170results about "Plant parameters regulation" patented technology

The present application relates to a method and system for filling water in a primary loop after overhaul charging of a nuclear power unit, comprising: performing preparation work before filling water; after the preparation work is completed, performing switch operation test of a to-be-tested actuator and water filling and exhaust work of a flowmeter; after the switch operation test of the to-be-tested actuator is completed, performing performance test of a high-pressure safety injection pump; after the performance test of the high-pressure safety injection pump is completed, performing performance test of a low-pressure safety injection pump; after the performance test of the low-pressure safety injection pump is completed, filling water in the primary loop according to a preset water filling mode; judging whether a real-time water level of a reactor pool reaches a full water level; if yes, performing alarm action test. The present application can solve the problem of long time consumption, reduce workload, significantly improve work efficiency, reduce work risk and effectively save production cost by completing the preparation work in advance and arranging the alarm action test at the end.

The invention discloses a power coordination control method of a high-temperature gas cooled reactor nuclear steam supply system, which relates to the technical field of nuclear power, and comprises the following steps of: coupling multiple modules in the high-temperature gas cooled reactor nuclear steam supply system through a common conventional island; the key state parameters of the to-be-adjusted module are used as main control variables, and the other modules are used as auxiliary control variables, so that the mutual influence relationship among the modules can be comprehensively considered; reasoning the control quantities based on a preset fuzzy rule to determine a fuzzy control strategy, so that adverse effects among modules can be counteracted, overall parameters are kept in a design value range, and safe and stable operation of the system is guaranteed; flexible adjustment can be performed according to key state parameters of different modules by using a fuzzy control strategy, the power of the reactor is quickly and accurately controlled, the dynamic response capability of the system is improved, and the requirement that the power of the reactor is quickly adjusted along with the load requirement of the steam turbine generator unit is met.

Techniques for detecting, forecasting, and reducing or avoiding disruptions in fusion devices are described. Data corresponding to operation of a fusion device may be received. Based on the data and one or more physics-based models, an occurrence of an event and a criticality level of the event may be determined. Based on criticality level of the event, whether a disruption of plasm will occur may be determined. An indication identifying the event and whether the disruption of plasma in the fusion device will occur may be output to a computing device associated with a user. In some embodiments, based, at least in part, on the criticality level of the first event, a control system associated with the fusion device may be instructed to perform one or more actions.

The invention provides a multipurpose small reactor control method, device, equipment and medium, and the method comprises the following steps: in an automatic temperature control mode, if the deviation between the actually measured temperature of a reactor outlet and the set temperature meets a preset automatic temperature control condition, determining a first control instruction of each adjusting rod of an adjusting rod group; a first control instruction is sent to an adjusting rod set executing mechanism so as to control an adjusting rod to move till the deviation between the actually measured temperature and the set temperature is within the temperature adjusting dead zone range; in the power automatic control mode, if the deviation between the set power and the actually measured power of the reactor meets a preset power automatic control condition, determining a second control instruction of the circulating pump of the intermediate loop; and a second control instruction is sent to the circulating pump executing mechanism so as to control the rotating speed of the circulating pump until the deviation between the set power and the actually measured power is within the power adjusting dead zone range. According to the invention, the temperature and the power are respectively controlled, and the control performance is optimized.

The invention discloses a method for determining the feed water flow of an evaporator of a nuclear power station and a related device, and relates to the technical field of thermoelectric conversion, and the method comprises the steps: obtaining the steam flow of an outlet of the evaporator, the secondary loop pressure of the evaporator, and the feed water temperature and nuclear power of an inlet of a main water supply pipe; the specific enthalpy coefficient is determined by looking up a table according to the secondary loop pressure of the evaporator and the feed water temperature, and the specific enthalpy coefficient is the ratio of the steam specific enthalpy to the feed water specific enthalpy; the nuclear power is converted into theoretical steam flow, the theoretical steam flow is used for verifying the steam flow of an evaporator outlet, and if verification passes, the steam flow of the evaporator outlet is used as one of the inputs of the feed water flow soft measurement model; and inputting the specific enthalpy coefficient, the steam flow at the outlet of the evaporator and the secondary loop pressure of the evaporator into a water flow soft measurement model to obtain the theoretical feed water flow. According to the method, high-precision dynamic measurement and calculation of the water supply flow of the evaporator are realized based on the water supply flow soft measurement model.

