89results about "Steam generator primary side" patented technology

The invention discloses a self-adaptive control method and system for a condenser for a nuclear power station, relates to the field of condenser control, and aims to predict a future load change trend in a prospective manner by analyzing historical data of a unit power instruction and steam flow, and calculate feed-forward circulating water flow based on a prediction result. On the basis, a feedback control loop based on actually measured pressure is reserved. More importantly, a self-adaptive bias correction mechanism is introduced, and continuous learning and correction are carried out on output of feedback control, so that steady-state errors of the system are eliminated, and uncertainty of the model is compensated. Finally, feed-forward, feedback and self-adaptive correction components are intelligently fused to form a final control instruction which can quickly respond to load changes and can accurately maintain pressure stability, and the problems of severe pressure fluctuation and energy waste under the dynamic working condition are effectively solved.

The invention provides a nuclear power plant primary side post-accident passive residual heat removal system and a control method, the system comprises a reactor coolant system, a containment inner water tank, a collecting mechanism and a passive water replenishing tank, and the reactor coolant system comprises a hot section and a cold section; the containment inner water tank is installed on the inner wall of the steel containment, the heat exchanger is arranged in the containment inner water tank, the hot section is connected with an inlet pipeline of the heat exchanger, the cold section is connected with an outlet pipeline of the heat exchanger, and the collecting mechanism is arranged on the inner wall of the steel containment and communicated with the containment inner water tank through a connecting pipe. The collecting device is used for collecting condensate on the inner wall of the steel containment in a grading manner along the axial direction of the containment and enabling the condensate to flow to a water tank in the containment through a connecting pipe; one end of the passive water replenishing tank is connected with the water tank in the containment through a runner pipe, and the passive water replenishing tank replenishes water to the water tank in the containment when the liquid level of the water tank in the containment drops to a target value.

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 belongs to the technical field of austenitic stainless steel pipes, and particularly relates to an austenitic stainless steel pipe for a heat exchanger and a preparation method and application thereof. The austenitic stainless steel pipe for the heat exchanger comprises the following components in percentage by mass: less than or equal to 0.03% of C, less than or equal to 0.75% of Si, 3-6% of Mn, less than or equal to 0.04% of P, less than or equal to 0.03% of S, 16-19% of Cr, 4-7% of Ni, 2-4% of Mo, 1-3% of Cu, 0.3-0.6% of Hf, 0.2-0.5% of Ti, 0.1-0.3% of Co and the balance of Fe and inevitable impurities. The austenitic stainless steel pipe for the heat exchanger has excellent mechanical performance and is resistant to chloride ions and microbial corrosion.

A nuclear power plant comprising at least one nuclear reactor having a steam generator, a steam cycle including the steam generator, a first turbine fluidly coupled to the steam generator, a first condenser fluidly coupled to the first turbine, and a feedwater heater fluidly coupled to the first condenser and the steam generator, an intermediate thermal loop including the first condenser, and an evaporator, and an organic cycle including the evaporator, a second turbine fluidly coupled to the evaporator, a second condenser fluidly coupled to the second turbine, and a heat exchanger fluidly coupled to the second condenser and the evaporator.

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.

The application provides a high-temperature gas cooled reactor type supercritical ultra-supercritical boiler steam generator, which comprises a shell and a heat exchange device arranged in the shell, a maintenance space for people to enter is arranged between the shell and the heat exchange device, the shell is provided with a hot helium gas inlet and a cold helium gas outlet connected with a reactor, and the application further comprises a helium gas isolation valve for opening and closing the hot helium gas inlet and the cold helium gas outlet. In this way, since the helium gas isolation valve can close the hot helium gas inlet and the cold helium gas outlet, the helium gas passage with the reactor can be cut off in an accident, and the supply of high-temperature helium gas is stopped, so that the safety level of the steam generator is reduced, the steam generator is no longer limited by the nuclear power safety I level design and manufacturing specification, and the thermal power boiler design and manufacturing specification can be used. Moreover, the helium gas cannot be activated, the radiation dose inside the boiler steam generator is very low, and personnel can enter the maintenance space to check and maintain the boiler steam generator.

A nuclear plant steam generator is provided. The steam generator has an upright elongate body and has tension members extending from proximal ends attaching to respective attachment positions on the elongate body to distal ends which are configured to attach to respective load-supporting attachment positions in the nuclear plant. Each proximal end is lower than its respective distal end. In use, the tension members are loaded in tension by the steam generator to transmit substantially all of the weight of the steam generator to the load-supporting attachment positions in the nuclear plant.

