52results 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 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.

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.

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 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 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.

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 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.

A heat exchanger for cooling a nuclear reactor core is disclosed herein. The heat exchanger can include a first stage including an input configured to receive a working fluid from an external source into the heat exchanger, and a first plenum configured to envelope a moderator heat pipe extending from the nuclear reactor core. The heat exchanger can further include a second stage including an output configured to remove a working fluid from the heat exchanger to the external source, and a second plenum configured to envelope a power heat pipe extending from the nuclear reactor core, wherein the first plenum and the second plenum are in fluid communication and configured such that the external fluid must traverse the first plenum and over the moderator heat pipe before entering the second plenum and traversing over the power heat pipe.

The invention particularly relates to a multi-steam-source multipurpose auxiliary steam system of a high-temperature gas cooled reactor, which belongs to the technical field of nuclear power supply of the high-temperature gas cooled reactor and comprises an auxiliary steam header, a plurality of auxiliary steam sources and a plurality of auxiliary steam users, the auxiliary steam distribution system is used for distributing auxiliary steam from one or more auxiliary steam sources to one or more auxiliary steam users in the starting, normal operation or shutdown period of the high-temperature gas cooled reactor unit, and qualified auxiliary steam is provided for the one or more auxiliary steam users. Stable steam supply of the system under the working conditions of starting, stopping, normal operation and accidents of the high-temperature gas cooled reactor unit is effectively guaranteed, meanwhile, the operation duration of the auxiliary boiler is remarkably shortened, and a brand new technical path is provided for efficient operation, energy conservation and emission reduction of nuclear power.

The invention relates to the field of nuclear reactors, in particular to a method for correcting the loop average temperature of a primary loop of a reactor. The method comprises the following steps: performing numerical simulation on typical working conditions according to actual operation parameters of a nuclear power plant to obtain fluid mechanics simulation data; on the basis of streamline distribution and flow distribution conditions in the fluid mechanics simulation data, dividing a reactor core into regions, enabling the regions to correspond to a hot section outlet, and determining a flow distribution coefficient according to the flow of a cold section inlet coolant flowing into the regions respectively; quantizing the inlet coolant flow, the initial temperature, the area temperature, the area inlet specific enthalpy and the power of each area based on the area division result and the flow distribution coefficient; performing heat balance calculation on each area to obtain the specific enthalpy increasing amount of the coolant flowing through each area and the coolant temperature at the outlet of each area; and averaging the coolant temperature at the inlet and the outlet of each area to obtain the average temperature of the coolant of each loop. The measurement accuracy is improved, and the calculation deviation of the primary loop heat balance power is improved.

This invention provides a heating system and control method for a high-temperature gas-cooled reactor (HTGR) nuclear power plant. The HTGR heating system includes: a HTGR nuclear island, a turbine generator set, a steam generator, a primary superheater, a secondary superheater, a heating mains, a user end, a feedwater preheater, a feedwater inlet, a deaerator, a feedwater heater, a feedwater pump, and a condenser. This invention fully utilizes the main steam from the HTGR nuclear island and the extracted steam from the turbine generator set as the heating steam source for the primary and secondary superheaters of the external heating system, producing externally supplied heating steam. Based on the operating status of the HTGR nuclear island, the heating system is controlled to adopt different operating modes, thereby achieving coordinated operation of power generation and heating. This provides high flexibility, high reliability, and good safety, and helps alleviate the pressure of carbon emissions from industrial heat consumption.

The invention discloses an elastic supporting device, a steam generator assembly and a nuclear reactor system, an elastic piece is arranged between a sealing pipe nozzle and a base plate, expansion difference generated by a steam generator from a cold state to a hot state is compensated, and the bottom of the device is in close contact with a shunting assembly under the action of external force such as vibration and seismic load by means of elastic force; and the steam generator is always in close contact with the shunting assembly. According to the main technical scheme, the elastic supporting device is used for the steam generator, a first through hole is formed in a base plate, the base plate is connected with an outlet connecting pipe, the outlet connecting pipe is communicated with the first through hole, and a reactor coolant flows out from the first through hole; the lower section of the outlet connecting pipe is inserted into an annular cavity of a sealing pipe nozzle below the outlet connecting pipe; a second through hole is formed in the other side, opposite to the outlet connecting pipe, of the sealing pipe nozzle; the elastic piece is arranged between the sealing pipe nozzle and the base plate and used for applying elastic force away from each other to the sealing pipe nozzle and the base plate. The support is mainly used for supporting.

