Abstract: A system for monitoring a limited number of operating conditions in a nuclear reactor and for graphically displaying same on a display device is disclosed. The system parameters selected are those which typically change as a result of a reactor trip. As long as these system parameters remain within a "post-trip window" on the graphic display, the power plant is operating normally. If, however, any of the foregoing system parameters falls outside this "window", this is an immediate indication to the power plant operator to institute the necessary procedures to remedy same. A sequential controller (26) is provided to continuously sequentially sample the foregoing operating parameters and to transmit data regarding same to the display device.

Abstract: A redundant reactivity control system, operating in conjunction with other systems in a nuclear power plant, to mitigate the potential consequences of an anticipated transient without scram event, is described. The invention interfaces to a boiling water nuclear reactor with the reactor's neutron monitoring system, standby liquid control system, control rod drive system, reactor recirculation system, reactor water clean-up feed water control system, and control room and local panels. Two divisionally separated control panels include associated detection and actuation logic and the necessary interface logic to other systems to perform the specific functions in response to an anticipated transient without scram event. Each separate division includes two identical channels. Actuation of any associated reactor system by the invention requires the agreement of the two channels within a division.

Abstract: A control system for overpressure protection of the reactor coolant boundary (14) in a nuclear power plant, comprising a pressure relief line (18a) having a power operated relief valve (24a) controlled by a circuit responsive to the condition of the reactor coolant and a control logic. The circuit includes a mode selector switch (80) having a NORMAL position (82) for establishing a full power overpressure relief valve setpoint (62) and a LTOP position (84) for establishing a low temperature overpressure relief setpoint (58). The control system logic opens the relief valve (24a) when (a) the mode switch is in the NORMAL position and the coolant pressure exceeds the full power setpoint (62), or (b) when the mode switch is in the LTOP position, the coolant temperature is less than a preselected value (52) and the coolant pressure exceeds the LTOP setpoint (58).

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: The invention relates to a method of protection of a pressurized-water nuclear reactor, which enables the crisis of boiling and the melting of fuel to be avoided. An emergency shutdown is triggered when a parameter representative of the crisis of boiling or the power released per unit length of the fuel elements exceeds a limiting value. For the determination of these parameters, the most exact representation possible of the distribution of the power in the core of the reactor is employed. This distribution is obtained by measurement of the axial power by ionization chambers, and determination of the radial power by measurement of the positions of the control rods.

Abstract: The operation of a nuclear fueled, electric power generating unit is monitored by utilizing decision tree analysis of selected system parameters to generate a representation of the real time system status. Current system status is presented to the operator either in the form of conclusory indications, such as verbal statements, representative of system status, or in the form of a visual display of the entire decision tree including indications of the path through the tree representative of current conditions. In either case, indications of off-normal conditions are accompanied by directions as to appropriate action to be taken, either by the operator or the automatic control system, to steer the system toward more acceptable conditions. In addition, the status indications are prioritized to indicate the seriousness of the off-normal conditions and to direct the sequence of corrective action to be taken when multiple off-normal conditions exist.

Abstract: A system for regulating the liquid level (20) in a vapor generator (10), in which the incoming flow of feed liquid (54) is regulated (48) in response to the difference between the measured liquid level (76) and a reference level (80), the difference between the exiting vapor mass flow rate (64) and the incoming liquid mass flow rate (56), and a function of the measured incoming liquid temperature (92). The temperature function produces a gain value (98), which increases in response to decreasing incoming liquid temperature.

Abstract: In a nuclear reactor power control system, when a control rod or a safety rod drops into a nuclear reactor by a failure of a control rod drive mechanism, a power of a summing and averaging circuit which sums and averages measurements of neutron detectors is smaller than a value A preset by an operator of the nuclear reactor. The averaged value of the neutron detector outputs is compared with the preset value A, and if the difference therebetween exceeds a preset limit signal level B for the output drop, automatic withdrawal of the control rod is stopped, and if the difference is smaller than the limit signal level B for the power drop, the control rod is withdrawn by the amount corresponding to the power drop.

