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

Abstract: 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: 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: 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 method is disclosed for controlling a PWR comprising a plurality of control bars selectively and individually positioned in the reactor core. The current value of an operating parameter indicative of the power level (for instance the average temperature of the core) is measured as well as the current positions of the control bars and the axial power offset in the core. When the difference between the current value of the operating parameter and the set value exceeds a predetermined deadband interval or the axial power offset in the core exceeds a reference value, the amount by which the reactivity of the core should be modified is computed. Through a simulation process on a model of said reactor, it is determined which ones of the bars may be moved and which are the amounts of travel to be given to the bars for bringing the difference within the deadband and the offset toward the reference value while minimizing a core enthalpy increase factor.

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: The present invention is a computer based system supporting plant operators in caring out nearly routine operations. The computer based system monitors available plant instrumentation signals and processes the gathered information to detect satisfaction of initial conditions to the state change. Once the state change is started, the system then compares the sequence of observed changes in the plant with a preprogrammed sequence and alerts the operators to any undesirable deviations from the preplanned sequence by providing appropriate attention grabbing displays on a system monitor. The system also continuously monitors constraining conditions and alerts the operator when the system moves toward an undesired state. The system requires no input from the operator when the operation underway is following the prescribed sequence and minimal input from the operator when a deviation is detected. The system internally and automatically tracks the evolution of plant states during the nearly routine operations.

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 display, a display method and an apparatus are disclosed which produce a summary display depicting function states using discrete state bars centered in one window of a two-window display. The other window on the summary display includes meters for the parameters along the active path of the status tree for the most relevant function as determined by the operator. Each meter indicates not only the current value of the parameter being monitored but the ranges of the different function states as they correspond to the parameter being displayed by the meter. Each meter also carries a data quality indicator. A second level display is provided which depicts the status tree with the active path highlighted allowing the operator to review the decision made by the status tree.

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 display, a display method and an apparatus are disclosed which produce a summary display depicting function states using discrete state bars centered in one window of a two-window display. The status bars represent the states of functions in a process as determined by status decision trees. The discrete states are emphasized by reference lines which indicate the state of each function. Priority among functions is indicated by vertically ranking the bars with the most important function bar appearing at the top. Each bar includes the name of the function, the state of the function and a procedure name for a procedure which can be used to solve the problem identified. The discrete function state is also indicated by the width of the bar and the bar color. On both sides of each bar, a data quality indicator can appear when the quality of the data is less than the best.

Abstract: It consists of determining by an iterative calculation the neutron flux .phi.(j,g) at each point j of the core in the form of a sum of two components, the first component being associated with a predetermined state of the core defined by predetermined interaction probabilities of the neutrons in the core and the second component being linked with the real state of the core, with said predetermined state and with said first component. The power given off P(j) and the number of emitted neutrons S(j) at each point of the core are then deduced from the neutron flux. The power given off P(j) and the neutron flux .phi.(j,g) are used for controlling the core regulating means, such as the control rods or the boron concentration.

Abstract: An interactive display apparatus and method for a plant operating at a plurality of process parametes from which a current plant status can be calculated, and from which symptoms can be diagnosed which indicate the existence of transient conditions of plant status for which mitigating action must be taken within a time interval, utilize a central computer for storing all appropriate mitigating actions which should be taken upon the occurrence of a symptom. The computer is also programmed with algorithms for selecting and prioritizing mitigating actions that are appropriate. A display is connected to the computer for displaying the list of mitigating actions which are appropriate for a symptom diagnosed by the central computer. Sets of symptom indicators and extreme action indicators are also driven by the computer. Each symptom is represented by a separate card which can be independently illuminated to indicate the existence of that symptom.

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

Abstract: A hybrid parallel/serial alarm management system in which sensor signals are intelligently processed to produce abnormality indication signals. The abnormality indication signals are divided into groups defined by functions of the system being monitored and are used to generate abnormality messages according to goal violation, process disturbance and process unavailability requirements for each function. Within each goal, process and support category the messages are ranked according to a predetermined local priority and chronologically within the same priority. Each function includes a spatially dedicated parallel display location which displays the highest priority portion of each category of messages associated with each function. The messages not displayed on the parallel display are displayed on a serial display unit in ranked order by category within function.

