Abstract: A method for calculating a PCI margin associated with a loading pattern of a nuclear reactor including a core into which fuel assemblies are loaded according to the loading pattern is implemented by an electronic system. The fuel assemblies include fuel rods each including fuel pellets of nuclear fuel and a cladding surrounding the pellets. This method includes calculating a reference principal PCI margin for a reference loading pattern of the fuel assemblies in the core; calculating a reference secondary PCI margin for the reference pattern; calculating a modified secondary PCI margin for a modified loading pattern of the fuel assemblies in the core, and calculating a modified principal PCI margin for the modified pattern, depending on a comparison of the modified secondary PCI margin with the reference secondary PCI margin.

Abstract: The present disclosure relates to a novel gas turbine system having applications, for example, in thermal power generation in an environmentally friendly manner. The multiloop gas turbine system may have multiple functional units each comprising a compressor, a regenerator, a combustion unit, and a turbine. Typically, exhaust flow of a turbine of a preceding loop may be routed to the combustion unit of the next loop, allowing mixing of exhaust flow with hot compressed air of the next loop, and the expanded exhaust from the turbine of the ultimate loop is fed back into the regenerators of each loop to recover exhaust heat.

Abstract: A power generator system with anomaly detection and methods for detecting anomalies include a power generator that includes one or more physical components configured to provide electrical power. Sensors are configured to make measurements of a state of respective physical components, outputting respective time series of said measurements. A monitoring system includes a fitting module configured to determine a predictive model for each pair of a set of time series, an anomaly detection module configured to compare new values of each pair of time series to values predicted by the respective predictive model to determine if the respective predictive model is broken and to determine a number of broken predictive model, and an alert module configured to generate an anomaly alert if the number of broken predictive models exceeds a threshold.

Abstract: A method for the nitro-carburization of a deep-drawn article or a stamped-bent article made of austenitic rustproof stainless steel includes inserting the article into an oven in a first process step and heating the article to a first temperature, wherein an oxygen containing standard atmosphere that is present in the oven is replaced by a first gas mixture, and in which the article is heated up to a second temperature in a second process step, wherein the first gas mixture is replaced by a second gas mixture, and in which the article is maintained on the second temperature in a third process step, wherein the article is treated with the second gas mixture, and in which the article is cooled down to a third temperature in a fourth process step, wherein the second gas mixture is replaced by a third gas mixture.

Abstract: A reactor power regulator that adjusts an output of a reactor on a basis of an operation pattern or a reactor output target value and a reactor output change rate that are input by a central load dispatching center or an operator, including: a reactor output calculating device that performs computation based on a thermal equilibrium from various necessary plant state quantities to calculate a reactor output signal; a correcting device that corrects a continuously obtained reactor output equivalent signal that is considered to be equivalent to a reactor output at a calculation interval of the reactor output signal, for each calculation interval in the reactor output calculating device so that the reactor output equivalent signal coincides with the reactor output signal calculated by the reactor output calculating device, and calculates a continuous corrected reactor output equivalent signal; a reactor output controlling device that calculates at least one type of reactor output control signal for controlling t

Abstract: An apparatus for monitoring nuclear thermal hydraulic stability of a nuclear reactor, contains: a calculation unit configured to calculate a stability index of a nuclear thermal hydraulic phenomenon based on nuclear instrumentation signals, the signals being outputted by a plurality of nuclear instrumentation detectors placed at regular intervals in a reactor core; a simulation unit configured to simulate the nuclear thermal hydraulic phenomenon based on a physical model by using information on an operating state of the nuclear reactor as an input condition; a limit value updating unit configured to update a limit value of the nuclear thermal hydraulic phenomenon based on a result of the simulation; and a determination unit configured to determine, based on the stability index and the limit value, whether or not to activate a power oscillation suppressing device.

