Abstract: An apparatus and process for the accurate measurement and calibration of core flow in a boiling water reactor is disclosed. The reactor includes a pressure vessel having internal forced circulation through the core. An annular downcomer region establishes reactor coolant flow downwardly in the periphery of the reactor vessel and finally upwardly and centrally into the core across a core plate. A pressure sensor detecting the pressure differential across the core plate is utilized. This pressure sensor has its measurement enhanced by input from local power range monitors in the core to utilize both the sensed pressure differential and the power to predict more accurately flow in the reactor. An algorithm is developed for utilizing the pressure differential and the real time readings from the local power range monitor to accurately gauge overall reactor coolant flow. To ensure accurate calibration, two calibration standards are itilized at steady states of reactor power output and coolant flow.

Abstract: A control system for a nuclear steam power plant having a reactor, one or more once through steam generators and a turbine includes a reactor control subsystem responsive to a power demand input to cause the reactor to deliver thermal power commensurate with the power demand input to the once through steam generator, one or more feedwater control subsystems sensing the thermal power delivered by the reactor to control feedwater flow to the once through steam generator, such as via a regulating valve and/or a feed pump, to match the thermal power delivered by the once through steam generator to the turbine with the thermal power delivered by the reactor, and a turbine steam demand control subsystem responsive to the thermal power delivered by the once through steam generator by sensing steam pressure to control a governor valve to admit steam to the turbine to cause the turbine operation to match the steam supplied by the once through steam generator such that the nuclear steam power plant is plant process con

Abstract: A safety qualified interface device is interposed between an operator manned control position having an operator interface and qualified safety system logic monitoring and control devices. This microprocessor driven interface incorporates a memory carrying a library collection of valid safety directives as well as valid parameter inputs and permissible parameter deviation ranges. Control or parameter requests are asserted by the operator from non-safety related but conveniently positioned interface devices at the control position which are treated as trial inputs which are matched at the interface with memory retained valid directives and parameters. Where a match between memory retained data and the trial input signals occurs, then the operator is apprised through a qualified safety related display and a corresponding acknowledgement is made to the interface.

Abstract: In the event of an incident in the secondary cooling circuit of a nuclear reactor preventing cooling the water in the primary circuit of the steam generators (2) which are common to both circuits, an emergency stop instruction is issued in order to cause control and stop rods to fall into the core of the reactor. If this instruction is not executed, and if the reactor remains at a high power level, an instruction is issued to stop the primary pumps. A rapid increase in the temperature of the water in the core then slows down the nuclear reaction soon enough to prevent subsequent excess pressure from damaging the primary circuit. The invention is particularly applicable to electricity-generating nuclear power stations.

Abstract: The steam generator water level, the feed water temperature, and the reactor power are continuously measured and a level signal (76), a temperature signal (78), and a power signal (80) commensurate with these measurements are respectively generated. From the power signal and the temperature signal, a target flow signal (84) is generated corresponding to the feedwater flow rate that would produce a stable steam generator water level at the steady state condition of the power signal and the temperature signal. The level signal is compared to a level signal setpoint (54) and a resulting level error signal (86) is generated. The level error signal is adjusted by a gain factor (88) that is dependent on the temperature signal to produce an adjusted level error signal (90). The feedwater flow through the bypass valve is then controlled in response to the target flow signal (84') and the adjusted level error signal.

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: Jet pumps and a feed water sparger are disposed in a reactor pressure vessel incorporating a core. A first feed water pipe supplies a portion of feed water to the jet pumps as their driving water. A second feed water pipe connected to the first feed water pipe supplies the remaining portion of the feed water to the feed water sparger. A first feed water heater is disposed in the first feed water pipe downstream of a connecting point of the first feed water pipe and the second feed water pipe. An amount of extracted steam introduced into the first feed water heater is controlled on the basis of a difference in the temperature between feed water flowing through the first feed water pipe and the feed water flowing through the second feed water pipe.