A heat exchanger includes a body having an inlet header through which a fluid is introduced, and an outlet header through which the fluid is discharged; and one or more plates accommodated in the body and provided with flow path modules providing flow paths for the fluid introduced through the inlet header to flow to the outlet header. The heat exchanger further includes at least one flow path adjuster each having at least a portion thereof accommodated in the body and being movable or rotatable to open or close a part or all of the flow paths or to change directions of the flow paths so that a flow of the fluid is adjusted.

According to the method for detecting the flow instability boundary of the spiral tube once-through steam generator, influence factors such as heating power, mass flow, pressure, inlet supercooling degree, inlet throttling resistance and tube shell side coupling are comprehensively considered, and the flow instability boundary of a spiral tube once-through steam generator system can be accurately described. And support is provided for design and safety analysis of the spiral tube once-through steam generator.

The invention relates to the technical field of nuclear power control, in particular to a response method for tripping of a single main feed pump under the full-power working condition of a nuclear power unit. The method comprises the steps that during power operation, operation parameters and states of subsystems and equipment of a power plant are continuously monitored, and when it is monitored that a reactor operates at high power and a single main water feeding pump of a main water feeding system trips, a steam turbine load shedding signal is triggered, and meanwhile a rod falling signal is generated; hard wiring is added between the control rod cabinet and the rod controller, and after a rod falling signal is received, pin claw coils of the corresponding pre-selected control rod group are all de-energized, so that the pre-selected control rod group quickly falls into the reactor bottom under the action of gravity, and the reactor power is reduced to PL. According to the method, the power of the steam turbine is rapidly reduced, and the predetermined control rod group is rapidly inserted into the reactor core, so that the nuclear power is rapidly reduced, the power balance of the first loop and the second loop is maintained, the SG liquid level is not too low, and shutdown is avoided.

The present application discloses a reactor safety system combining active and passive functions and a control method thereof. The reactor safety system includes a safety injection subsystem combining active and passive functions, a containment energy management subsystem, a secondary side heat removal subsystem, and a reactor cavity water injection subsystem; wherein the safety injection subsystem includes two DVI safety injection pipelines and two independently arranged safety injection sub-trains, each DVI safety injection pipeline includes a DVI pipe, which receives the safety injection flow from a safety injection pump in each of the two safety injection sub-trains, and injects the safety injection flow into the reactor through the DVI pipe after being cooled by a series-connected safety injection heat exchanger; the control method includes: for a large breach accident, under a single fault criterion, providing the reactor with the safety injection flow of at least two safety injection boxes and two safety injection pumps, with the safety injection boxes providing a large short-term safety injection flow in the early stage, and the safety injection pumps providing a continuous safety injection flow thereafter; for other accidents, at least one DVI safety injection pipeline is effective.

This application provides a method and apparatus for water injection in the primary loop of a nuclear reactor, belonging to the field of nuclear control. The method includes: obtaining the original enthalpy and residual enthalpy of the reactor core; obtaining the critical time for successful water injection and the critical time for unsuccessful water injection corresponding to the primary loop of the reactor core; determining the critical total flow rate for successful water injection in the primary loop based on the critical time for successful water injection, the original enthalpy and the residual enthalpy, and the energy conservation principle of the reactor core; and determining the critical total flow rate for unsuccessful water injection in the primary loop based on the critical time for unsuccessful water injection, the original enthalpy and the residual enthalpy, and the energy conservation principle of the primary loop; and controlling the total flow rate of water injection in the primary loop based on the critical total flow rate for successful water injection and the critical total flow rate for unsuccessful water injection. This application embodiment can improve the efficiency of primary loop water injection.

The system comprises a plurality of nuclear reactors that use coated fuel particles, a moderator and a coolant, a steam generator, and a steam turbine in communication with the steam generator, with the plurality of nuclear reactors being connected to one of the steam generators.

The invention discloses movable nuclear energy equipment, which is divided into a plurality of modules which can be independently transported and can be mutually butted into a whole based on functions, the movable nuclear energy equipment comprises a loop system module and a working module, the loop system module is integrally packaged by a shielding body, and a reactor body used for generating heat and a reactor body used for generating heat are integrally arranged in the loop system module; the loop pipeline is connected between the reactor body and the intermediate heat exchanger; the first fan is arranged on the primary loop pipeline and is used for driving a primary loop medium in the primary loop pipeline to absorb heat of the stack body and conveying the heat to the intermediate heat exchanger; the intermediate heat exchanger is used for exchanging heat between a primary loop medium and a working medium of the working module; and the working module is detachably butted with the loop system module and is used for supplying energy through the working medium subjected to heat exchange in the intermediate heat exchanger. The equipment disclosed by the invention does not need civil construction, can be quickly deployed, and has relatively high maneuverability and relatively high factory site adaptability. The invention further provides a nuclear energy system.