Abstract: A system for discharging the residual power of a nuclear reactor includes: a containment vessel incorporating a primary vessel including the core; a water reserve; a steam source wherein the heated primary water circulates and heats the secondary water circulating in the steam source; a condenser in the containment vessel including: a recovery unit; a condenser link linked to an intermediate water circuit and ensuring the circulation of the intermediate water between the water reserve and the condenser; a hot link ensuring the circulation of the steam to the condenser; a cold link ensuring the circulation, by gravity, of the water from the recovery unit to the secondary water inlet of the steam source; a heat recovery unit on the intermediate water circuit, the heat recovery unit being traversed by a feed water circuit, the feed water capable of heating up by thermal contact with the intermediate water circulating through the heat recovery unit.

Abstract: A method for producing a wear-resistant and corrosion-resistant stainless steel part for a nuclear reactor is provided. This method includes steps of providing a tubular blank in austenitic stainless steel whose carbon content is equal to or lower than 0.03% by weight; shaping the blank; finishing the blank to form the cladding; hardening the outer surface of the cladding by diffusing one or more atomic species; the blank, before the providing step or during the shaping or finishing step, being subjected to at least one solution annealing with sub-steps of: heating the blank to a sufficient temperature and for a sufficient time to solubilise any precipitates present; quenching the blank at a rate allowing the austenitic structure to be maintained in a metastable state at ambient temperature and free of precipitates.

Abstract: A system for controlling the power level of a natural circulation boiling water nuclear reactor (NCBWR) may include a heating subsystem for heating feedwater flowing into an annulus of the NCBWR to increase the temperature of recirculation water flowing through the core above a predetermined recirculation water operating temperature. Additionally the system may include a temperature sensor operable to sense the temperature of the feedwater flowing into the annulus. The temperature sensor is communicatively connected to a temperature controller operable to command the heating subsystem to increase the temperature of the feedwater flowing into the annulus to a requested temperature above a predetermined operating temperature of the feedwater flowing into the annulus.

Abstract: A pressurized water reactor (PWR) comprises a pressure vessel, a reactor core disposed in the pressure vessel, an integral or external pressurizer, primary coolant disposed in the pressure vessel and heated by operation of the reactor core, and a steam generator disposed in the pressure vessel and configured to convert secondary coolant in the form of feedwater into steam by heat transfer from the primary coolant heated by operation of the reactor core to secondary coolant in the steam generator. A controller is configured to perform a PWR control method including the operations of (i) adjusting one or more parameters of the PWR and (ii) adjusting a pressurizer water level setpoint based on a predicted direction and magnitude of change of a pressurizer water level of the PWR predicted to result from the adjusting (i).

Abstract: A method for monitoring and acoustic damping a boiling water reactor which includes: a reactor pressure vessel; a steam pipe for transporting steam out from a steam dome of the reactor pressure vessel; a high pressure turbine driven by the steam; a feedwater heater which heats feedwater supplied to the reactor pressure vessel using bleed steam from the high pressure turbine; a bleeding valve which adjusts a flow rate of the bleed steam; and a pressure sensor provided in a main steam line including the steam dome and the steam pipe. The method includes monitoring a fluctuating pressure in the main steam line and controlling an opening degree of the bleeding valve based on a magnitude of the fluctuating pressure.

Abstract: A tool for use in servicing an LPRM assembly of a nuclear reactor vessel includes a structural member having a first bore that is structured to receive an LPRM device associated with the LPRM assembly, a headpiece provided at a first end of the structural member, and a nut engaging assembly slideably mounted on the structural member. The headpiece has a plurality of projections structured to mate with a plurality of bores provided in a seal of the LPRM assembly to enable the seal to be removed, and the nut engaging assembly has a housing that defines a second bore and that has first and second nut engaging portions. The nut engaging assembly is free to slide along the structural member and over the headpiece to a position wherein the engaging portions extend beyond the headpiece so that they may be used to remove the assembly nut.