The invention discloses a reactor coolant system flow monitoring device and a signal control method, and belongs to the technical field of reactor coolant systems.The reactor coolant system comprises at least three main loops, a first pressure tapping hole located in the outer wall is formed in a main pipeline transition section elbow of each main loop, and a second pressure tapping hole located in the outer wall is formed in the main pipeline transition section elbow of each main loop; the at least four second pressure tapping holes are formed in the inner wall; the flow monitoring device comprises at least four first differential pressure sensors and at least two second differential pressure sensors; the positive pressure ends of the at least four first differential pressure sensors are communicated with the first pressure tapping hole, and the negative pressure ends of the at least four first differential pressure sensors are communicated with the at least four second pressure tapping holes in a one-to-one correspondence mode. The at least two second differential pressure sensors are arranged in parallel, and the two ends of the at least two second differential pressure sensors communicate with an inlet pipeline and an outlet pipeline of the main pump on the main loop correspondingly.

The invention relates to the technical field of compressed air system intelligent control, and discloses an intelligent control and air supply distribution method and system for a nuclear facility compressed air system pipe network. Carrying out numerical simulation on the distributed arrangement and operation conditions of different air compression stations and the flow and pressure characteristics of the compressed air pipe network under the intermittent load of different process equipment; on the basis of data generated by numerical simulation, the pipe network tail end pressure of the compressed air pipe network system is adjusted through an intelligent fuzzy self-tuning PID control method, and the operation state of an air compressor unit of the compressed air pipe network system is optimized on the basis of a flow matching strategy; pipe network operation parameters of the compressed air pipe network system are collected in real time, and the control strategy is fed back and adjusted. The whole compressed air system can operate efficiently and stably, the requirements of all links of nuclear facilities for compressed air are guaranteed, meanwhile, the operation intelligence and reliability of the system are improved, and adverse effects caused by unstable pressure and the like are reduced.

A nuclear reactor system for use with an electric power grid is disclosed, the nuclear reactor system including a nuclear reactor, an energy storage system coupled to the nuclear reactor, and control circuitry coupled to the nuclear reactor and the energy storage system, wherein the control circuitry is configured to monitor power demand on the electric power grid, monitor power output generated from the nuclear reactor, detect changes in power demand, cause the energy storage system to temporarily compensate for the changes in power demand, and adjust the power output based on the changes in power demand.

To provide a nuclear power plant and a method for generating power in a nuclear power plant which can improve the load following property of a nuclear power generation plant and can further stabilize the output of a nuclear reactor.SOLUTION: The nuclear power plant having a nuclear reactor includes: an output control unit 70 for controlling the opening degree of an air extraction valve 13 according to a load request; and an energy storage unit for storing renewable energy. The output control unit 70 controls the temperature of water to supply to the nuclear reactor on the basis of the opening degree of the air extraction valve 13 and the amount of stored energy in the energy storage unit.SELECTED DRAWING: Figure 1

To adjust a core flow rate higher than that in a stationary state when a power change operation of a natural circulation reactor is conducted by a control rod, thus improving a power change speed while maintaining a stable condition.SOLUTION: A nuclear power plant includes a large-diameter space chimney at an upper part of a reactor core within a shroud of a reactor pressure vessel. The nuclear power plant includes: a coolant flowmeter configured to measure a coolant flow flowing into the reactor core; a water level gauge configured to measure a water level of a coolant within the reactor pressure vessel; a water-supply flowmeter connected to the reactor pressure vessel and configured to measure a water-supply flow rate; a monitoring device configured to monitor the coolant flow, the water level and the water-supply flow rate; and a controlling device configured to control a control rod and the water-supply flow rate. The controlling device is configured to: calculate, in advance, a fluctuation margin of a core flow rate by a time lag of the water-supply flow rate; change a target value of the water level to be higher than a fluctuation margin of the water level corresponding to the fluctuation margin of the core flow rate so as to change the water-supply flow rate and then change the water level; and thereafter operate and control a position change of the control rod.SELECTED DRAWING: Figure 4