The invention relates to the technical field of nuclear power reactors, and discloses a nuclear power unit secondary circuit support system and a once-through steam generator nuclear power system. According to the nuclear power unit secondary loop support system, a plurality of groups of steam generator output pipelines are communicated with a main steam pipeline in parallel; the start-stop reactor main pipe is communicated with the output pipeline of the steam generator through a start-stop branch, a steam-water separator is mounted on the start-stop reactor main pipe in parallel, a steam supply branch is communicated between the start-stop reactor main pipe and the main steam pipeline, the steam supply branch is communicated with a steam exhaust pipeline, and two groups of steam exhaust outlets are formed in the tail end of the steam exhaust pipeline in parallel; and the drainage pipeline is communicated with the liquid phase outlet of the steam-water separator, and two groups of drainage outlets are arranged on the drainage pipeline in parallel. Start-stop steam exhaust of multiple groups of steam generators is treated through the shared start-stop reactor main pipe and the steam-water separator, graded recycling of working media and heat energy in the start-stop process is achieved, vapor-phase media are recycled, and waste of heat energy resources and water resources in the initial stage of start is reduced.

The application belongs to the field of coolant purification of nuclear power plants, and particularly relates to a coolant circulating purification system and method applied to nuclear power plants. The system comprises: a primary coolant storage system cation exchanger, a primary coolant storage system anion exchanger, a primary coolant storage system resin collector, a primary coolant treatment system cation exchanger, a primary coolant treatment system anion exchanger, and a primary coolant treatment system resin collector. A third bypass is arranged in front of the inlet of the primary coolant treatment system cation exchanger, and a valve is arranged in the third bypass to connect to a position after the inlet of the primary coolant storage system cation exchanger. The application can effectively improve the reliability of the primary coolant treatment system in purifying the coolant, and can switch the primary coolant storage system ion exchanger to operate in the case of failure of the primary coolant treatment system ion exchanger.

Disclosed is a micro nuclear reactor that includes a core filled with nuclear fuel and moderator which are formed of particles, a heat pipe inserted in the core and transferring outwards heat generated by a nuclear reaction, and a power converter receiving heat from a condenser of the heat pipe and converting thermal energy into electrical energy.

The invention provides a method for reducing tritium emission of a pressurized water reactor unit, which comprises the following steps of: discharging a primary loop coolant into a refueling water tank in a containment vessel for temporary storage for t time, reducing tritium concentration in the primary loop coolant by utilizing natural decay, and then discharging the primary loop coolant so as to reduce the tritium emission of the pressurized water reactor unit, the t time is a fuel circulation period of the pressurized water reactor unit. The method is reasonable in design and easy to operate, the primary loop coolant is temporarily stored through the refueling water tank in the existing containment, the tritium concentration in the primary loop coolant is reduced through tritium natural decay, tritium emission of the pressurized water reactor unit can be reduced, additional devices and equipment are not needed, and the tritium treatment cost of the pressurized water reactor unit cannot be increased.

The invention discloses a test device suitable for energy utilization of a high-temperature gas cooled reactor and a working method of the test device. The device comprises a fused salt heat exchange system, a primary loop helium system and a secondary loop helium system, the fused salt heat exchange system comprises a fused salt-helium heat exchanger, the primary loop helium system comprises a primary loop heat regenerator and a helium cooler, and the secondary loop helium system comprises a gas-gas heat exchanger, a secondary loop heat regenerator and a cooling device. Helium cooled by the helium cooler sequentially enters the primary loop heat regenerator and the fused salt-helium heat exchanger to be heated and then enters the gas-gas heat exchanger to heat SCO2, and therefore heat of the primary loop is transferred to the secondary loop. The test device can be used for analyzing coupling characteristics of a flow field, a temperature field and a stress field of the gas-gas heat exchanger under various operation conditions; the gas-gas heat exchanger is heat exchange equipment for connecting and isolating a primary loop and a secondary loop, and can ensure that heat emitted by a reactor is transferred from the primary loop to the secondary loop under various expected operation conditions when the high-temperature gas cooled reactor is coupled.