Abstract: In one embodiment, the method includes determining the safety limit minimum critical power ratio using the operating limit minimum critical power ratio, a change-in-critical-power-ratio distribution bias and a change-in-critical-power-ratio distribution standard deviation.

Abstract: Provide a nuclear reactor power monitoring technology for enhancing the monitoring accuracy and reliability in nuclear thermal hydraulic stability.

Abstract: A pressurized water reactor (PWR) comprises a pressure vessel, a reactor core disposed in the pressure vessel, an integral or external pressurizer, primary coolant disposed in the pressure vessel and heated by operation of the reactor core, and a steam generator disposed in the pressure vessel and configured to convert secondary coolant in the form of feedwater into steam by heat transfer from the primary coolant heated by operation of the reactor core to secondary coolant in the steam generator. A controller is configured to perform a PWR control method including the operations of (i) adjusting one or more parameters of the PWR and (ii) adjusting a pressurizer water level setpoint based on a predicted direction and magnitude of change of a pressurizer water level of the PWR predicted to result from the adjusting (i).

Abstract: Discussed is a method and device for increasing the operational flexibility of a current-generating system with a turboset, comprising a turbine coupled to an electrical generator, a power set point value being predefined, and a future target time at which the turboset is supposed to be at the power set point value, so that a power curve is determined via the power set point value and the target time, the turboset being operated starting from a actual power value at an actual time along the power curve such that the predefined power set point value is reached at the predefined target time. Also discussed is a gas turbine and to a steam turbine in accordance with the above description.

Abstract: A boiling water reactor includes a reactor pressure vessel; a steam pipe for transporting steam generated in the reactor pressure vessel out from a steam dome positioned at an upper part of the reactor pressure vessel; a high pressure turbine connected to the steam pipe and driven by the steam; a feedwater heater which heats feedwater supplied to the reactor pressure vessel using bleed steam flowing from the high pressure turbine to the feedwater heater; a bleeding valve which adjusts a flow rate of the bleed steam; and a pressure sensor provided in a main steam line including the steam dome and the steam pipe. The boiling water reactor further includes a monitor and control system which controls an opening degree of the bleeding valve based on a magnitude of fluctuating pressure in the main steam line that is detected by the pressure sensor.

Abstract: A nuclear power plant and method of operation for augmenting a second reactor thermal power output in a second operation cycle to a level larger than a first reactor thermal power output in the previous operation cycle. The plant is equipped, for example, with a reactor; a steam loop comprising high and low pressure turbines; a condenser for condensing steam discharged therefrom the low pressure turbine; a feedwater heater for heating feedwater supplied from the condenser; and a feedwater loop for leading feedwater discharged from the feedwater heater to the reactor. The operation method includes decreasing a ratio of extraction steam which is led to the feedwater heater from a steam loop in the second operation cycle.

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

Abstract: A power generation system is disclosed. The power generation system includes a nuclear reactor, a steam turbine, a gas turbine, and a primary generator. The steam turbine is in thermal connection with the nuclear reactor via a transfer medium and converts thermal energy to rotation. The gas turbine converts thermal energy to rotation and is in thermal connection with the transfer medium to increase a thermal energy of the transfer medium. The primary generator is in mechanical connection with the steam turbine to generate power in response to a rotation of a rotor of the primary generator.

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

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

Abstract: A method has been developed to select fuel rod enrichments for a fuel bundle of a nuclear reactor, the method including: creating an ordered list of fuel rod types in an initial fuel bundle design; perturbing at least a subset of the fuel rod types in the initial fuel bundle design to generate a plurality of perturbed fuel bundle designs; selecting perturbed fuel bundle designs having fuel rods with allowable fuel enrichment types and an allowable average enrichment for the perturbed bundle; determining a difference between each of the selected perturbed fuel bundle design and the initial fuel bundle design, and creating a group of the perturbed fuel bundle design having a difference less than a predetermined threshold difference value.