Abstract: A method for regulating the power supplied to a steam turbine by a pressurized water nuclear reactor driving an electric current generator supplying a network by utilizing the movement of control rods in the reactor and the by-passing of a fraction of the steam produced by the reactor out from the turbine circuit. In the case of electrical uncoupling between the generator and the network or load rejection, the parameter determining the position of the control rods is selected as the greater of two values of a reference temperature. The by-passing of the steam is kept in operation and the operation of the reactor remains automatic. The opening signal for the by-pass valves is modified by a gain signal determined from a signal representing the power level. The invention is particularly applicable to the regulation of the power of pressurized water nuclear power stations.

Abstract: A power control apparatus of a boiling water atomic reactor for transferring the atomic reactor to a low power stand-by operation when a load of a generator is rejected. The necessity of insertion of control rods into a core is decided in accordance with the state of the atomic reactor when a load rejection occurs. When the insertion of control rods is required selected control rods are inserted into the core and a core flow rate is reduced to transfer the atomic reactor to the low power stand-by operation. When the insertion of control rods is not necessary only the reduction of core flow rate is performed.

Abstract: Method and apparatus for monitoring the movement of the control rods of a nuclear reactor.The number of steps of movement in either direction of the rod from which the control rod is suspended is counted. According to the height of the step, an indication of the position of the suspension rod and of the control rod. The apparatus comprises devices for measuring the speed of movement of the control rod, for logging variations in speed higher than a given value, and for counting such variations according to their sign. The invention is particularly useful in pressurized water nuclear reactors.

Abstract: A monitoring system 10 for providing a display 38 of the margin between actual and saturation pressure as well as a display 68 between actual and saturation temperature for the cooling fluid of a nuclear reactor. The system 10 also has an alarm 46 which is set off whenever the pressure margin to saturation pressure reaches a predetermined limit as well as a temperature margin alarm 78 which sets off an alarm whenever the temperature margin to saturation temperature reaches a predetermined limit.

Abstract: A system for controlling the feedwater system in a nuclear power plant is disclosed. The system utilizes a logic calculations subsystem (50) which selects the operational mode of the system to be monitored from a plurality of operational modes. A disturbance estimate calculations subsystem (52) contains a dynamic mathematic model of the feedwater system and compares measured system variables with corresponding dynamic model variables contained therein to produce signals representative of deviations therebetween. A modified feedforward calculations subsystem (54) utilizes these signals and other system parameters as inputs to a steady-state mathematical model to produce control signals to determine valve positions and pump speeds so that the control valve flows, minimum control valve pressure drop, and feedwater pump flows are at their respective demanded values.

Abstract: A load control system for a boiling water reactor power plant. A recirculating pump speed demand signal and a total steam flow demand signal are derived on the basis of combination of a first control signalproduced in accordance with a power difference signal representative of difference between an actual value of an output power of an electric generator and a set value for the generator output and a second control signal produced in accordance with a pressure difference signal representative of difference between an actually measured value of the reactor pressure and a set value therefor. Speed of a recirculating pump is controlled in dependence on the recirculating pump speed demand signal, while opening degrees of a governor valve and a bypass valve of a turbine are controlled in accordance with the total steam flow demand signal.

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: When monitoring the period of a nuclear reactor, a signal representing the reactor power is generated and compared with a reference signal. When a difference between the power signal and the reference signal is detected, the level of the reference signal is changed at a predetermined rate independent of the magnitude of said difference in a direction to reduce said difference. Hereby the reference signal is brought to follow the power signal as long as the latter does not change faster than said predetermined rate of change for the reference signal. At least one warning level signal is generated, the level of which being equal to the instant level of the reference signal multiplied by a predetermined factor greater than one. The power signal is compared also with this warning level signal, and a warning signal is generated when the level of the power signal becomes equal to or exceeds the level of said warning level signal.

Abstract: The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The protection system has a number of separate protection units, each unit receiving the process signals from the like sensors of each assembly in its turn. The sets of process signals derived from the sensor parameter assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector is interposed between the protection system and the control system.

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: This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus uses a separate bypass loop for conveying part of the reactor coolant away from the core, and at least three separate delayed-neutron detectors mounted proximate this detector loop. The detectors are spaced apart so that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the dealy time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector.

Abstract: A method for determining the transit time and flowrate of coolant traversing a cooling channel of a nuclear reactor fuel element from the effective center of the fuel element to a point where the cooland exits therefrom.

