155results about "Steam generator primary side" patented technology

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 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 discloses a passive nuclear steam supply system which is characterized in that two ends of a hot-section main pipeline are respectively connected with a reactor pressure vessel and a steam generator, a main pump is vertically connected with the steam generator, two ends of a cold-section main pipeline are respectively connected with the pressure vessel and the main pump, the hot-section main pipeline is provided with a first liquid level pressure tap and a second liquid level pressure tap, and the cold-section main pipeline is provided with a second liquid level pressure tap. The first liquid level pressure tap is located at the bottom of the hot section main pipeline, the second liquid level pressure tap is close to the steam generator and located at the top of the hot section main pipeline, the reactor coolant system further comprises a pressure stabilizer and a surge pipe, one end of the surge pipe is connected to one of the hot section main pipelines, and the other end of the surge pipe is connected to the bottom of the pressure stabilizer. The passive nuclear steam supply system can allow the rated power of the reactor core to be remarkably improved.

A passive nuclear steam supply system (100), comprising a reactor body (1), the reactor body (1) comprising a reactor pressure vessel (11); and a reactor coolant system (2), the reactor coolant system (2) comprising two coolant loops (21), a pressurizer (22) and a surge line (23), and each coolant loop (21) comprising a steam generator (24), two primary pumps (25), a hot leg main pipe (26) and two cold leg main pipes (27), wherein the hot leg main pipe (26) is provided with a first liquid level pressure tap (261) and a second liquid level pressure tap (262), the first liquid level pressure tap (261) being located at the bottom of the hot leg main pipe (26), and the second liquid level pressure tap (262) approaching the steam generator (262) and being located at the top of the hot leg main pipe (26).

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.

A nuclear reactor includes a heat exchanger that transfers thermal energy from a primary reactor coolant to a secondary coolant. The heat exchanger is formed with a hot flow channel, a cold flow channel, and a porous layer between the hot flow channel and the cold flow channel. The porous layer may be thermally insulative to reduce the efficiency of thermal energy transfer from the hot flow channel to the cold flow channel. The porous layer may have a control gas passed therethrough that can be tailored to control the thermal energy transfer through the porous layer. The control gas can be tested for leakage within the heat exchanger. The control gas may also be used to sequester fission or activation products.

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.

The invention particularly relates to a large sodium-cooled fast reactor power station, which belongs to the technical field of sodium-cooled fast reactors, and comprises a plurality of sodium-cooled fast reactor units and an auxiliary system shared by the sodium-cooled fast reactor units, and each sodium-cooled fast reactor unit comprises a sodium-cooled fast reactor, a main heat transmission system, a power conversion system, an accident waste heat removal system and an energy storage system; the sodium-cooled fast reactor is arranged in a reactor building; the sodium-cooled fast reactor comprises a reactor pit, a containment vessel in the reactor pit and a reactor body in the containment vessel, and is constructed through the following steps that the reactor pit, the containment vessel and the reactor body are constructed in parallel, the reactor body is loaded into the containment vessel after the containment vessel and the reactor body are constructed, the reactor body is modularly manufactured, the containment vessel is a capsule type small steel containment vessel, and the containment vessel is a capsule type small steel containment vessel. The pile pit is built by concrete; and integrally pushing the containment and the reactor body in the containment into a reactor pit. While the nuclear safety and the industrial safety are improved, the economical efficiency of the sodium-cooled fast reactor power station is greatly improved, and the construction period of the sodium-cooled fast reactor power station is greatly shortened.

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 invention discloses a nuclear storage and supply integrated method and system based on a pressurized water reactor and fused salt heat storage. The method comprises the following steps that steam is provided for a steam turbine device through nuclear energy of a reactor device; power is generated through the steam turbine device; heat is provided for the fused salt heat storage device through the steam turbine device, and heat is stored and provided through fused salt in the fused salt heat storage device. A nuclear power'machine '(steam turbine device) side is coupled with the fused salt heat storage device to form a nuclear storage and supply integrated system, that is, flexible connection is added between nuclear power and a user, so that decoupling of a nuclear power reactor device and the steam turbine device is realized, a flexible nuclear energy concept is realized, and the system can be applied to peak regulation, large-scale heating and steam supply scenes; the requirements for nuclear power peak regulation and nuclear energy comprehensive utilization are met through the fused salt heat storage device, the reactor power can not be adjusted frequently, potential safety hazards caused by frequent reactor power adjustment are avoided, and the nuclear power safety is improved.

An apparatus and method for conditioning fluid flow within a steam generator of a nuclear power plant is disclosed. The flow conditioning apparatus includes an outer enclosure defining a plurality of inlet openings arranged in an array and a plurality of outlet openings 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 inlet openings and the outlet openings to form alternating directional flow paths. The flow channels receive fluid flow from the plurality of inlet openings, direct the fluid flow from the inlet openings through the flow channels in alternating directions, impart swirl and frictional pressure losses to the fluid flow, and direct the exiting fluid flow through the outlet openings to a tube lane region of the steam generator.