Abstract: An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

Abstract: A computer controlled system and method accurately and rapidly verifies the in-place, dynamic response of a transducer/sensor to a time dependent state variable. The verification system includes a programmable ramp generator which introduces a time varying signal simulating a time dependent state variable, e.g., pressure, temperature, flow rate, etc., to a protection system such as used in a nuclear reactor power plant. The verification system further includes means for measuring and displaying the response of the protection system to input signal variation over time. The verification system and method ensures that the protection system outputs an appropriate control signal, such as a shutdown signal, for terminating nuclear reactor power plant operation when the magnitude or time dependent variation of the state variable is not within predetermined limits.

Abstract: A system which measures direct current drift and noise in sensor signals from redundant sensors utilizing a parity-space algorithm. Parity vector signals produced by the parity-space algorithm are averaged to provide a direct current (DC) drift signal. An instantaneous noise signal is found by subtracting the direct current drift signal from a parity vector signal for one of the samples of the sensor signals. The RMS value of the instantaneous noise signals are averaged to provide sensors noise signals.

Abstract: The present invention is a multimode sensor system that transmits power down a common bus coaxial cable typically using an alternating current power source. Each remote unit connected to the coaxial cable and through an isolation transformer converts the alternating current power to direct current power for an integrated circuit bus interface. The interface is connected to the sensors. The interface is externally pin programmable to provide a carrier at a frequency for a channel assigned to the remote unit. The carrier is provided by a ripple counter producing a frequency divided signal compared to a fixed reference frequency, where the result of the comparison controls a voltage controlled oscillator. When plural low frequency analog signals are to be transmitted over the common bus, an on-chip multiplexer multiplexes the signals to an off-chip, external analog-to-digital converter.

Abstract: A method and apparatus for controlling the stability of a boiling water reactor using a digital computer to calculate on-line, from distributed steady state values of only power, flow, enthalpy and pressure; a stability index for selected fuel assemblies taking into account nuclear feedback as well as detailed hydrodynamic effects. Such calculations are only made for fuel assemblies selected as most susceptible to instability on the basis of the level and axial distribution of power generated. If the least stable fuel assembly is unstable, its stability index is iteratively recalculated using assumed incremental changes in either flow or control rod position until a stable condition is indicated. The cumulative adjustment in flow or rod position required for stability is reported to the operator and can be implemented manually or by an automatic control system.

Abstract: This invention provides a method and an apparatus for controlling a polymerization reaction. The temperature difference between a target temperature, calculated from the degradation rate of a polymerization initiator, which varies as a function of temperature, and a polymerization reactor temperature is calculated. The difference in polymerization rate between a target polymerization rate previously set and a polymerization rate calculated from a measured heat release value at a certain time in the polymerization process is calculated. The temperature inside the reactor is finely adjusted on the basis of the temperature difference and the difference in polymerization rate.

Abstract: Closed-loop controller for altering the power level of nuclear reactors in a safe manner and without overshoot. Apparatus is provided for moving a control element such as a control rod or for adjusting the concentration of a soluble neutron absorber (chemical shim) for altering the nuclear reactor power level. A computer computes at short time intervals the two functions [.rho.(t)-.vertline..rho..sub.c .vertline./.lambda..sub.e '(t)]/.vertline..rho..sub.c .vertline. and .tau.(t)ln(P.sub.F /P(t)). The direction of motion of the control element is altered when these two functions become equal thereby resulting in the attainment of a new power level without overshoot. In a preferred embodiment, these two functions are computed at intervals of approximately one second.

Abstract: A signal processor which applies non-linear dynamic compensation to an applied analog signal is tested by applying to a reference ramp signal compensation having a transfer function equal to the inverse of the transfer function of the signal processor. This test signal is applied to the signal processor in place of the sensor signal so that after the compensation of the processor is applied to it, the resultant signal should match the reference signal. When the signal processing is carried out digitally in a microcomputer and the response to the test signal is multiplexed back to the tester along with response signals from other microcomputers, variable time skewing of the returned test signal is eliminated by feeding the reference signal through a processing path parallel to that of the test signal and then comparing those two signals in the tester.