Abstract: A nuclear power plant and method of operation for augmenting a second reactor thermal power output in a second operation cycle to a level larger than a first reactor thermal power output in the previous operation cycle. The plant is equipped, for example, with a reactor; a steam loop comprising high and low pressure turbines; a condenser for condensing steam discharged therefrom the low pressure turbine; a feedwater heater for heating feedwater supplied from the condenser; and a feedwater loop for leading feedwater discharged from the feedwater heater to the reactor. The operation method includes decreasing a ratio of extraction steam which is led to the feedwater heater from a steam loop in the second operation cycle.

Abstract: A method and arrangement of determining pin enrichments for a fuel bundle of a nuclear reactor, where a plurality of input parameters and target conditions may be input and enrichment changes, to be made across the fuel bundle, may be calculated using response matrix technology. Fuel bundle pin enrichment data may be output that satisfies the target conditions. The method and arrangement may enable production of fuel bundles having a desired local peaking, exposure peaking and R-factor performance. Consequently, given fuel cycles typically may be loaded and operated such that less fuel may be needed for identical cycle lengths, potentially resulting in improved fuel cycle economics. Additionally, because fuel bundle development may require fewer iterations, there may be a substantial cycle time reduction in the bundle design process, potentially reducing cost and enhancing profitability.

Abstract: The invention relates to a control system for controlling an average temperature of a coolant at a reactor core outlet. The control system method detects an average temperature of the coolant at the reactor core outlet, compares the actual average temperature of the coolant at the reactor core outlet with a reference temperature thereby to generate an error signal, and adjusts the actual average temperature of the coolant at the reactor core outlet in response to the error signal. The invention extends to a cascade controller for a nuclear reactor, and to a nuclear power plant.

Abstract: A main steam system around a nuclear reactor which comprises two main steam nozzles in a reactor pressure vessel, each of the main nozzles being disposed at a symmetrical position with respect to a plane parallel to steam outlet faces of steam dryers provided inside the reactor pressure vessel and passing through a center of the reactor pressure vessel; and main steam pipes each connected to the main steam nozzles.

Abstract: An improved method of increasing the power output of an existing nuclear power plant includes increasing the thermal power output of the plant's nuclear island and constructing of an auxiliary BOP to handle the increased thermal power. The thermal power of the nuclear island can be increased such as by increasing the thermal power of the plant's reactor, by replacing the plant's steam generator with one that is more efficient, and by increasing the flow rate and/or change in temperature of a coolant in a secondary cooling loop of the plant. The thermal power of the reactor can be increased such as by replacing existing cylindrical fuel rods with fuel rods having a relatively greater surface area to volume ratio and/or by increasing the flow rate and/or the change in temperature of a coolant of a primary cooling loop. The auxiliary BOP can be constructed while the plant is in operation, and can then be connected with the nuclear island during a maintenance operation on the reactor.

Abstract: A reactor core (104) includes fuel assemblies (202) arranged in a loading pattern (1100) in response to corresponding power levels of the fuel assemblies (202). A crud deposition model (412) is used to predict crud deposition on the fuel assemblies (202) and fuel pins (300) within the fuel assemblies (202). The prediction of crud deposition enables generation of the loading pattern (1100) and design of lattice structures (1600) for the fuel assemblies (202) that results in the reduction of total crud deposition in the reactor core (104) and causes a substantially uniform crud deposition on the fuel assemblies (202) of the reactor core (104).