Abstract: A natural circulation reactor includes a reactor pressure vessel, a cylindrical shroud disposed within the reactor pressure vessel to surround a reactor core and to extend to a position above the reactor core, and a coolant descending passage defined between the reactor pressure vessel and the shroud for allowing the coolant discharged from the top of the reactor core to be again supplied to the bottom of the reactor core. In such a natural circulation reactor, a liquid level of coolant in the coolant descending passage is adjusted on the basis of a reactor power request signal at a position below a liquid level of coolant in the shroud, and thus reactor power is controlled so that it may reach a predetermined reactor power by adjusting the liquid level of coolant in the coolant descending passage.

Abstract: A smart reactivity computer is disclosed that processes signals from multiple neutron detectors to generate values of regional core reactivity that are output to a strip chart recorder. The computer compares the regional reactivity values and, when the values are coincident for a period sufficient to satisfy a statistical fit test to a straight line, indicates to the operator that true core reactivity can be determined. The computer determines control rod worth by first fitting a straight line to the coincident portion of the reactivity values. The computer then determines intersections between the straight line and the reactivity value lines produced when the control rods are moved on both ends of the straight line. The difference between intersections for a single control rod movement line determines control rod worth for that movement.

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: The invention improves detecting that a control cluster has fallen into the reactor of a nuclear fission power station. Variations are monitored firstly in the nuclear power (3) and secondly in at least one parameter (1, 2) external to the reactor and having an influence on reactor control. A cluster is considered as having fallen when a rapid drop (12j) is detected in said power while no variation (12a, 12b, . . . 12h) is simultaneously detected in an external parameter that could have given rise to said power drop.Application to producing electricity.

Abstract: In moisture separator reheaters of a nuclear power plant, a control valve is mounted in a steam vent line of the plant to control the flow rate of vent steam in conformity with the load applied to the plant, to thereby avoid the occurrence of any instable flow phenomenon in the entire range of loads applied to the plant.

Abstract: The present invention provides a power distribution synthesis system that uses a small number of local axial power distributions determined from less than a full complement of fixed incore neutron detectors and a power sharing map produced from a full complement of core exit thermocouple temperature signals to synthesize the full core power distribution. The localized tilt of the core is determined from the power sharing map and used to remove the tilt from the local distribution. The tilt adjusted local distribution is assigned to symmetric core locations where uninstrumented core power values are determined by interpolation. The interpolated, assigned and tilt corrected local distribution power values are transformed into the three dimensional power distribution using the tilt.

Abstract: The low alarm threshold (B) of the overheating ratio of a nuclear reactor is determined in such a manner that the higher the position of the temperature-regulating rods, the lower the value determined for said threshold. The invention is applicable, in particular, to pressurized water nuclear reactors used for producing electrical power.

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: A computer based system aids an operator in proceeding step-by-step through procedures for a complex process facility. At each step, monitored plant parameter values are used to evaluate relevant plant status and recommend action to be taken. The status and recommended action are presented to the operator on a display device together with prompts for generating appropriate responses. The step logic is carried out repetitively to provide the operator with feedback and to verify operator actions. The complete display picture including operator responses, and other plant conditions monitored in parallel with the current step, is logged for later review. An online review feature permits review of plant conditions and operator actions while the operator continues to execute the procedure. High-level textual statements of all steps of a current procedure can be reviewed and prior steps can be executed or re-executed.

Abstract: Signals from redundant sensors located throughout a pressurized water reactor (PWR) nuclear power plant are processed in four independent channel sets each of which includes a plurality of independent microcomputers which calibrate, convert to engineering units and calculate partial trip signals and engineered safeguard actuation signals from the sensor signals for use in the conventional voting logic of a plant protection system. The primary and secondary partial trip and engineered safeguard actuation functions associated with various postulated abnormal events are allocated to different independent microcomputers in the channel set for reliability. A test unit common to the channel set automatically, rapidly bypasses and tests each protection function independently while the other protection functions in the channel set remain on-line and also continually tests each microcomputer through a dummy test function performed along with the assigned protection functions.