The present application relates to a kind of nuclear power steam generator pipe blocking equipment, by the setting of driving device and fluid guide device, greatly simplify the operation process of pipe blocking, reduce the work intensity of operator. By the cooperation of hydraulic guide assembly and gas guide assembly, the pipe blocking work of plug to heat pipe can be quickly completed, the time required for pipe blocking is reduced, and the efficiency of pipe blocking operation is improved. Further, the nuclear power steam generator pipe blocking equipment can be applied to heat pipes of different sizes and types, and can be used in different nuclear power plant environments, with good versatility and adaptability.

The invention belongs to the technical field of nuclear energy, and particularly relates to a high-temperature gas cooled reactor primary loop system which comprises a reactor pressure vessel, reactor internals, a high-temperature gas cooled reactor core, a fused salt heat exchanger and a hot gas conduit. Reactor internals are mounted in the reactor pressure vessel, and a high-temperature gas cooled reactor core is filled in an inner cavity of the reactor internals. And a low-temperature helium inlet is formed in the top of the reactor internals. And one end of the hot air conduit is connected with the high-temperature gas cooled reactor core, and the other end of the hot air conduit is connected with the molten salt heat exchanger to form a circulation loop. Helium-molten salt heat exchange is realized by adopting the molten salt heat exchanger, so that the problems that the helium side and the water side of a steam generator adopted at present are both under high-temperature and high-pressure working conditions, a complex auxiliary system needs to be arranged, the system is huge, the equipment cost is high, and a reactor core water inlet accident is easily caused by breakage of a heat exchange pipe on the water side of a secondary loop of the steam generator are solved.

The invention relates to the field of nuclear power operation, in particular to a system and method for nuclear power plant peak regulation through a nuclear power plant auxiliary steam heat supply system. The method comprises the steps that when a unit in a nuclear power plant operates in a peak-load-regulation and load-reduction mode, saturated steam provided by the nuclear power unit is used for heating a heat storage heat exchanger, and heating steam condensation water returns to a unit condensation water system; cold water in the hot water storage tank is heated by the heat storage heat exchanger and then returns to the hot water storage tank for heat storage; when the heat supply unit in the nuclear power plant operates in a heat supply working condition, hot water in the hot water storage tank heats the heat release heat exchanger, and cold water after heat release returns to the bottom of the hot water storage tank; feed water of the heat supply system is conveyed to the heat release heat exchanger and is conveyed to the first-stage preheater of the heat supply system after being heated; or the hot water in the hot water storage tank is directly supplied to the first-stage preheater of the heat supply system; when the heat supply system of the heat supply unit of the nuclear power plant stops running, hot water in the hot water storage tank heats the heat release heat exchanger, and cold water after heat release returns to the bottom of the hot water storage tank; and unit condensation water is conveyed to the heat release heat exchanger, heated and then returned to the unit condensation water system. The method reduces the loss of peak-load-reduction operation of the nuclear power unit, improves the electric power of the heat supply unit in the nuclear power plant during operation, and improves the economic benefits of the nuclear power plant.

The present invention discloses a nuclear energy steam supply system, comprising a nuclear energy system, a steam supply system, and a seawater desalination system. The nuclear energy system is suitable for generating nuclear steam. The steam supply system comprises a steam extraction pipeline and a steam network. The steam extraction pipeline is connected to the nuclear energy system and is suitable for nuclear steam to pass through the steam extraction pipeline. The seawater desalination water in the steam network is suitable for heat exchange with the nuclear steam in the steam extraction pipeline to generate industrial steam. The seawater desalination system comprises a seawater desalination device, a heat source pipeline, and a supply pipeline. The seawater desalination device is arranged on the heat source pipeline and the supply pipeline. The seawater desalination device is suitable for generating seawater desalination water. One end of the heat source pipeline is connected to the steam extraction pipeline, and the other end of the heat source pipeline is connected to the nuclear energy system. The supply pipeline is connected to the steam network and is suitable for transporting seawater desalination water to the steam network. The nuclear energy steam supply system of the present invention can realize the supply of industrial steam based on nuclear energy, achieving clean and efficient industrial steam supply.

The invention is applicable to the technical field of nuclear power stations, and provides a drainage control method and device, terminal equipment and a storage medium, and the method comprises the following steps: starting a target water pump after loop circulation of a secondary loop system is established; the opening degree of the water outlet valve of the target water pump is adjusted according to the water temperature change condition of the secondary loop system, water of loop circulation of the secondary loop system can be controlled to be drained in advance under the condition that the cooling rate is not affected after loop circulation of the secondary loop system is established, the drainage rate can be adjusted according to the cooling efficiency, and the cooling efficiency of the secondary loop system is improved. Water is drained in advance under the condition that water circulation cooling of the secondary circuit system is guaranteed, so that water of the secondary circuit system can be emptied more quickly, the time required by water drainage of the secondary circuit system of an overhaul unit can be greatly shortened, the influence of water drainage of the secondary circuit system on an overhaul path is reduced, and the overhaul efficiency is improved.

