1425results about "Cooling arrangement" 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 relates to a containment pressure suppression, filtration and discharge system and method and application thereof, and belongs to the field of special safety facilities and radiation protection of nuclear power devices. The containment pressure suppression, filtration and discharge system comprises a discharge valve (2), a discharge pipeline (3), a spray pipe (4), a pressure suppression pool, an expansion chamber (7), a steam-water separator (9), a metal fiber filter layer (10), a rupture disk (12), a heat exchanger and a valve. According to the containment pressure suppression filtering and discharging system, the seawater side heat exchanger and the pressure suppression pool heat exchanger are arranged, and a natural circulation loop is constructed by means of the potential difference between the seawater side heat exchanger and the pressure suppression pool heat exchanger and the temperature difference between seawater and a liquid phase in a pressure suppression pool under the accident condition; the heat of the suppression pool can be guided out without time limit under the condition of an LOCA accident or a main steam pipeline fracture accident in the containment; by adopting a pressure suppression washing method with multi-stage bubbling, rapid pressure suppression of the containment and efficient retention of airborne radioactive substances can be realized; occupation of total resources is reduced by arranging the combined metal fiber filter layer.

The invention provides a deep learning method for predicting thermal hydraulic characteristics of a heat exchanger of a sodium-cooled fast reactor, and belongs to the technical field of passive residual heat removal systems of sodium-cooled fast reactors. The method comprises the following steps of: approximating low-fidelity flow and temperature distribution by using a first DNN; constructing a nonlinear relationship between the low-fidelity data and the high-fidelity data through the second DNN; the physical information partial differential equation and the boundary condition are used as constraints to be injected into the third DNN, and a prediction result is optimized; according to the method, the fluid heat transfer partial differential equation and the data driving advantages are combined, high-fidelity prediction of the flow velocity and the temperature in the sodium-cooled fast reactor PCHE can be achieved only through low-fidelity data, the problem that physical information in a heat exchanger is difficult to measure is solved, and the method has great significance in improving the inherent safety of the sodium-cooled fast reactor and the safety of a reactor core.

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.

The invention relates to the technical field of nuclear reactor safety, and particularly discloses a reactor waste heat export, recovery and comprehensive utilization system and method based on a heat pipe, and the system comprises a nuclear energy generation system, a secondary loop energy conversion system, a primary loop waste heat discharge system, a secondary loop waste heat discharge system, and a waste heat recovery and comprehensive utilization system. According to the waste heat recycling system, the heat pipe can be used for heat transfer and storage under the peak regulation operation condition, the nuclear heat use efficiency is remarkably improved, the requirements of various types of energy sources such as electricity and heat are met, and passive export of reactor core waste heat can be achieved through multi-natural-circulation coupling cooperative operation under the condition of a whole-field outage accident. By means of the efficient heat transfer characteristic and the modularization characteristic of the heat pipes, independent operation of the heat pipes can be achieved, the problem of single-point failure of the system can be avoided, meanwhile, separation of a reactor coolant and a cooling medium of a waste heat removal system can be achieved, the inherent safety of a nuclear reactor is remarkably improved, and the requirements of multi-purpose scenes are met; and the purpose of fully utilizing resources is achieved.

The invention discloses a pressurized water reactor primary loop boron concentration prediction and control method and system, and the method comprises the steps: obtaining an original helium flow sequence of a pressurized water reactor primary loop, and obtaining M boron concentration IMF components IMF1,..., IMFM; the M boron concentration IMF components IMF1,..., IMFM are put into a trained LSTM model for prediction, predicted values Y1, Y2,..., YM of the IMF components are obtained, and error coefficients delta Y1, delta Y2,..., delta YM corresponding to the boron concentration IMF components are calculated according to the predicted values Y1, Y2,..., YM of the IMF components; and inputting error coefficients delta Y1, delta Y2,..., delta YM corresponding to each boron concentration IMF component into the trained DenseNet prediction model to obtain an optimal IMF component predicted value # imgabs0 # of the boron concentration signal sequence, based on a DEM-DAC algorithm, controlling the boron concentration of the pressurized water reactor primary loop according to the optimal IMF component predicted value # imgabs1 # of the boron concentration signal sequence, and calculating the boron concentration of the pressurized water reactor primary loop according to the optimal IMF component predicted value # imgabs0 # of the boron concentration signal sequence. The method and the system can solve the technical problems that the boron concentration fluctuation of the primary loop is large and the unit load regulation performance is poor during the reactor variable load operation period.