Abstract: In order to provide an operation and maintenance planning aiding system for a power generation installation which prepares an operation plan for a plurality of power generation units by making use of actual plant data and based on a total judgement including a variety of circumstances of the machine and apparatus or the parts thereof in the power generation units, in the system the plurality of power generation units 41, 42, 51 and 52, a power supply command center 3 and a service center 1 are arranged and connected via a communication network 6, the service center 1 obtains the plant data via the communication network 6 from the plurality of power generation units 41, 42, 51 and 52, calculates in real time a power generation efficiency of a concerned power generation unit for every plurality of power generation units 41, 42 51 and 52 by making use of the obtained plant data and design data of the concerned power generation unit and prepares an operation and maintenance plan for each of the power generation u

Abstract: In order to provide an operation and maintenance planning aiding system for a power generation installation which prepares an operation plan for a plurality of power generation units by making use of actual plant data and based on a total judgement including a variety of circumstances of the machine and apparatus or the parts thereof in the power generation units, in the system the plurality of power generation units 41, 42, 51 and 52, a power supply command center 3 and a service center 1 are arranged and connected via a communication network 6, the service center 1 obtains the plant data via the communication network 6 from the plurality of power generation units 41, 42, 51 and 52, calculates in real time a power generation efficiency of a concerned power generation unit for every plurality of power generation units 41, 42 51 and 52 by making use of the obtained plant data and design data of the concerned power generation unit and prepares an operation and maintenance plan for each of the power generation u

Abstract: A method of controlling a nuclear power plant includes determining a first maximum linear heat generation rate and a first minimum critical power ratio by a core monitoring system, and determining a second maximum linear heat generation rate and a second minimum critical power ratio by utilizing the first maximum linear heat generation rate and the first minimum critical power ratio determined by the core monitoring system and plant data. The second maximum linear heat generation rate and the second minimum critical power ratio are compared with predetermined thermal limit values and a control signal outputted from an automatic power regulator system to a re-circulation flow control system and to a control rod control system is held when at least one of the second maximum linear heat generation rate and the second minimum critical power ratio exceeds the predetermined thermal limit values.

Abstract: A control system, and a control method, of a nuclear power plant capable of easily executing automatic output regulation by an automatic power regulator system even when a lower-accuracy maximum linear heat generation rate and a lower-accuracy minimum critical power ratio are determined in a short cycle by using brief calculation. While a core monitoring system does not calculate a maximum linear heat generation rate and a minimum critical power ratio, they are determined by utilizing the maximum linear heat generation rate and the minimum critical power ratio calculated by the core monitoring system as well as plant data, and the lower-accuracy maximum linear heat generation rate and the lower-accuracy minimum critical power ratio so determined are compared with predetermined thermal limit values.

Abstract: During operation of the nuclear reactor, at specified time intervals, a neutron flux is measured using a set of neutron flux detectors which are fixed and arranged in the core (1) of the nuclear reactor, the maximum number n of detectors (8) being 15% of the number of fuel assemblies in the core. The measured signals are processed and the instantaneous distribution of neutron flux or of power throughout the entire core (1) is calculated from the measured signals. At least one core operating parameter is calculated from the instantaneous neutron flux distribution and an alarm is raised if at least one parameter exceeds a set range.

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

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

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

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

Abstract: An apparatus for measuring the power of a nuclear reactor having a sensor assembly disposed in a core section inside the nuclear reactor. The sensor assembly includes a cover tube made of metal and press-fitted on the outer surface of a rod-shaped core tube made of metal, a plurality of annular gaps defined between the cover tube and the core tube at different positions spaced from each other in the axial direction of the core tube, a plurality of axial grooves formed in the core tube while opening on the outer surface of the core tube and extending in the axial direction of the core tube, and a plurality of temperature sensors individually guided in the axial grooves and disposed on the core tube. The temperature sensors of the sensor assembly detect temperature distribution in portions of the core tube which face the annular gaps so as to measure the power of the nuclear reactor.

Abstract: A system for monitoring power, particularly a power distribution, of a nuclear reactor includes a plurality of neutron flux measuring units disposed in a core of the nuclear reactor for measuring neutron flux in the core and generating neutron flux signals. A unit, generally constructed by a process computer, for calculating a neutron flux distribution in the core in response to the neutron flux detection signals from the neutron flux measuring units, is provided. A unit for calculating a higher mode of the neutron flux distribution in accordance with results of calculations performed by the neutron flux distribution calculating unit, is provided. A filter calculating unit is provided for obtaining a filter for extracting characteristics of change of the neutron flux detection signal in response to the neutron flux detection signal and an input/output unit for transmitting the neutron flux detection signal filtered by the filter obtained by the filter calculating unit.