A control rod drive mechanism having a torque tube, a control rod assembly including a connecting rod disposed within the torque tube and including an annular collar defining a key slot, an elongated key that is slidably receivable within the key slot, a lock cam assembly rotatably secured to a top end of the connecting rod and including a locking cam, and a holdout collar disposed non-rotatably within the torque tube, the holdout collar including a locking recess with an entry slot, wherein in a first axial position the lock cam assembly is rotatable with respect to the torque tube, and a second axial position the lock cam assembly is non-rotatable with respect to the torque tube.

A nuclear fission power plant (10) configured for extra-terrestrial use comprising a gas-cooled moderated nuclear reactor core (20) with a metallic fuel (30), a moderator (60), and a containment vessel (40) with a containment vessel inlet (42) and a containment vessel outlet (44). A neutron reflector (70) is disposed around a periphery of the gas-cooled moderated nuclear reactor core, and a plurality of neutronic control elements (50) comprising rotatable control drums disposed around the periphery of the gas-cooled moderated nuclear reactor core. A fluid circuit (90) is configured to receive hot gas from the containment vessel outlet and deliver cooled gas to the containment vessel inlet. An electricity generating system (91) is coupled to the fluid circuit so as to generate an electrical current.

Systems and methods provide data gathering and execution on the same without human operations. Systems may include controls and sensors that electronically provide data and operations to a processor networked with the same. For a nuclear reactor, the processor may determine reactivity from the sensors and issue commands to actuators to operate the reactor. Reactivity may be determined based on all reactivity 10 factors determined from the plant data, including the use of modelling. The processor may position control elements or moderator feeds to achieve a desired reactivity. The processor may be networked to plant switches and sensors, and multiple processors may be used to independently calculate and decide on plant operations. Human operator input is not required at discreet instances of plant operational change; systems 15 may include displays and input interfaces to permit observation and / or intervention if absolutely necessary.

This invention provides a method for controlling the oxygen content in reactor coolant, comprising: determining the relationship between the growth thickness of the oxide layer formed on the surface of the fuel rod cladding within a preset time period under conditions of no coolant corrosion and the oxygen content in the coolant; determining the removal rate of the oxide layer on the surface of the fuel rod cladding after reaction with the coolant; determining a functional relationship between the remaining oxide layer thickness on the cladding surface and the oxygen content of the coolant under varying oxygen content conditions, based on the relationship between the oxide layer growth thickness and the oxygen content in the coolant and the oxide layer removal rate; using the functional relationship between the remaining oxide layer thickness on the cladding surface and the oxygen content of the coolant as the objective function, and determining the optimal oxygen content control strategy of the coolant corresponding to the minimum value of the remaining oxide layer thickness based on a preset optimization algorithm. This invention alleviates the oxidation and corrosion problem of fuel rod cladding materials in metallic coolants by determining the optimal oxygen content control strategy of the coolant.

The application discloses a method for determining feedwater flow of an evaporator of a nuclear power plant and related devices, and relates to the technical field of thermoelectric conversion. The method comprises the following steps: obtaining steam flow at an outlet of the evaporator, secondary loop pressure of the evaporator, feedwater temperature at an inlet of a main feedwater pipe, and nuclear power; using the secondary loop pressure of the evaporator and the feedwater temperature to determine a specific enthalpy coefficient by table lookup, the specific enthalpy coefficient being a ratio of specific enthalpy of steam to specific enthalpy of feedwater; converting the nuclear power into theoretical steam flow, and verifying the steam flow at the outlet of the evaporator by using the theoretical steam flow, if the verification is passed, using the steam flow at the outlet of the evaporator as one of inputs of a feedwater flow soft measurement model; inputting the specific enthalpy coefficient, the steam flow at the outlet of the evaporator and the secondary loop pressure of the evaporator into the feedwater flow soft measurement model to obtain a theoretical feedwater flow. The application realizes high-precision dynamic measurement and calculation of the feedwater flow of the evaporator based on the feedwater flow soft measurement model.

A hybrid power generation system which includes a nuclear facility comprising a nuclear reactor and an exclusion zone. A thermal energy storage, a nuclear steam supply system located, and a solar energy collection system are all located within the exclusion zone. The thermal energy storage vessel contains a thermal mass composition operable to store thermal energy. The nuclear steam supply system has a nuclear reactor and a working fluid loop. The working fluid loop is configured to circulate a working fluid from a steam generator through the thermal energy storage vessel to absorb thermal energy and heat the working fluid for introduction to an electricity generating system. The solar energy collection system includes a heat transfer loop heated via a solar collector. The heat transfer loop is configured to circulate a heated heat transfer fluid to add thermal energy to the thermal mass composition in the thermal energy storage vessel.

The present invention relates to a system and a method for automatically setting a main steam condensate drain valve closing setpoint of a nuclear power plant, and, specifically, to a system and a method for automatically setting a main steam condensate drain valve closing setpoint of a nuclear power plant, which apply different setpoints to respective condensate drain valves since the pressure of main steam and a condenser of a nuclear power plant varies according to operating conditions such as turbine output, and the discharge speeds of the condensate vary according to the different installation positions of condensate drain valves.

The present invention relates to a method of operating an MSR, said MSR comprising a reactor core, a fuel loop for molten fuel salt providing a flow path of molten fuel salt into the reactor core through a reactor core inlet and out of the reactor core through a reactor core outlet, said method comprising the steps of: providing at least one location mi on the fuel loop at a distance Li measured along the flow path from the reactor core outlet, wherein the one or more locations mi is outside the reactor core, and the one or more locations mi is shielded from the reactor core measuring the neutron source strength at the one or more locations mi obtaining one or more calculated operating parameters by calculation with input of the value of the measured neutron source strength at one or more locations mi.

The present invention provides a method for controlling the reactivity of a boric acid-free reactor in a secondary system of a pressurized light-water reactor, wherein: a turbine and a main water feeding system of a pressurized light-water reactor comprise a steam generator (S / G), a high-pressure turbine, a low-pressure turbine, a high-pressure feedwater heater, and a low-pressure feedwater heater; the high pressure turbine is provided with extracted steam ① and ② capable of heating feedwater, the low pressure turbine is provided with extracted steam ③ and ④ capable of heating feedwater, the high-pressure feedwater heater is provided with high pressure heaters ① and ② for receiving extracted steam from the turbines to heat feedwater, and a low-pressure feedwater heater is provided with low pressure heaters ③ and ④ for receiving extracted steam from the turbines to heat feedwater; and when operated with constant reactivity and temperature set as default values, the high-pressure heater ① receives extracted steam ② of the high-pressure turbine to heat feedwater, the high-pressure heater ② receives extracted steam ① of the high-pressure turbine to heat feedwater, the low-pressure heater ③ receives extracted steam ④ of the low-pressure turbine to heat feedwater, and the low-pressure heater ④ receives extracted steam from the low-pressure turbine ③ to heat feedwater.

The invention provides a method, a device and a product for adjusting a quadrant power inclination value of a reactor core, and the method comprises the steps: determining a quadrant power inclination value exceeding a preset threshold value in the reactor core, and obtaining a target quadrant power inclination value; determining a corresponding target temperature adjusting rod according to the target quadrant power inclination value, and determining a target out-of-step number of the target temperature adjusting rod; and performing out-of-step operation on the target temperature adjusting rod according to the target out-of-step number so as to adjust the target quadrant power inclination value. According to the method, the power of the corresponding quadrant is pointedly reduced by locally adjusting the temperature adjusting rod, so that direct economic loss caused by power reduction is avoided; the economic benefit influence during the supplementary safety evaluation period is avoided; the reactor is prevented from being in a reactor core operation limiting state for a long time, and the influence on the safe operation margin is reduced. Therefore, the method can guarantee the economic benefit of the nuclear power plant and improve the margin of safe operation when the quadrant power inclination value of the reactor core of the reactor is adjusted.

The present invention provides a control method, device and electronic equipment for a helium-xenon cooled reactor, which relates to the technical field of nuclear reactors. The above-mentioned control method for a helium-xenon cooled reactor includes: during the operation of the helium-xenon cooled reactor, monitoring the operation state information of the helium-xenon cooled reactor, and obtaining a mathematical model of the helium-xenon cooled reactor constructed in advance; performing numerical simulation on the mathematical model based on the current operation state information to obtain a numerical simulation result; adjusting each control parameter of the helium-xenon cooled reactor based on the numerical simulation result so that the power generation power of the helium-xenon cooled reactor reaches a preset power generation power. The present invention can adaptively control each control parameter of the reactor according to the operation state of the reactor, and improves the reliability of the control of the helium-xenon cooled reactor.

The application relates to the technical field of periodic testing, and discloses a safety injection logic test optimization method, device, equipment and medium for a nuclear power station. The method comprises the following steps: in the process of performing a safety injection logic test on the nuclear power station, if it is detected that a fault scene simulation of the nuclear power station is ended, a small-flow electric valve on a small-flow pipeline of an RCV pump of the nuclear power station is controlled to be started, and after the small-flow electric valve is started, an adjustable isolation valve is controlled to be adjusted from a closed state to a fully open state within a set opening time range, so that the adjustable isolation valve is used for slowly injecting reactor coolant into a primary pump No. 1 sealing leakage flow pipeline of the nuclear power station through the small-flow pipeline of the RCV pump, wherein the adjustable isolation valve is installed on the small-flow pipeline of the RCV pump and located upstream of the small-flow electric valve. The method can improve the safety of nuclear power station equipment operation.

The invention discloses an intelligent control system for heat energy recovery of circulating cooling water in a nuclear power station. The system comprises a heat energy recovery device used for transferring heat energy of circulating cooling water to other media through a heat exchanger; the heat energy recovery device is used for receiving circulating cooling water, the cooling water pump is used for conveying the circulating cooling water to the position between the heat energy recovery device and a reactor, and the controller is used for adjusting working parameters of the heat energy recovery device. Therefore, the water consumption and the discharge amount of the nuclear power station can be effectively reduced, the discharge of greenhouse gases is reduced, and the reliability and the service life of the nuclear power station are improved.

The invention relates to the technical field of reactor power flow control, in particular to an automatic rapid power and flow reduction control method for a sodium-cooled fast reactor, which comprises the following steps: step 1, when the reactor power is greater than 80% of rated power, setting a target set value of the reactor power; 2, a reactor power adjusting system is adjusted to be in a manual mode, and a first loop flow control system and a second loop flow control system are adjusted to be in an automatic mode; 3, the rod control and rod position indication system controls the first compensation rod to execute descending action; 4, when the reactor power is smaller than or equal to 80% of rated power, the rod control and rod position indicating system controls the first compensation rod to stop descending, and the reactor power adjusting system adjusts the reactor power to a power target set value. According to the invention, the effect of rapidly reducing the nuclear power and the flow of the primary loop and the secondary loop to 75% of the rated power level is realized through a non-rod-falling type automatic rapid power and flow reduction technical means.

The invention is suitable for the technical field of system control, and provides a control method for the rotating speed of a main water supply pump of a nuclear power plant and a steam generator system.The method comprises the steps that when it is detected that a main water supply valve breaks down, the first rotating speed of the main water supply pump is determined according to the first output power of the steam generator system at the current moment; correcting the first rotating speed according to the numerical value of the process parameter of the steam generator system to obtain a corrected second rotating speed; and the rotating speed of the main water supply pump is adjusted to be the second rotating speed. According to the method, under the condition that the main water supply valve of the steam generator breaks down, the rotating speed of the main water supply pump can be timely and accurately adjusted, and the loss of the nuclear power station can be reduced.

The invention discloses a pressurized water reactor nuclear power unit power adjusting system and method, and belongs to the technical field of pressurized water reactor nuclear reactors, the system comprises a pressurized water reactor nuclear steam supply system and a bypass system arranged on a steam generator secondary side steam pipeline of the pressurized water reactor nuclear steam supply system; wherein the bypass system comprises a steam bypass subsystem, a composite material heat storage subsystem and an air power generation subsystem; the steam bypass subsystem is used for introducing excess steam in the pressurized water nuclear reactor nuclear steam supply system into the composite material heat storage subsystem for heat energy storage, and the air power generation subsystem absorbs the heat energy stored in the composite material heat storage subsystem through compressed air to drive a turbine to generate power. Under the condition that the operation power of the nuclear reactor is not changed, the steam heat energy of the bypass steam turbine is stored in a mode of sensible heat energy of the composite material, the problem of mismatching between power supply and power demand is solved, and power adjustment of the pressurized water reactor nuclear power unit is achieved.

A method and system for calibrating nuclear instrumentation of a reactor is disclosed. The system comprises a reactor configured to operate at a low power level and an external heater configured to heat the reactor to its operating temperature and maintain the reactor in a critical state. The reactor maintains its temperature at a steady state by repeating steps of reducing the heater's power and increasing the reactor's power by changing one or more reactivity settings, wherein the reduced heater's power is compensated by using the increased power generated by the reactor itself. The nuclear instrumentation is configured to measure the reactor's output power and compare it with the calculated reactor's power. Based on the discrepancy between the calculated power and the measured power exceeds a threshold; and the nuclear instrumentation is appropriately calibrated by adjusting one or more settings.

To obtain an operation plan having high operability including width and a period of adjustment force.SOLUTION: A nuclear power generation plant 100 includes a power generation system 101 and an output control device 102. The output control device includes: an evaluation criterion input device 28 for inputting a market price prediction value of wholesale power and a consumption adjustment; an adjustment force setting device 26 for setting an adjustment force supply amount for adjusting consumption from information on a price prediction value and a reactor core / plant state including a margin for a thermal restriction of at least a reactor core and a control rod position; an operation plan evaluation device 25 for generating an operation plan from an adjustment force supply amount and evaluation propriety of the plan; a dynamic characteristic evaluation system 24 for evaluating a reactor core / plant state on the basis of an operation plan; and an operation plan output device 27 for outputting an operation plan obtained by a repeated evaluation of the adjustment force setting device and the operation plan evaluation device.SELECTED DRAWING: Figure 1

Systems and methods provide data gathering and execution on the same without human operations. Systems may include controls and sensors that electronically provide data and operations to a processor networked with the same. For a nuclear reactor, the processor may determine reactivity from the sensors and issue commands to actuators to operate the reactor. Reactivity may be determined based on all reactivity factors determined from the plant data, including the use of modelling. The processor may position control elements or moderator feeds to achieve a desired reactivity. The processor may be networked to plant switches and sensors, and multiple processors may be used to independently calculate and decide on plant operations. Human operator input is not required at discreet instances of plant operational change; systems may include displays and input interfaces to permit observation and / or intervention if absolutely necessary.

The application provides a unit control method, device and equipment suitable for a third generation pressurized water reactor, and the method comprises the following steps: if it is detected that a single-column AHP system fails to be isolated, the feedwater temperature of a secondary loop feedwater system is monitored in real time; if the difference of the feedwater temperature in a preset unit time is greater than a preset feedwater deviation value, the real-time operation power of the unit is obtained; whether the real-time operation power meets a preset unit power adjustment condition is judged according to a preset operation control strategy; if yes, the operation power of the unit is adjusted according to the unit power adjustment value in the operation control strategy. The embodiment of the application can adjust the operation power of the unit according to the unit power adjustment value in the operation control strategy, effectively improves the availability of the unit operation after the single-column AHP system is isolated, and ensures the safety and reliability of the unit operation.