Abstract: The operator of a complex process facility is aided in executing step by step procedures by a computer based system which incorporates monitored plant data and interactive plant operator responses to prompts to progress step by step through selected procedures. At each procedure step the plant status is evaluated and a recommended course of action is identified. The results are displayed on a visual display device to provide operator guidance for executing the procedure in effect. Parallel process monitoring visually alerts the user to conditions not considered by the current step of the active procedure, and where the seriousness of the other condition warrants, provides procedure-based recommendations for priority action. The system can be accessed by multiple users at separate locations for coordinated execution of a single procedure or simultaneous execution of two different procedures.

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: Closed-loop control law for altering the power level of nuclear reactors in a safe manner and without overshoot and in minimum time. Apparatus is provided for moving a fast-acting control element such as a control rod or a control drum for altering the nuclear reactor power level. A computer computes at short time intervals either the function:.rho.=(.beta.-.rho.).omega.-.lambda..sub.e '.rho.-.SIGMA..beta..sub.i (.lambda..sub.i -.lambda..sub.e ')+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e '.omega.]or the function:.rho.=(.beta.-.rho.).omega.-.lambda..sub.e .rho.-(.lambda..sub.e /.lambda..sub.e)(.beta.-.rho.)+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e .omega.-(.lambda..sub.e /.lambda..sub.e).omega.]These functions each specify the rate of change of reactivity that is necessary to achieve a specified rate of change of reactor power.

Abstract: A rod drive control system for nuclear such as boiling water reactors is comprised of a plurality of stepping motors for inserting and withdrawing control rods disposed within the nuclear reactor, which the stepping motors are each provided for each control rod, terminal units which are each provided for each stepping motor and are each provided with a motor controller to drive the related stepping motor and to check the operation of the motor, a central processing unit for transferring an operation signal to the motor control unit and for detecting faults in the terminal units, and transmission lines between the terminal units and the central processing unit.

Abstract: A water level control for a steam generator of a pressurized water type of nuclear steam supply system varies the water level demand as a function of power. The control also varies the water feed rate after a reactor trip initially as a function of reactor coolant average temperature.

Abstract: A method and apparatus for generating core-average and radially peaked, pointwise core axial power distributions of a nuclear reactor core from monitored responses of reactor flux obtained exterior of the core, such as responses obtained from ex-core power range nuclear detectors, and from responses of temperature obtained from selected core-exit thermocouples.

Abstract: Disclosed are methods and apparatus for the control of dynamic systems, more particularly complex dynamic systems and especially those systems exhibiting minimum phase behavior. The method utilizes observer theory to estimate the state variables of the system. Based in part upon the estimated values of the state variables, a pseudo or compensated output for the system is generated. This compensated output represents the steady state asymptote of the actual system output without the unstable or transient system responses. The actual outputs are then controlled by controlling the compensated output.According to another embodiment, a bounded output may also be generated from the state variable estimates. This bounded output represents the minimum and/or maximum excursion that the actual output takes in its approach to the steady state value represented by the compensated output.

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: 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 control apparatus for electric power plant comprises turbine speed control means for detecting the speed of a turbine coupled to an electric generator to control the turbine speed by controlling a steam flow supplied to the turbine on the basis of the detected speed, pressure control means for controlling the pressure in a steam generator on the basis of the pressure of the steam generator, and correcting means for correcting the steam flow supplied to the turbine only when fluctuation in the pressure of the steam generator brought about by rapid change of a set load value is deviated from a predetermined range in which the steam pressure is allowed to vary, the correction of the steam flow being made in correspondence to the deviation.