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

Embodiments of the present disclosure include an integral nuclear reactor having a core fluidly coupled to an inlet chamber and to an outlet chamber. In some instances, the nuclear reactor may include a hub to house the core, the chambers, and the protective plug. The hub may include a window allowing a heat transfer fluid to flow from the outlet chamber, through the hub, and enter an annular space between the hub and a separation shell. The heat transfer fluid may flow into an inlet of a heat exchanger. The heat transfer fluid may flow through the heat exchanger before exiting at a heat exchanger outlet. The heat transfer fluid may enter an annular delay tank before flowing down an annular downcomer duct formed by the separation shell being housed in a reactor vessel. The heat transfer fluid may flow from the annular downcomer duct and into the inlet chamber.

The invention relates to the technical field of lead-based stacks, in particular to a nuclear storage integrated lead-based stack with a load autonomous following function. The nuclear storage integrated lead-based reactor with the automatic load following function comprises a reactor core, a phase change energy storage device and a heat energy utilization device, the reactor core is used for heating a coolant; the heat energy utilization device is used for absorbing heat in the coolant; the phase change energy storage device is arranged on the inlet side of the reactor core, the phase change energy storage device is used for carrying out heat interaction with a coolant, and the phase change temperature of the phase change energy storage device is consistent with the preset inlet temperature of the reactor core. According to the nuclear storage integrated lead-based reactor with the automatic load following function, natural circulation flow is not prone to oscillation and diffusion, the risk of overheating melting of the fuel assembly is not prone to occurring, thermal fatigue of the structural assembly is not prone to occurring, and high safety performance is achieved.

A mixed-spectrum molten-salt reactor has a primary containment vessel. A reactor core is housed in the primary containment vessel, the reactor core has an atypical geometry that holds a fuel salt. A blanket region is formed within the primary containment vessel and surrounding the reactor core, the blanket region comprises a blanket salt.

The invention belongs to the technical field of fluid machinery, and particularly relates to a novel tank type molten salt reactor main pump structure. Comprising a driving device, a sealing section, an air cavity section and a working section which are sequentially connected from top to bottom. The driving device is connected with the working section through a rotating rod; the rotating rod is rotatably connected with the sealing section and the air cavity section; the air cavity section comprises a third outer shell, an air outlet pipe and an air inlet pipe penetrate through the side wall of the third outer shell, an air outlet control unit is arranged at the end, close to the third outer shell, of the air outlet pipe, an air inlet control unit is arranged at the end, close to the third outer shell, of the air inlet pipe, and the air outlet control unit and the air inlet control unit are connected with the working section. The gas outlet control unit and the gas inlet control unit perform negative feedback regulation on the molten salt liquid level of the third outer shell, so that the molten salt liquid level is prevented from being in direct contact with the sealing section; the working section comprises an impeller, a guide vane, a fused salt inlet and a fused salt outlet, fused salt enters the third outer shell from the fused salt inlet, then flows into the impeller and the guide vane section through the pump inlet to be pressurized and stabilized, and finally flows out through the fused salt outlet.

The embodiment of the invention provides a pressure control method, electronic equipment, a pressure stabilization control system and a storage medium, and relates to the technical field of nuclear reactors. The method comprises the following steps: acquiring a pressure measurement value and a pressure setting value of the pressure stabilizer; determining a pressure deviation value according to the pressure measurement value and the pressure setting value; if the absolute value of the pressure deviation value is smaller than or equal to a first preset value, the charging and discharging flow of a loop is adjusted according to the pressure deviation value and the pressure measurement value so as to control the pressure of the pressure stabilizer; if the absolute value of the pressure deviation value is larger than the first preset value, the pressure stabilizer is controlled to be inflated or deflated according to the pressure deviation value so as to control the pressure of the pressure stabilizer. According to the embodiment of the invention, the voltage stabilization effect of the voltage stabilization control system can be improved.

