated upon receipt of one or more reactor trip signals and are halted at the closure of switches for detecting the fully inserted or other preselected position of the control rods in the reactor core. The monitor system is designed for continuous operation. Reactor trip signals and limit switch signals are fed to latch circuits following a trip event or limit switch activation and are unaffected by subsequent changes in the state of the limit switches or trip signals. A manual reset is required to clear rod drop times. Also, a test mode is provided wherein all of the timers can be started and stopped, thereby providing a functional test of 100% of the monitor circuitry.

Abstract: A nuclear reactor coolant system pressure/temperature system based on real-time data is disclosed. An input subsystem receives and performs initial processing of real-time plant analog and digital signals, and transmits same to a processing subsystem which develops real-time, nuclear reactor coolant system pressure/temperature limits. The real-time, pressure/temperature information is transmitted to an output subsystem for display purposes, for recording operating results, and for control and/or protection system setpoint generation.

Abstract: An apparatus for measuring nuclear reactor power distribution comprises a Fourier coefficient calculation means and a transformation means, whereby Fourier coefficients of neutron detector outputs are transformed into Fourier coefficients of integrated values of the reactor power on each interval, and hence conditions required of data for initial calibration are made less severe and the initial calibration is enabled to be performed much easier. The apparatus further comprises an abnormal data deletion means, whereby detection and deletion of abnormal data are made easy so that the apparatus performs a stable operation with high accuracy.