Method (100) for checking a state of a tube sheet (41) and of a plurality of tubes in a steam generator (3), the method (100) comprising the following steps: acquisition (101) of at least two images by a movable image acquisition device inside the steam generator (3), one of the at least two images being acquired by the acquisition device in one position, and another of the at least two images being acquired by the acquisition device in another position, so that an image of the portion of the tube(s) (4) and / or of the tube sheet (41) is present in each of the at least two images acquired;determination (102) of the 3D coordinates, in a reference frame attached to the steam generator (3), of points on the tube(s) (4) and / or points on the portion of the tube sheet (41), from at least two acquired images, and from parameters representative of the tubes (4) and the tube sheet (41), the determination (102) of the 3D coordinates being carried out by computer-implemented photogrammetry. Figure 3;

The application provides a nuclear energy steam heat supply system and method capable of improving superheat degree, a second outlet of a steam generator is communicated with an inlet of a steam turbine and an inlet of an electric heating boiler through pipelines respectively, an outlet of the electric heating boiler is connected to a steam user through a pipeline, the steam turbine is connected with a generator, an outlet pipeline of the steam turbine is communicated with a first inlet of a condenser, a first outlet of the condenser is communicated with a condensate tank, the condensate tank is communicated with a second inlet of the steam generator through a second main pump, a storage and discharge circuit comprises a storage and discharge module, inlets of the storage and discharge module are connected with a power grid and the generator respectively, an outlet of the storage and discharge module is connected with the power grid and each electric equipment, and the application effectively improves the superheat degree of the secondary circuit steam and can meet the temperature parameter requirements of various users.

One variation of a system includes a pressure vessel: including a wall; defining a primary working fluid circuit extending vertically within the wall; defining a secondary working fluid circuit extending vertically within the wall and fluidly isolated from the primary working fluid circuit; and configured to store a nuclear fuel, a primary working fluid, and a secondary working fluid. The wall of the pressure vessel: defines a heat exchanger configured to transfer thermal energy from the primary working fluid flowing through the primary working fluid circuit into the secondary working fluid flowing through the secondary working fluid circuit; and defines a radiation shield configured to attenuate radiation emitted by the nuclear fuel.

The present invention discloses a steam generator accident protection system and its working method. The pipe-side outlet of the steam generator is connected to the steam output pipeline via a steam rapid isolation valve. An emergency drain valve is provided on the pipeline between the water supply rapid isolation valve and the steam generator. An atmospheric release valve is provided between the steam generator and the steam rapid isolation valve. The outlet of the inlet sodium valve is connected to the sodium inlet of the steam generator shell side, the inlet of the outlet sodium valve is connected to the sodium outlet of the steam generator shell side, and a flow differential meter is connected to the outlet of the inlet sodium valve and the inlet of the outlet sodium valve. A steam generator air cavity is provided at the upper portion of the steam generator shell side, a hydrogen-in-gas meter is installed at the top of the steam generator air cavity, and a hydrogen-in-sodium meter is installed at the lower portion of the steam generator shell side. This system and its working method can timely monitor small leaks and prevent contamination of the main circuit coolant in the event of a large leak.

The invention provides a nuclear reactor primary loop coolant flow measurement and calibration device. A main pump main path comprises a first pipeline and a main pump arranged in the first pipeline; the main pump bypass comprises a second pipeline and a first flow measuring assembly arranged in the second pipeline, the second pipeline and the first pipeline are arranged in parallel, the first flow measuring assembly is used for measuring the second flow of the second pipeline, and the second flow is used for determining the first flow of the first pipeline; the calibration pump bypass comprises a third pipeline and a second flow measurement assembly arranged in the third pipeline, the third pipeline and the second pipeline are arranged in parallel, the second flow measurement assembly is used for measuring third flow of the fourth pipeline, and the third flow is used for calibrating the first flow; the calibration pump main path comprises a fourth pipeline and a calibration pump, and the fourth pipeline and the third pipeline are arranged in parallel. The independent closed-loop calibration loop with the calibration pump is arranged on the bypass of the main pump in parallel, and real-time and accurate measurement and calibration of the coolant flow of the primary loop are achieved.