Abstract: A monitored nuclear reactor parameter, such as a startup range neutron monitoring signal, is filtered at a filter having a preselected time constant and amplified with the output being passed to one channel of a two channel comparator. The level of the amplification and the filter time constant are chosen to discriminate between normally increasing startup neutron density including the prompt-jump effect and exponentially increasing neutron density due to the occurrence of reactor casualty in view of the design parameters of the reactor. This same startup range neutron monitoring signal is passed unprocessed to the other channel of the two channel comparator. Comparison of the two signals is made with the binary output of the comparator flagging excess of the unprocessed signal over the filtered and amplified signal. Presence of the appropriate reactor flag is typically used to cause rod withdrawal blockage and/or SCRAM.

Abstract: A rod position indicating system for a pressurized water reactor generates within containment redundant sets of digital detector signals and applies both sets to each of two communications buses each controlled by a separate bus controller/serial output device located inside of containment. A pair of redundant serial data links transmits the redundant data from each bus controller/serial output device through containment to a dedicated CRT display and the plant computer each of which generates rod position displays from redundant information selected from one of the data links. The data for each rod is selected individually from the redundant signals received on the one data link. Rod position is presented in bargraph form on a multi-page display which includes overall status information along with the detailed presentations.

Abstract: The present system uses measurement inputs which reveal fission product barrier functional characteristics for emergency event classification in nuclear generating stations. The process determines the functional status of barriers to fission product release under transient or emergency conditions. The first part of the present process specifies which key indicators within a nuclear plant provide data indicative of fission product barrier status, and develops a logic table or diagram relating specific key symptoms to barrier status such that a computer identifies the functional status of each fission product barrier. In the second part of the present process nuclear power plant operators utilize the computer to determine emergency event classification.

Abstract: A supervisory operation control system responds to detection of vibration occurring at a predetermined location of an equipment for causing the equipment susceptible to adverse influence of the vibration to perform a predetermined supervisory operation. Transition to the supervisory operation is enabled when a product of speed and displacement of the vibration exceeds a predetermined value.

Abstract: On-line, real time monitoring of core-wide and local power in an advanced pressurized water reactor is provided by a system having four channel sets of fixed incore nuclear detector strings distributed in fuel assemblies across the core in a pattern which produces a uniquely identifiable detector response to rod movement even with one channel set out of service. Separate signal processors for each channel generate totally independent core-wide overpower protection signals and exchange detector responses through fiber optic data links for local power monitoring. The local power density at each detector level for any desired point across the core is determined by fitting a surface spline function to the detector responses adjusted to eliminate the local effects of rod movement from a reference condition, interpolation of the fitting function, and factoring back in the local effects of rod movement.

Abstract: Closed-loop controller for altering the power level of nuclear reactors in a safe manner and without overshoot. Apparatus is provided for moving a control element such as a control rod or for adjusting the concentration of a soluble neutron absorber (chemical shim) for altering the nuclear reactor power level. A computer computes at short time intervals the two functions [.rho.-.vertline..rho..vertline./.lambda..sub.e ]/.vertline..rho..vertline. and .tau..sub.i 1n(P.sub.F /P.sub.i). The direction of motion of the control element is altered when these two functions become equal thereby resulting in the attainment of a new power level without overshoot. In a preferred embodiment, these two functions are computed at intervals of approximately one second.

Abstract: Process and device for the detection of the variations, due to a fault in the position of the core control clusters, in the reactivity of the core of a reactor. A momentary value (6) representing the reactivity calculated directly from the core neutron power (2) is compared to a momentary value (30) representing the reactivity calculated from the momentary mean temperature (25) in the core and the mean boron concentration (29) in the primary fluid passing through the core. The invention applies to pressurized water nuclear reactors.