Abstract: Components and systems which impact technical specifications (tech specs) in a nuclear power plant are modeled in a set of cascading operability trees which represent the various combinations of component, system and parameter statuses which can result in non-compliance with the tech specs. Cutsets for the operability trees are entered into a cutset table in the data base of a digital computer which tracks the effect of a selected event up through the trees by repetitively searching the cutset table using the top event activated by the selected event as the selected event for the next search, until an activated top event is reached which is not entered in the cutset table as an event. A separate cutset evaluation table is used to assure that all the members of a multiple event cutset have been activated before the associated top event is activated.

Abstract: Excessive primary cooldown and start up of the auxiliary feedwater system in a pressurized water reactor (PWR) following a reactor trip are avoided by switching to primary coolant T.sub.avg programmed, main feedwater flow control in place of normal level programmed control. The Low T.sub.avg set point at which the main feedwater isolation valves are closed is lowered below the no load T.sub.avg temperature set point and the flow rate is decreased gradually to zero at a T.sub.avg between these two set points. In addition, the low-low level set point at which the auxiliary feedwater system is started is lowered if possible, and/or the level signal is filtered, or the start signal delayed, to increase the margin to the low-low set point.

Abstract: A nuclear reactor start-up method of starting up a boiling water reactor which is in the cold shut-down state. The method includessetting a subcooled temperature of a coolant in a reactor core inlet to a range smaller than a critical subcooled temperature which is determined by the condition whereby the coolant starts boiling in the reactor core and the condition whereby the instability due to a periodic variation of buoyancy begins,thereafter starting an increase of power of the reactor.

Abstract: A full range steam generator level control system and method is provided by a weighted linear combination of a pair of inputs from a low power and a high power controller to actuate feedwater regulation valves.

Abstract: A control system for a nuclear reactor is disclosed. A control rod strategy computer provides for dynamic control of core power distribution in both radial and axial directions and forms the basis for a partial trip capability. Several microprocessor-based computation centers are combined together in a data-sharing network which processors determine local power density, determine the instantaneous differential and integral reactivity worth of each group of control rods, determine and effectuate partial trip for immediate power reduction, determine and provide for uniform core burnup, and effectuate core reactivity changes by directing the movement of groups of four control rods from zero to one hundred percent of travel while minimizing power distribution factors throughout the core. The power control circuitry to move the groups of control rods are bus arranged such that the power circuitry is shared among the groups of control rods.

Abstract: A protection system such for a nuclear power plant shuts down the reactor in response to the detection of a negative flux rate characteristic of a dropped rod, but inhibits the trip if the specified negative flux rate is preceded within a selected time interval by a positive flux rate characteristic of a net load rejection when the moderation temperature coefficient is highly negative.

Abstract: A calibrator for a pressurized water reactor (PWR) core predictor utilizing a distribution expression for the transverse buckling factor, in a one dimensional diffusion model, which has coefficients correlated with overall neutron balance and with a set of power distribution characterization parameters including axial offset, axial pinch, axial quarters and axial fifths. The coefficients are periodically adjusted incrementally one at a time to produce criticality with the model at measured values of rod position, reactor power level, cold leg temperature and boron concentration until the power distribution characterization parameters calculated from a power distribution generated by the model are within preset tolerances of those from a power distribution measurement generated by movable, or if provided, fixed, incore detectors.

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: In a pressurized water reactor in which a dropped control rod is detected by a negative rate of change of neutron flux which exceeds a preset negative limit, the reactor is only shutdown if the axial offset exceeds a preselected limit, so that it is possible to maintain power, and in many cases full power, despite the presence of a dropped rod. Where part length control rods are used during power changes, separate limiting values on axial offset are used when these rods are fully inserted and when they are fully retracted.