The invention relates to the field of nuclear reactor engineering design, and discloses a molten salt reactor loop system and a molten salt reactor, the loop system comprises a molten salt heat exchanger and a molten salt pump; the molten salt heat exchanger is provided with a supporting flange; the molten salt pump is provided with a pump tank flange; the molten salt reactor loop system further comprises a coupling flange, and the coupling flange is connected to the lower portion of the supporting flange and the lower portion of the pump tank flange, so that the molten salt pump is located on the peripheral side of the molten salt heat exchanger. According to the molten salt reactor loop system, the molten salt heat exchanger and the molten salt pump are connected into a whole which is arranged left and right through the connecting flange, so that the structure of the molten salt reactor loop system is compact, the molten salt pump is directly arranged on the peripheral side of the molten salt heat exchanger, and the molten salt pump and the molten salt heat exchanger can be communicated through a short pipe; and the complexity of pipeline arrangement of the whole system is reduced.

The embodiment of the invention provides a safety injection system and method, and relates to the field of reactor safety systems. The problem that the water injection flow of a safety injection system is not matched with the flow needed in the reactor water loss accident stage is solved. The safety injection device comprises a pressure-bearing container and a damping device, wherein a safety injection outlet pipe is arranged at the bottom of the pressure-bearing container; the damping device is arranged in the pressure-bearing container, the damping device is provided with a water inlet and a water outlet, and safety injection water in the pressure-bearing container flows through the damping device from the water inlet and then flows to the safety injection outlet pipe from the water outlet. The drainage flow of the safety injection water in the pressure-bearing container is gradually reduced along with the process of the water loss accident, and compared with a pressure-bearing container without the damping, the flow in the initial stage is lower, so that the water quantity of the safety injection bypass can be reduced; the flow in the middle stage can quickly realize refilling of water in the lower cavity of the reactor core and resubmerging of the reactor core; the flow submerged in the reactor core is maintained to be lower in the later period of the accident, the safety injection time is fully prolonged, and safety injection water is effectively utilized.

The invention provides a reactor core structure capable of controlling an assembly clearance and a manufacturing method thereof. The reactor core structure comprises a reactor core heat management element and a filling material; wherein the base body is provided with a plurality of assembly holes; the reactor core heat management element comprises a fuel rod or a heat pipe; the reactor core heat management element is assembled in the assembly hole of the base body; the filling material is filled in an assembly gap between the reactor core heat management element and the assembly hole; and the filling material is a shape memory alloy. The reactor core structure can improve the safety margin of the heat pipe reactor.

The embodiment of the invention provides an integrated small reactor passive residual heat removal test heat exchange structure, system and method, and relates to the technical field of nuclear reactor safety system tests. The system aims to solve the problems that in an integrated small reactor passive quay crane test system, split type and pipeline connection are adopted for containment cooling, so that the system is tedious and large in occupied space. And the heat exchange structure is directly connected with the containment simulation device and the cooling water tank through the shared heat exchange part. The test system comprises the test heat exchange structure, and the test method is implemented by adopting the test system. And heat is transferred from the high-temperature containment simulation device to the low-temperature cooling water tank fluid through the heat exchange part. Redundant connecting parts of the common heat exchange part are omitted, and occupied space is reduced. Efficient and rapid heat transfer is achieved, waste heat is discharged in time, safety is guaranteed, and due to the compact structural design, the whole test device is more reasonable in layout, saves space and is convenient to install, debug and maintain.

The invention belongs to the technical field of nuclear power, and particularly relates to a nuclear power generation steam supply system and a using method thereof. The nuclear energy power generation steam supply system comprises a sodium-cooled fast reactor, an intermediate heat exchanger, a secondary loop sodium pump, a sodium-fused salt heat exchanger, a low-temperature fused salt conveying pump, a first adjusting valve, a low-temperature fused salt tank, a high-temperature fused salt tank, a second adjusting valve, a high-temperature fused salt conveying pump, a fused salt-water heat exchanger, a steam loop control valve, a power generation loop control valve, a superheater and a main water feeding pump. The invention discloses a steam turbine generator, a check valve and a condenser. By adopting a sodium-sodium-fused salt-water four-loop design, the problems that the possibility of sodium-water reaction is not eliminated and peak regulation and frequency modulation capabilities are not achieved due to the fact that an existing fast reactor adopts a sodium-sodium-water three-loop design are solved.