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

Abstract: A system for improving the performance of nuclear power plant feedwater control systems and simplifying steam generator low water level reactor protection logic includes a plurality of water level channels for measuring steam generator water level and generating a plurality of signals representative thereof. The median steam generator water level signal is selected from among the plurality of steam generator water level signals. The median steam generator water level signal is then communicated to the feedwater control system through an output interface.

Abstract: An emergency stop of a pressurized water reactor to be protected is triggered when the nuclear power (ET) thereof reaches an emergency stop limit (LP). The value of this limit is lower with lower operating temperatures (ST) of the reactor.

Abstract: A drive for positioning a control rod in a nuclear reactor core is disclosed. The drive includes a housing having a piston disposed therein, with a piston rod extending from the piston and through the housing for being joinable to the control rod. A driving fluid is provided into the housing for exerting a pressure force against the piston for moving the piston and the control rod. The output requirements for the driving fluid are varied in response to the position of the piston for selectively controlling intermediate positions of the piston.

Abstract: A natural circulation boiling-water reactor system using free-surface steam separation incorporates a steam separator near the inner wall of its reaction vessel and extending into a downcomer. Steam so collected supplements the main steam flow from the reactor vessel to drive a turbine, which in turn drives a generator. A flow controller opens and closes a valve to adjust the flow of steam from the steam collector to the turbine. The flow controller can be programmed to implement a predetermined power output level or to adjust power output as a function of load or time. When this level is exceeded, as indicated by feedback from the generator, the flow controller valve is constricted. The auxiliary steam flow to the turbine is directly decreased, reducing power output. Concomitantly, less steam is removed from the steam collector, and thus from the downcomer. As a result there is more carryunder steam, which reduces the power generated at the reactor core.

Abstract: A simple and accurate method for determining the enthalpy of wet steam using pressure and moisture content values. Moisture content is determined by measuring sonic velocity of the steam, which is dependent upon enthalpy. A small sample of the wet steam is superheated by pressure reduction to a single-phase state prior to measuring its sonic velocity in order to achieve greater accuracy of this measurement. From its sonic velocity, enthalpy of the sample is calculated. Since the process takes place at constant enthalpy, the enthalpy value of the sample is the same as that of the fluid from which it was drawn.

Abstract: In a nuclear electric power generating plant connected to an electric power system, an output of a nuclear reactor of the plant is reduced during a nighttime operation, increased to a predetermined level during a daytime operation and reduced during a lunch time operation in which a poisonous effect of fission products, i.e., X.sub.e.sup.135 produced during a decreased output running during a nighttime operation becomes significant.

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 mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements having differing neutron absorbing capabilities for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The displacer elements may comprise substantially hollow cylindrical low neutron absorbing rods and substantially hollow cylindrical thick walled stainless rods. Since the stainless steel displacer rod have greater neutron absorbing capability, they can effect greater reactivity change per rod. However, by arranging fewer stainless steel displacer rods in a cluster, the reactivity worth of the stainless steel displacer rod cluster can be less than a low neutron absorbing displacer rod cluster.

Abstract: The setpoints for control systems in a pressurized water reactor are adjusted by an amount corresponding to the expected change in the controlled process variable resulting from rapid fluctuations in load to reduce the duty time of the control system components such as the rod drive mechanisms in the rod control system and the spray and heater units in the pressurizer pressure control system. In addition, the deadband in the response of the rod control system is continuously varied as a function of the variance of the magnitude of the rapid fluctuations in load to further reduce wear while maintaining good response during load following.