The invention provides a thorium-based molten salt reactor system with fuel salt and proliferation salt isolated and an operation method of the thorium-based molten salt reactor system. The system comprises: a loop system comprising a reactor body, a fuel salt loop, a fuel salt circulating in the fuel salt loop, a proliferative salt loop, and a proliferative salt circulating in the proliferative salt loop, where the fuel salt comprises uranium-233 and does not comprise thorium-232, the proliferative salt comprises thorium-232 and does not comprise fissionable nuclear fuel, and the proliferative salt comprises uranium-233 and does not comprise thorium-232 and does not comprise fissionable nuclear fuel; the fuel salt loop and the proliferation salt loop are physically isolated in the reactor body and realize neutronics coupling; the reactor body comprises a main container and a plurality of fuel assemblies arranged in the main container, each fuel assembly comprises a fuel pipe made of an isolation material, and the two ends of each fuel pipe are in sealed connection with a lower cavity and an upper cavity in the main container respectively; the fuel pipe is wrapped with moderator blocks, and circulation channels are formed between the moderator blocks. Fuel salt of the fuel salt loop flows through the interior of the fuel pipe, and proliferative salt of the proliferative salt loop flows through a circulation channel between the moderator blocks.

A plurality of signal anomalies are identified in a number of tubes in a steam generator. Since the geometry of the steam generator is known, the location of each signal anomaly along each tube is converted into a location within the interior of the steam generator. If a plurality of signal anomalies are at locations within the steam generator that are within a predetermined proximity of one another, such a spatial confluence of signal anomalies is determined to correspond with a loose part situated within the steam generator. Additional methodologies can be employed to confirm the existence of the loose part. Historic tube sheet transition signal data can be retrieved and subtracted from present signals in order to enable the system to ignore the relatively strong eddy current sensor signal of a tube sheet which would mask the relatively weak signal from a loose part at the tube sheet transition.

The invention is suitable for the field of nuclear power station steam generator main water supply pump fault intervention, and provides a main water supply pump fault control method and system for a nuclear power station and terminal equipment, which are applied to a nuclear power station main water supply pump fault control system corresponding to a plurality of steam generators of the nuclear power station. The nuclear power station main water supply pump fault control system comprises at least one main water supply pump, and the method comprises the steps that when it is detected that at least one main water supply pump breaks down, state parameters corresponding to each steam generator are obtained; determining the fault type of at least one main water supply pump according to the state parameter corresponding to each steam generator; and performing fault intervention according to the fault type so as to eliminate the fault of at least one main water supply pump. According to the method, when the main water supply pump of the steam generator of the nuclear power station breaks down, intervention can be performed timely and accurately, and the loss of the nuclear power station can be reduced.

Steam generators in power plants exchange energy from a primary medium to a secondary medium for energy extraction. Steam generators include one or more primary conduits and one or more secondary conduits. The conduits do not intermix the mediums and may thus discriminate among different fluid sources and destinations. One conduit may boil feedwater while another reheats steam for use in lower and higher-pressure turbines, respectively. Valves and other selectors divert steam and / or water into the steam generator or to other turbines or the environment for load balancing and other operational characteristics. Conduits circulate around an interior perimeter of the steam generator immersed in the primary medium and may have different cross-sections, radii, and internal structures depending on contained. A water conduit may have less flow area and a tighter coil radius. A steam conduit may include a swirler and rivulet stopper to intermix water in any steam flow.

A method treats an internal surface of a metal part made from a nickel-based or cobalt-based alloy of a primary circuit during an operating cycle of a PWR reactor. The method includes estimating a quantity of nickel discharged into the primary circuit by the metal parts of the primary circuit in contact with water of the primary circuit, during an operating cycle of the reactor; calculating a quantity of magnetite necessary for fixing the whole of the quantity of nickel estimated; forming a treated water by dissolving the quantity of magnetite calculated in the water of the primary circuit when the water has a temperature of between 80° C. and 180° C.; and putting this treated water in circulation in the primary circuit so as to form, on the internal surface of at least one metal part, a layer of a protective oxide to fix the quantity of nickel estimated.