The invention discloses a system and a method for testing thermal cycle characteristics of multidirectional horizontal pipelines of a nuclear power plant, the system comprises a circulating water path and a pressure stabilizing water path, the circulating water path comprises the multidirectional horizontal pipelines (1a, 1b), paired pressure signal detectors (17a, 17b), pipeline heaters (14a, 14b, 14c) and a pipeline heat exchanger (3), and the pressure stabilizing water path comprises a cooling water supply system (18) and a pressure stabilizer (23); the circulating water path is communicated with the pressure stabilizing water path, and the number of the multi-direction horizontal pipelines (1a and 1b) is more than two. The thermal cycle characteristics of the upstream horizontal pipeline, the lateral horizontal pipeline and the downstream horizontal pipeline connected with the main pipeline under the conditions of different temperatures, pressures and volume flows can be researched more effectively, conveniently and accurately, and data support and reference are provided for design optimization of the multi-direction horizontal pipeline of the nuclear power station.

A method includes: during a first time period, pumping propellant: to a set of moderator coolant channels to cool the moderator; through a set of fuel elements to heat the propellant to a first outlet temperature; out of a thrust nozzle to produce a first thrust; and locating the reflector at a first closed position to reflect neutrons toward the set of fuel elements to maintain a fission rate of the set of fuel elements above a threshold fission rate. Additionally, during a second time period: locating the reflector at a second open position to reduce the fission rate of the set of fuel elements to below the threshold fission rate; and pumping the propellant through the fuel elements to heat the propellant to a second outlet temperature less than the first outlet temperature and out of the thrust nozzle to produce a second thrust less than the first thrust.

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).

An installation includes: a plurality of pilings securable to a bed under a surface of a body of water; a base structure disposed atop the plurality of pilings; and a module disposable on the base structure, wherein the module is positioned and securable on the base structure after being floated on the surface of the body of water over the base structure.

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 provides a long-term passive cooling system of a compact small reactor, which is characterized in that a reactor and steam generators are arranged in a containment, the steam generators are connected with the reactor, and each steam generator is provided with a secondary side residual heat removal system; the liquid storage chamber is mounted on the outer wall surface of the containment in a surrounding manner; the passive heat exchanger is arranged in the liquid storage chamber and is connected with a secondary side residual heat removal system of the corresponding steam generator, so that steam of the steam generator enters the passive heat exchanger for heat exchange; the plurality of passive heat pipes are mounted on the wall surface of the containment in a penetrating manner and are used for guiding heat in the containment to the liquid storage chamber. The secondary side waste heat removal system is arranged on the secondary side of the steam generator and used for relieving non-loss-of-water accidents, short-term and long-term reactor core cooling is achieved outside the containment through the passive heat exchanger combining water cooling and air cooling, the peak pressure of the containment after the accidents is effectively reduced, long-term passive heat carrying is achieved, and the unit operation safety is improved.

The invention discloses a direct injection safety system for a reactor pressure vessel. The direct injection safety system comprises a first water injection pipeline and / or a second water injection pipeline, the first water injection pipeline penetrates through an upper seal head of the reactor pressure vessel, the input end of the first water injection pipeline is located outside the reactor pressure vessel and connected to a coolant source, and the output end of the first water injection pipeline is located in the reactor pressure vessel and penetrates through the reactor core upper supporting plate to extend to a reactor core area; the output end of the first water injection pipeline can be directly positioned to a reactor core area to output a coolant, so that the reactor core cooling efficiency is improved; the second water injection pipeline penetrates through the reactor pressure vessel, the input end of the second water injection pipeline is located outside the reactor pressure vessel and connected to the coolant source, the output end of the second water injection pipeline is located in the reactor pressure vessel and at least extends to the descending section of the reactor pressure vessel, and the output end of the second water injection pipeline can directly output coolant from the descending section. And more coolants are promoted to flow to the reactor core area from the reactor core lower support plate, so that the function of cooling the reactor core is achieved.

A nuclear reactor is configured with a primary coolant loop for transferring heat away from the nuclear reactor core. In a shutdown event, the primary coolant pump may stop pumping primary coolant through the reactor core, resulting in decay heat buildup within the reactor core. An inertial energy coast down system can store kinetic energy while the nuclear reactor is operating and then release the stored kinetic energy to cause the primary coolant to continue to flow through the nuclear reactor core to remove decay heat. The inertial energy coast down system may include an impeller and a flywheel having a mass. During normal reactor operation, the flowing primary coolant spins up the impeller and flywheel, and upon a shutdown event where the primary coolant pump stops pumping, the flywheel and impeller can cause the primary coolant to continue to flow during a coast down of the flywheel and impeller.