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: A device and method for determining the power of a pressurized water nuclear reactor, comprising measurement of the neutron power (15) and of the temperature of the primary fluid in the cold branch (1) and in the hot branch (1'), computing (2, 4, 6, 8, 12) the thermal power of the reactor from these temperature measurements (1, 1'), comparing (14) the thermal power signal (13) and the neutron power signal (16) and correcting (22) the neutron power signal (16). The device in addition comprises means for fast and precise determination of the secondary power of the reactor by comparison of a first signal of the secondary thermal power (13) established by means of the temperature measurements (1, 1') and a second signal of the thermal power (16') established by means of a simplified secondary heat balance (25).

Abstract: An analog output network provides multiple communication paths for control of an actuator in response to an input signal. The analog output network includes at least two controllers, each of which is connected to an analog output circuit and a controller selector which is connected to each of the controllers and the analog output circuits. The controller selector ensures that at least one of the analog output circuits is enabled at all times and identifies one of the controllers as an on-line controller. The analog output circuit connected to the on-line controller generates an actuation signal which passes through the actuator and the other analog output circuit. An alternate current path comprising a diode string with a voltage drop larger than the voltage drop of the actuation signal in the other analog output circuit is provided so that if an analog output circuit is removed or defective, the actuation signal from the on-line analog output circuit is still provided with a current path.

Abstract: A control apparatus and method for restricting liquid flow in a liquid moving pump, usually of the centrifugal type, to prevent pump cavitation and pump prime mover overloading. The control apparatus includes sensors to detect liquid temperature and pressure at the inlet of the pump. It may further include a device such as a current transformer to develop a signal indicative of power consumed by the prime mover of the pump where the prime mover is an electrical motor. The liquid pressure and temperature indications are used to generate a specific indication of the subcooling of the liquid. The temperature indication is used to derive an indication of the instantaneous required subcooling of the pump. The subcooling indication and the required subcooling indication are introduced to a comparator. Should the subcooling of the liquid fail to exceed the required subcooling, a first control signal is generated.

Abstract: The invention relates to a process for monitoring leaks in the primary circuit of a pressurized water nuclear reactor.The primary circuit consists of the vessel (1), the primary part of the steam generators (6), pipework (3, 4), the pressurizer (10) and various stocktanks (20). Connected to the primary circuit is at least one auxiliary circuit such as the circuit for volumetric and chemical monitoring comprising a stocktank (17). A sectioning of the primary circuit and of the circuit for volumetric and chemical monitoring into volume elements, excluding the stocktanks (20, 17) is determined. At regular time intervals, or steps, the mass of water in each of the volume elements and in each of the stocktanks is determined from measurements of pressure, temperature and levels and then, by addition, of the total mass of water.

Abstract: An operation control method for a nuclear reactor performing a load follow-up operation in accordance with a load variation program, wherein a core reactivity which changes with time in a first cycle of operation is predicted based on the load variation program and data for analyzing dynamic characteristics of the reactor, and a change in a liquid poison concentration in the first cycle of operation is obtained based on the predicted reactivity, and reactor power is controlled in the first cycle of operation by adjusting the liquid poison concentration in accordance with the obtained change. When the liquid poison concentration is adjusted and control rods are manipulated, a reactivity introduced by these operations in the first cycle of operation is obtained, and an adjustment to be made to the liquid poison concentration in the second cycle of operation which is the next cycle of operation following the first cycle of operation is obtained from the reactivity introduced in the first cycle of operation.

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: An apparatus for monitoring and evaluating the degree of risk associated with continued operation of a nuclear power plant while one or more plant components has failed or is otherwise unavailable. The apparatus has several functional sections, including means for storing a plant-specific data base of component-level core damage logic paths and component level failure probabilities, means for selecting plant condition scenarios by modifying the component failure probabilities to represent plant components actually or potentially unavailable, means for associating a figure of merit with the change in risk of core damage resulting from the unavailable components, and means for displaying the figure of merit relative to a base or reference value. As an interactive tool in the plant, the apparatus PSES displays the probability or risk of core damage almost instantly for any given state of plant readiness.

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).