The present application relates to nuclear power technology field, specifically to a kind of high temperature gas cooled reactor steam generator feedwater treatment method, comprising: before starting, using first reducing full volatile treatment mode to treat feedwater;During starting, using second reducing full volatile treatment mode to treat feedwater;When running and after completing two-phase flow transition in steam generator secondary side, using oxidizing full volatile treatment mode to treat feedwater;When stable operation, using oxygen treatment mode to treat feedwater;When shutdown and before entering two-phase flow in steam generator secondary side, using second reducing full volatile treatment mode to treat feedwater;During shutdown stage and after completing two-phase flow transition to become liquid phase in steam generator secondary side, using third reducing full volatile treatment mode to treat feedwater;When unscheduled shutdown, adding hydrazine and ammonia water in feedwater;The present application selects appropriate feedwater treatment mode by different state of high temperature gas cooled reactor, reduces corrosion product generation, prevents and slows down deposition, ensures safe and stable operation.

A nuclear power system has an open volume between containment vessel and a reactor vessel containing a reactor core. Located in the open volume is a container holding a neutron-absorbing chemical in solid form. The container is configured to release the chemical in solid form directly into the open volume in response to a predetermined temperature and / or a predetermined pressure within the open volume. The released chemical can assist in maintaining the reactor core in a sub-critical state.

The invention provides an asymmetric composite runner plate heat exchanger, and belongs to the field of heat exchange equipment. The problems that in a fourth-generation lead-cooled fast reactor, when a hot working medium exchanges heat with a cold working medium, the pressure drop is too large, a micro-channel is prone to being blocked, and the heat exchange efficiency is low are solved. Comprising a plurality of heat exchange units which are arranged in a stacked mode, a shell is arranged outside the heat exchange units which are arranged in a stacked mode, each heat exchange unit comprises an upper cold plate, a lower cold plate and a hot plate which are arranged from top to bottom, the upper cold plates and the lower cold plates are arranged in a mirror symmetry mode, and arc-shaped channels are correspondingly formed in the bottom faces of the upper cold plates and the top faces of the lower cold plates. After the upper cold plate and the lower cold plate are connected, the arc-shaped channels are in butt joint to form a cold working medium channel with an oval section, the cold working medium channel is of a snakelike structure, the hot plate is provided with a plurality of parallel straight channels to form a hot working medium channel, and the section size of the hot working medium channel is larger than that of the cold working medium channel. And the flowing directions of the cold working medium channel and the hot working medium channel are crossed.

A steam generator for a nuclear power plant is provided. The steam generator has an upright, elongated body and a tension member that extends from a proximal end, which is mounted at each mounting position on the elongated body, to a distal end, which is configured to be mounted at each load-bearing mounting position within the nuclear power plant. Each proximal end is lower than its respective distal end. When in use, the tension member receives a tensile load from the steam generator and transmits substantially all of the weight of the steam generator to the load-bearing mounting positions within the nuclear power plant.

Disclosed is an apparatus and method for conditioning fluid flow in a nuclear power plant steam generator. A flow conditioning device includes an outer enclosure defining a plurality of entrance apertures arranged in an array and a plurality of exit apertures arranged in an array. A plurality of baffle plates are defined within the outer housing. The baffle plates define flow channels in fluid communication with the entrance and exit apertures to create a flow path of alternating directions. The flow channels receive fluid flow from the plurality of entrance apertures, direct the fluid flow from the entrance apertures in alternating directions through the flow channels to impart turning and frictional pressure loss to the fluid flow, and direct exiting fluid flow through the exit apertures into the tubelane region of the steam generator.

UNIVERSAL PERFORMANCE TEST METHOD FOR NUCLEAR STEAM SUPPLY SYSTEM This disclosure specifically relates to a universal performance test method for a nuclear steam supply system, which includes: determining a limit of the nuclear steam supply system, and arranging performance test measurement points of a nuclear power generating unit and the nuclear steam supply system; determining a power generated by the nuclear power generating unit under the TMCR condition; performing a performance test on the nuclear steam supply system under a design condition based on the power generated by the nuclear power generating unit under the TMCR condition, and correcting a steam flow rate exported from the nuclear steam supply system subsequent to the design condition;and the correction of the power generated by the nuclear power generation unit under the TMCR condition, and the calculation of a test result for the nuclear steam feed system. This disclosure objectively and accurately reflects an operating state of the nuclear steam feed system and provides reliable support for design verification, performance evaluation, and subsequent optimization design of the steam feed system. Summary figure: Fig. 1;