The invention relates to the technical field of nuclear reactor safety, in particular to a heat exchange device, a waste heat removal system and a use method. A heat exchange device comprises a first cooling water tank, a second cooling water tank and a heat exchanger, a heat exchanger is suitable for being arranged in the first cooling water tank, the first cooling water tank is suitable for being connected with a containment, and cooling liquid flows in the first cooling water tank; the second cooling water tank is suitable for being connected with the containment vessel, and cooling liquid flows in the second cooling water tank; and the first cooling water tank and the second cooling water tank are communicated through the pipeline assembly. The invention provides a heat exchange device, a waste heat removal system and a using method, and aims to solve the problems that a cooling water tank of an existing waste heat removal system is usually designed to be a tank body, so that the flowability of cooling liquid in partial space in the cooling water tank is insufficient, and the cooling liquid does not participate in heat exchange.

This invention discloses a firecracker-style controlled nuclear fusion and fission synergistic power generation device and method. The power generation device includes a fission-fusion reaction system, a circulating cooling system, and other accessories. The fission-fusion reaction system includes a neutron gun, fuel spheres, a fuel sphere storage and delivery device, and a vacuum reactor. The circulating cooling system includes a circulating fan, a voltage regulator, a steam turbine, a steam generator, delivery pipelines, and a water pump. Other accessories include an energy storage and control system and a vacuum pump. This invention innovatively integrates the start-up mechanism, drawing on the firecracker-style multi-reactor relay concept, combining nuclear fission and nuclear fusion. It utilizes the high temperature and pressure generated by nuclear fission to start nuclear fusion, reducing start-up difficulty and cost. Simultaneously, it efficiently converts and utilizes energy, employing helium for heat transfer. Helium's high thermal conductivity, low density, high specific heat capacity, and chemical inertness allow it to effectively carry away the heat stored in the phase change microspheres, comprehensively collecting and converting nuclear reaction heat energy, thus improving energy conversion efficiency.

The invention discloses a nuclear power plant water system and a cooling method.The nuclear power plant water system comprises a plurality of series and non-safety-level pools, each series is provided with a safety-level pool, the safety-level pools are communicated pairwise, and each series has a first working condition and a second working condition; the sum of the water quantity borne by any two safety-level pools can be at least used for the system to operate for 3 days under the second working condition. The non-safety-level water pool can at least store water for the system to operate for 27 days under the second working condition. Water resources stored in the non-safety-level water pool and the safety-level water pool can be flexibly allocated to different series according to actual needs, idling and waste of the water resources are avoided, the requirement for the total water storage amount is effectively reduced, and therefore the building and maintaining cost of the water pool is reduced. Reduction of the total water storage capacity of the nuclear power plant means that occupation and consumption of land and water resources in the pool building process are reduced.

A small-scale nuclear reactor cooling device and a cooling method are provided. The small-scale nuclear reactor cooling device includes a reactor vessel in which cooling water circulates along a flow path structure, and a cooling water storage unit that is provided inside the reactor vessel and is filled with the cooling water based on the circulating flow, and when a molten core occurs in the reactor vessel due to the occurrence of a serious accident, the cooling water storage unit comes into contact with the molten core via the cooling water filled inside to cool the molten core and prevent the molten core from leaking out to the outside.

The invention discloses a dual-mode self-starting residual heat removal system for a heat pipe reactor, which belongs to the technical field of nuclear reactor safety, and comprises a heat pipe heat exchanger for dissipating heat of a plurality of heat pipes; the air cooling system comprises an air channel formed in the heat pipe heat exchanger, the heat pipe is located in the air channel, circulation openings are formed in the top and the bottom of the air channel respectively, and when the heat pipe heat exchanger fails, the circulation openings are opened to communicate with the external atmosphere. The waste heat removal heat exchanger is provided with an inner-layer cavity and an outer-layer cavity, the inner-layer cavity and the outer-layer cavity are both filled with molten metal used for heat conduction, the heat pipe penetrates through the inner-layer cavity, and a self-opening assembly is arranged between the inner-layer cavity and the outer-layer cavity. And the self-opening assembly is automatically opened. According to the invention, self-starting dual-mode residual heat removal can be realized under extreme conditions, and the residual heat removal capability and safety margin of the heat pipe reactor system under accident and shutdown conditions are improved.