Abstract: Process and device for the detecting of faults in the distribution of the power of the core (2) of a pressurized water nuclear reactor, in which at least one parameter representing the power of the core (2) is measured, each parameter being measured at a determined number of points. For each parameter, the difference between the two extreme measured values and the ratio of this difference to the smallest measured value are determined, and this ratio is compared to a set value, a fault being detected if this ratio is greater than the set value. The sensors (4, 9) are arranged symmetrically close to the periphery of the core (2) and a further sensor (9) is placed close to the center of the core (2).

Abstract: Method of monitoring a nuclear power station by gathering information, called parameters, relating to the operation and state of the power station and its auxiliary equipment, comparing these parameters with reference data corresponding to normal operation, and signaling the differences, in order to permit an operator to evaluate the state of the power station very quickly, and to undertake any needed steps promptly.

Abstract: The fissile inventory required in operating a negative power coefficient nuclear reactor in an electric power generating system is reduced by cycling the load imposed on the system when 100% power can no longer be maintained at equilibrium due to xenon poisoning in order to induce an oscillation in the xenon concentration which is in antiphase wtih the power requirements so that 100% power can be maintained at least during part of the day. The load can be progressively reduced by a preset amount each night or a xenon suppression controller which forecasts the xenon reactivity at the time selected for a return to full power as a function reactor history, current reactor flux and an arbitrary load schedule can be used to determine the maximum reduced power level that will permit operation at full power at the selected time.

Abstract: A method of controlling the output power of a nuclear reactor core in which an output power distribution in the core is controlled with the use of an output power controlling means for controlling control rods, a flow of coolant or a concentration of neutron absorbing substance. Core data is calculated to include an output power distribution and a burn-up degree distribution in a fuel assembly, in accordance with detected data of the controlling means. Fuel rod data, which includes a temperature distribution, a stress distribution, a strain distribution and a concentration of fission products in a fuel rod, is calculated in accordance with the core data. Then, a failure event probability in the fuel rod is calculated in accordance with fuel rod data, the failure event probability being correlated with the history of the fuel rod concerning stress corrosion, strain and fatigue of the fuel rod. The soundness of the fuel rod is determined by comparing the failure event probability with a set value.

Abstract: A universal logic card is disclosed which contains both pre-programmed functional circuitry for actuating channel logic and dedicated test circuitry for system controller test and for constant and looping interrogation of the pre-programmed channel logic in absence of self-test system controller test. The disclosed card resides as a channel in a self-test system for a nuclear power plant supplementing the nuclear reaction protection system. The card is pre-programmed in digital logic to act equivalent to an analog circuit during actual plant operation. Additionally, the card responds to system controlled surveillance of the nuclear protection system. In system controlled surveillance, a central microprocessor serially addresses system circuit cards such as the one disclosed and loads these cards at predetermined input points with test commands. The addressed cards are thereafter simultaneously activated by a system-wide command.

Abstract: When the control parameter (.DELTA.T) is within its deadband, a correction of the axial deviation of power (DA) is effected by movement of at least one group of control rods, when DA is outside a predetermined range or deadband straddling a reference value DA ref. The group which is moved is chosen as a function of the sign of .DELTA.T, of the position of .DELTA.T with respect to its deadband and of the position of the other groups. When .DELTA.T is outside its deadband, the differences between the amplitudes of insertion of the groups participating in the control of the reactor are made to vary by stopping or starting of groups selected as a function of the positions of the groups of rods (G1, G2, N1, N2) in the core of the reactor and of the position of DA either with respect to its deadband if DA is outside this deadband, or with respect to DA ref, if DA is within this deadband.The invention is in particular applicable to pressurized-water nuclear reactors.

Abstract: Malfunction under water-solid conditions responsive either to increase in mass flow or increase in heat flow into the reactor coolant is suppressed. The power-actuable reactor-coolant relief valve is opened for increase in reactor-coolant mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). The interval is set to preclude opening of the relief valve for transients.

Abstract: A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Abstract: A method of operating a pressurized water nuclear reactor comprising determining the present core power and reactivity levels and predicting the change in such levels due to displacer rod movements. Groups or single clusters of displacer rods can be inserted or withdrawn based on the predicted core power and reactivity levels to change the core power level and power distribution thereby providing load follow capability, without changing control rod positions or coolant boron concentrations.