The invention provides a primary loop safety control system and control method.The control system comprises a pressure source, a receiving unit, a switch unit and a balance unit, one end of the pressure source is connected with a primary loop through the balance unit, the pressure source is further connected with the receiving unit and the switch unit, and one end of the switch unit is connected with the primary loop; the other end is connected to the receiving unit; the balance unit comprises a shell, a first channel and a second channel are arranged on the shell, a control mechanism is arranged between the first channel and the second channel, the first channel is connected with a pressure source, the second channel is connected with a loop, and the control mechanism is used for controlling connection and disconnection between the first channel and the second channel. Therefore, the functions of overpressure protection from the pressure source to the primary loop side, automatic pressure supplement from the primary loop to the pressure source side and pressure maintaining of the pressure source are achieved. According to the primary loop safety control system, the structural configuration of the primary loop side control system is simplified, and the spatial arrangement requirement of a small reactor is met.

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 present invention belongs to the technical field of nuclear power plant equipment maintenance, and specifically relates to an exhaust device of a nuclear island drain and exhaust system of a nuclear power plant, which is achieved by connecting a set of exhaust devices in parallel on the exhaust pipeline at the top of the nuclear island pressurizer equipment, and by adding two tees and isolation valves, vacuum pumps (vacuum ejectors) and other devices on the original pipeline. The vacuum ejector is powered by compressed air, and the exhaust gas mixed with the gas of a circuit returns to the nuclear island drain and exhaust system. The device of the present invention for the flushing stage of a circuit after the overhaul of a nuclear power plant is equipped with a set of exhaust devices of the nuclear island drain and exhaust system, and a vacuum ejector is added to the exhaust pipeline, and the gas of the circuit is discharged through the vacuum ejector, and the circuit is filled with water. The vacuum ejector is powered by maintenance compressed air, and the exhaust gas mixed with the gas of the circuit returns to the nuclear island drain and exhaust system. When the pressurizer liquid level reaches full water, the pressurizer exhaust valve and the isolation valve of the vacuum ejector are closed.

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 power generation system and related method for repowering a fossil-fueled power plant using a carbon-free nuclear steam supply system (NSSS) which replaces the existing fossil plant steam generator which burns fossil fuel such as coal, oil, or natural gas. The existing fossil plant energy conversion system including the turbine-generator (turbogenerator) and auxiliary components of the Rankine cycle is retained. The NSSS may include a small modular reactor (SMR) unit comprising a reactor vessel with nuclear fuel core and steam generator which receives heated primary coolant from the reactor to produce main steam to operate the Rankine cycle. The main steam output by the SMR unit is compressed in a steam compressor to increase its pressure to a level necessary to operate the turbogenerator. The compressor may be operated via a portion of the main steam. An intercooler of the compressor may be used for main steam reheating.

The invention discloses a hydrogenation system and a hydrogenation method for a high-temperature and high-pressure irradiation test loop. The hydrogenation system comprises a hydrogen supply unit, a flow control unit and a feeding unit which are connected in sequence; the flow control unit comprises an automatic adding unit and a manual adding unit which are arranged in parallel; the manual adding unit comprises a return water providing unit and at least two hydrogenation units, and the hydrogenation unit is connected to the feeding unit and the hydrogen supply unit; at least two hydrogenation units are arranged in parallel in at least two rows; each row of hydrogenation units is connected to the return water providing unit; each row of hydrogenation units is provided with a temperature measuring instrument, and a pressure measuring instrument is provided on a pipeline connecting the hydrogenation unit and the feeding unit; isolation valves are provided at both ends of each row of hydrogenation units; the operation is convenient, safe and reliable, the dissolved hydrogen concentration in the loop has a fast response speed, a high control accuracy, and a high hydrogen use efficiency, which meets the hydrogenation requirements in the high-temperature and high-pressure irradiation test loop.

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.

Certain aspects of the present disclosure are generally directed to tritium shunt heat exchangers that use a sweep gas. In some aspects, a heat exchanger system for a fusion power plant is disclosed herein. The system may advantageously allow for efficient energy and tritium extraction from a tritium -containing fluid, while minimizing tritium leakage into the environment. For example, the system may comprise components, such as a thermally conductive solid connector, a sweep gas, reactive materials, etc., that, allow for high heat transfer efficiency, and / or high tritium removal and extraction efficiency. In addition, some aspects of the disclosure are directed to methods for using or making such a system.

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.

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.

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.

A pressurized water reactor nuclear energy steam supply system (101), comprising: a pressurized water reactor nuclear power unit (11), which is connected to a first thermal treatment loop and outputs, to the first thermal treatment loop, steam to be treated; a steam generator (12), comprising a pipe side path and a shell side path that are isolated from each other, wherein the pipe side path is located in the first thermal treatment loop, the shell side path is located in a second thermal treatment loop independent of the first thermal treatment loop, and the steam generator (12) can use said steam the first thermal treatment loop as a heat source input to generate clean steam in the second thermal treatment loop; and a steam compressor device (13), which is connected to the second thermal treatment loop, wherein the steam compressor device (13) can treat the clean steam in the second thermal treatment loop to increase the pressure and / or temperature of the clean steam. The solution can produce medium-pressure superheated steam or high-pressure superheated steam, so as to satisfy the industrial steam parameter requirements of a chemical industry base, thereby achieving the purpose of comprehensively utilizing the nuclear energy of the pressurized water reactor nuclear power unit.