Abstract: A protection and control system for a nuclear reactor which monitors a variable operating parameter indicative of the state of reactor operation and identifies an abnormal rate of change in reactor operating conditions. The operating parameter monitored at a given time after the abnormal change in operating conditions is identified and is employed as a base for a variable setpoint defining a design limit within which the reactor can continue to operate. At a preselected time following the identification of the abnormal rate of change of the reactor operating conditions, control of the reactor is automatically modified to establish a reactor working environment in which the reactor can continue to operate. Should the established setpoint be exceeded, the reactor will automatically be tripped.In one embodiment the negative rate of change of neutron flux of the nuclear core is monitored and compared to a preselected reference.

Abstract: In a nuclear reactor having at least two control rod groups residing in the core for controlling the power level generated by the core, wherein one group is at a higher elevation in the core than the other, an improved method of controlling the core axial power distribution.

Abstract: In a nuclear reactor comprising control rods, nuclear fuel elements each having a cladding and pellets inserted therein, and a recirculation system, the reactor raises its power at least once by withdrawal of the control rods above a pellet-cladding interaction starting power (kw/ft) of the fuel element in a burnup range below the burnup at which the maximum pellet-cladding interaction starting power appears.

Abstract: A temperature control system for liquid plant is comprised of liquid metal, plant members, heaters for heating these liquid metal and the plant members at respective control points, a temperature control unit and a control unit. The plant members include a vessel containing liquid metal, pipe lines through which the liquid metal circulates, and valves. Further, temperature sensors, a level detector for detecting the level of the liquid metal, and a meter for detecting the flow rate of the liquid metal, are mounted to the plant members. Temperature sensors produce output signals relating to the condition parameter of the plant which data is delivered via an interface to the temperature control system and the control unit, as the information at each control point. The control unit previously stores the operating schedule of the plant members and transfers, the input signal relating to the plant condition to the temperature control system, with reference to the temperature control schedule.

Abstract: A nuclear reactor power monitoring system for monitoring the power level of a reactor and preventing an excessive rise thereof attributable to a transient increase in the core coolant flow rate before the reactor is scrammed. The system include an operating region monitor (ORM) for blocking the increase in the core coolant flow rate or running-back the flow rate when the power level exceeds a predetermined coolant block threshold power level which is a function of the core coolant flow rate.

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.

Abstract: In a control apparatus for residual heat removal system for a nuclear reactor in which the steam generated in the reactor vessel is cooled and condensed into water through a heat exchanger and the condensed water is fed back by a feedwater pump driven by the reactor steam so as to cool the reactor core, the pressure of the condensed water at the exit of the heat exchanger is increased with the rise of the reactor water level so as to decrease the flow of the condensed water, that is, coolant water and the flow-in of the make-up water from the make-up water reservoir communicating with the outlet port of the condenser while the quantity of the reactor steam supplied to the turbine is decreased with the increase in the reactor temperature fall rate so as to decrease the flow of the coolant water fed back into the reactor vessel.

Abstract: In supervising the channel stability of a nuclear reactor, the thermal power and the coolant flow quantity in each fuel assembly are determined by signals produced by a plurality of neutron flux detectors installed in the nuclear reactor, and signals regarding other operating conditions of the reactor, and the thermo-hydrodynamic stability of each fuel assembly is judged by the above-mentioned thermal power and coolant flow quantity as well as the measured values of such parameters as inlet subcooling of the core and of the pressure in the reactor vessel. Or the stability limit of a selected one of the parameters is determined. Then, the threshold of the thermo-hydrodynamic stability or the stability limit is compared with the actually measured value thereof so as to determine a stability margin.

Abstract: A control system and method for a nuclear steam supply system for continuously predicting the pending violation of one of the system's design limits. The prediction is made far enough into the future to allow corrective or protective action to be implemented in a timely manner in order to avoid violation of a design limit even on the occurrence of a worst case accident, or, restated, the occurrence of an accident which causes a most rapid approach to the design limit. Various parameters of the nuclear steam supply system are monitored and fed to a calculating mechanism. Samples of these parameters are periodically taken for the periodic detailed calculation of an index representative of the proximity of violation of the appropriate design limit. Continuous readings of these same parameters are taken for the purpose of making continuous updates of the last calculated index.