Abstract: Excessive swing of the feedwater in power supply apparatus on the occurrence of a transient is suppressed by injecting an anticipatory compensating signal into the control for the feedwater. Typical overshoot occurs on removal of a large part of the load, the steam flow is reduced so that the conventional control system reduces the flow of feedwater. At the same time there is a reduction of feedwater level in the steam generator because of the collapse of the bubbles under increased steam pressure. By the time the control responds to the drop in level, the apparatus has begun to stabilize so that there is overshoot. The anticipatory signal is derived from the boiling power which is a function of the nuclear power developed, the enthalpy of saturated water and the enthalpy of the feedwater injected into the steam generator. From the boiling power and the increment in steam pressure resulting from the transient are anticipatory increment of feedwater flow is derived.

Abstract: An apparatus which monitors a subset of control panel inputs in a nuclear reactor power plant, the subset being those indicators of plant status which are of a critical nature during an unusual event. A display (10) is provided for displaying primary information (14) as to whether the core is covered and likely to remain covered, including information as to the status of subsystems needed to cool the core and maintain core integrity. Secondary display information (18, 20) is provided which can be viewed selectively for more detailed information when an abnormal condition occurs. The primary display information has messages (24) for prompting an operator as to which one of a number of pushbuttons (16) to press to bring up the appropriate secondary display (18, 20). The apparatus utilizes a thermal-hydraulic analysis to more accurately determine key parameters (such as water level) from other measured parameters, such as power, pressure, and flow rate.

Abstract: System failure in the acquisition of data from a process plant such as a nuclear reactor owing to stuck signals is avoided by arranging input signals from transducers in the plant in a test pattern. The signal inputs are physically wired in the test pattern which is then decoded by either or both of a computer based system which allows for the test pattern and/or a hard wired system which mimics the input connections. Transmission of the multiplexed signals is made dynamic by provision of a polarity reverser switch (10) which alternates between signals or scans. A pattern recognition logic or a mimicing physical wiring decode the signals from the inputs or the computer system and the hard wired system respectively.

Abstract: A method of operating a nuclear reactor adapted to use cross-shaped control rods. The control rods are grouped into a plurality of groups having a first group consisting of a control rod located at the center of the core, a second group consisting of 8 control rods surrounding the control rod of the first group, a third group consisting of control rods located adjacent to and outside of the control rods of the second group, and other groups determined successively in the same manner as in the third group. During the operation period other than the period in which the number of the control rods to be inserted for realizing the critical condition of said nuclear reactor is less than 6, the nuclear reactor is operated with control rod patterns in which the inserted control rods are selected from control rods of alternate or every other groups of said groups so as to include at least a pair of control rods which are located in a similar manner to that of the KNIGHT's movement on a chess board.

Abstract: Disclosed is a control system and process for operating same for nuclear reactors, particularly gas cooled nuclear reactors, comprising in the control system a means for measuring the output produced in the reactor core and the output removed by a primary circuit or a secondary circuit, a means for integrating the integral of the core produced output over the output level, during the same time, of the primary or secondary circuit of the reactor and a means for forming a control signal from the value of the integral.

Abstract: A control apparatus for a boiling water reactor includes a monitoring section for monitoring power distribution in the reactor, a core flow control section for controlling the core flow and an arithmetic operational section for calculating change rate of the linear heat generation rate from the output of the monitoring section to thereby cause the core flow to be increased or decreased through the core flow control section when the calculated change rate exceeds a predetermined limit, whereby the increasing rate of the linear heat generation rate of fuel rods is maintained at a value not greater than the limit value. The power distribution monitoring section serves to generalize the current power distribution in the whole core in trms of typical monitoring quantities at selected locations of in-core neutron detectors.

Abstract: A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point (79) is disclosed. The set point (79) is dependent upon percent of full power load demand. A manually-actuated (85) "droop mode" of control is provided whereby the reactor coolant temperature is allowed to drop below the set point (79) temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow.