232results 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.

A method for controlling a pressurized water reactor, computer program product and control system, the pressurized water reactor includes a reactor core and a primary cooling circuit. The primary cooling circuit includes a primary cooling medium, which includes: acquiring a plurality of measurable reactor process variables and obtaining a plurality of non-measurable reactor process variables. The method further includes calculating future axial offsets at the end of a predetermined prediction time interval for a plurality of different possible boration / dilution actions based on the plurality of measurable reactor process variables and the plurality of non-measurable reactor process variables, the axial offset being a normalized difference between power of an upper half of the reactor core and a lower half of the reactor core. The calculation of the future axial offset for each of the plurality of different possible boration / dilution actions is performed in parallel.

The invention discloses a nuclear power plant primary loop zinc injection control method, and the method comprises the steps: carrying out sampling detection, determining that the detected primary loop water chemical parameters meet the thermal state function test zinc injection conditions, and preparing a zinc ion solution; monitoring the zinc ion concentration of the primary loop, filling the zinc ion solution and adjusting the adding rate; monitoring is carried out, and zinc injection is stopped when abnormal conditions occur; after passivation timing is finished, zinc injection is stopped before boronizing; the unit commercial operation zinc injection control method comprises the steps that sampling detection is carried out to determine that hydrochemical parameters meet unit commercial operation zinc injection conditions; monitoring the zinc ion concentration of the primary loop, and filling a zinc ion solution; and zinc injection is continued after abnormal conditions are processed, and zinc injection is stopped when the conditions are met. According to the invention, primary side stress corrosion cracking can be relieved, and the external radiation field of the reactor core and the collective dose during unit overhaul can be reduced.

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.

The invention provides an intelligent diagnosis method, system and device for nuclear power plant unit parameters and a storage medium, and belongs to the technical field of nuclear power. The method comprises the steps of obtaining historical operation parameters of unit equipment; determining a pre-triggering threshold value according to the historical operation parameters; obtaining current actual operation parameters of the unit equipment; and determining a parameter diagnosis result of the unit equipment according to the current actual operation parameter and a pre-triggering threshold value. According to the method, the pre-triggering threshold value is determined based on the historical operation parameters of actual operation of the unit equipment, so that the pre-triggering threshold value is set between the normal operation parameters and the fault parameters. The current actual operation parameters of the unit equipment obtained in real time are compared with the pre-triggering threshold value, whether the parameter diagnosis result of the unit equipment is abnormal or not is determined, early warning information can be triggered in advance, an operator can find equipment abnormity in time, intervention and response time is reserved for the operator, and the unit equipment is protected in time.

The invention provides a transient working condition control device and method under a multi-module high-temperature gas cooled reactor combined heat and power working condition. The transient condition control device under the multi-module high-temperature gas cooled reactor cogeneration condition comprises a plurality of NSSS module control system modules, a unit control system module, a steam turbine control system module, a main steam pressure control system module, a main feed water pressure control system module and a heat supply station control system module. And the unit control system module is used for controlling a plurality of transient working conditions according to a plurality of received signal parameters output by the NSSS module control system modules, the steam turbine control system module, the main steam pressure control system module, the main feed water pressure control system module and the heat supply station control system module. According to the system, the multiple control system modules are additionally arranged, so that the problems that when the unit occurs in a transient state, the simple jump stop of a steam turbine or a reactor can generate large influence on external steam supply, and industrial enterprise equipment is damaged and economic losses are large are solved.

The embodiment of the invention provides a reactor control method and device, electronic equipment and a storage medium, and belongs to the technical field of nuclear power. The method comprises the steps that the type of a steam turbine and a control strategy of a nuclear power unit are obtained, and the nuclear power unit comprises a reactor; according to the type of the steam turbine and the control strategy, an energy balance signal and a heat signal are calculated, the energy balance signal is used for indicating the minimum energy needed by operation of the steam turbine, and the heat signal is used for indicating the energy generated by the reactor; obtaining a feedforward signal according to the energy balance signal and the heat signal; and controlling the rod position action of the control rod according to the feedforward signal. According to the embodiment of the invention, the response speed and the adjustment capability of the nuclear power unit in the case of rapid load change can be improved.

Disclosed is an intelligent control system for thermal energy recovery of circulating cooling water in a nuclear power station. The system comprises: a thermal energy recovery apparatus used for transferring thermal energy of circulating cooling water to other media by means of a heat exchanger; and a cooling water pump used for conveying the circulating cooling water between the thermal energy recovery apparatus and a reactor. The intelligent control system for thermal energy recovery of circulating cooling water in a nuclear power station further comprises a controller used for adjusting working parameters of the thermal energy recovery apparatus. Thus, the water consumption and discharge amount of the nuclear power station can be effectively reduced, greenhouse gas emissions are reduced, and the reliability and service life of the nuclear power station are improved.

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

Method for estimating a quantity of a system (1) comprising the following steps: - for each variable of a plurality of variables, obtaining a sequence of successive measurements, - determining a sequence of successive values ​​of a stability parameter, the parameter being a weighted sum of rates of change of the variables, - identifying a time interval in which the stability parameter is less than or equal to a predetermined threshold for a duration greater than or equal to a predetermined duration, - estimating a sequence of successive estimates of a particular variable, a temporal start of the sequence being included in the time interval, - comparing the sequence of estimates and the sequence of measurements to determine a calibration parameter, and - estimating the physical quantity of the system (1) using the calibration parameter.

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.

The invention relates to the technical field of new energy. The invention discloses a multi-loop reactor system operation method, device and equipment and a medium, and the method comprises the steps: obtaining a plurality of measurement signals of a main pump, preprocessing each measurement signal, and obtaining a main pump characteristic value corresponding to each measurement signal; according to the main pump characteristic value corresponding to each measurement signal, the main pump health degree is calculated based on a preset main pump health degree calculation rule and the main pump characteristic reference value corresponding to each measurement signal; judging the running state of the main pump according to the health degree of the main pump; and determining the system operation mode according to the mapping relation between the operation state of the main pump and the system operation mode. The health degree of the main pump is obtained through calculation of the multiple measurement signals of the main pump, the operation state of the main pump is judged according to the health degree of the main pump, the system operation mode is further determined according to the operation state of the main pump, it is guaranteed that operation of the unit is matched with the state of the main pump, and economical efficiency and safety of operation of the unit are achieved.

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 present disclosure relates to a method and apparatus for controlling a pressurized water reactor core, and a computer device and a storage medium. The method comprises: determining an operation mode of a pressurized water reactor core, wherein the pressurized water reactor core is loaded with at least one first control rod assembly, a second control rod assembly, a third control rod assembly and a plurality of fuel assemblies; acquiring a rod assembly control instruction corresponding to the operation mode; and according to the rod assembly control instruction, controlling the pressurized water reactor core on the basis of the first control rod assembly, the second control rod assembly and the third control rod assembly. By using the method, the aim of automatically controlling a pressurized water reactor core by means of a first control rod assembly, a second control rod assembly and a third control rod assembly is realized, and it is not necessary to control the pressurized water reactor core by frequently adjusting the concentration of boron in a primary circuit coolant, such that the control flow of the pressurized water reactor core is simplified, and thus the control efficiency of the pressurized water reactor core is improved.

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.

To provide a nuclear power plant capable of increasing the output control width of steam extraction throttle operation responding to the load following command from the demand side.SOLUTION: The nuclear power plant includes: an atomic reactor; a turbine; a power generator; a steam condenser; a feed water heater; an air-bleeding line that supplies part of the steam produced in the atomic reactor to the feed water heater as extraction steam; a bleeding valve that adjusts the flow rate of the air-bleeding line; and a control unit that controls the valve opening of the bleeding valve. Any one of the operation modes is selected out of a first operating mode in which the output of the power generator is constant, a second operating mode in which the output of the power generator is adjusted to follow the load and a third operating mode in which the output of the power generator is constant. In the third operating mode, at least some of the bleeding valves have a large valve opening compared with the first operating mode.SELECTED DRAWING: Figure 1

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 invention relates to a main steam pressure control method and device for a multi-module nuclear thermoelectric system, and the method comprises the steps: obtaining a main steam pressure measurement value of the multi-module nuclear thermoelectric system under the condition that the opening degree of a main steam adjustment valve of the multi-module nuclear thermoelectric system is fixed; if the deviation between the main steam pressure measurement value and the main steam pressure set value of the multi-module nuclear thermoelectric system is greater than a first threshold value, determining the adjustment amount of a water feed pump rotating speed set value and the adjustment amount of a thermal power set value of the multi-module nuclear thermoelectric system according to the deviation; and adjusting the rotating speed set value of the water feeding pump of the multi-module nuclear thermoelectric system based on the adjustment amount of the rotating speed set value of the water feeding pump, and adjusting the thermal power set value of at least one nuclear steam supply system module of the multi-module nuclear thermoelectric system based on the adjustment amount of the thermal power set value so as to reduce the deviation.

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.