44 results about "Spent fuel pool" patented technology

The invention discloses a nuclear power plant fuel operation accident source item result determination method and device and a nuclear power plant fuel operation accident alarm method and system, and relates to the technical field of nuclear industry. The method for determining the fuel operation accident source item result of the nuclear power plant comprises the following steps: acquiring a target decontamination factor corresponding to a target nuclide in a first time period and the total concentration of aerosol iodine retained in a spent pool of the nuclear power plant in a second time period; determining a first release amount corresponding to the target nuclide in the first time period according to the target decontamination factor; according to the total concentration of iodine in the aerosol, determining the share of iodine volatilized in the form of elements in the spent fuel pool in the second time period; determining a second release amount of the element iodine in a second time period according to the iodine share; and generating a fuel operation accident source item result of the nuclear power plant according to the first release amount and the second release amount. According to the embodiment of the invention, the method can improve the accuracy of determining the fuel operation accident source item result of the nuclear power plant.

The invention provides a corrosion product centrifugal separation device and method for a spent fuel pool, and belongs to the technical field of spent fuel pool purification systems.The corrosion product centrifugal separation device comprises a solid collecting tank and a Venturi tube, a pore plate is installed in a cylindrical section of the upper portion of the solid collecting tank, and the bottom of the solid collecting tank is connected with an inlet pipe; a cyclone impeller is mounted in the inlet pipe, a separation section is arranged at the top of the inlet pipe, a cylindrical guide plate is externally connected to the separation section, the separation section is connected with a connecting pipe, a recovery impeller is mounted in the connecting pipe, the connecting pipe is connected with a Venturi pipe, and the Venturi pipe is connected with an outlet. Based on the obvious density difference of the solid phase and the liquid phase, a rotational flow field can be generated by pool water through the rotational flow impeller, solid-liquid separation is conducted on substances with different specific gravities, corrosion products are efficiently removed through gravity settling, the purpose of purifying the spent fuel pool is achieved, and compatibility with an existing circulating system of the spent fuel pool is high.

The invention relates to a method and system for measuring the thickness of a nuclear fuel cladding oxide film, the method is used for a differential probe, and the differential probe comprises a detection probe used for measuring the thickness of the oxide film, a reference probe arranged in a spent fuel pool and a sampling circuit used for collecting differential signals of the detection probe and the reference probe. The method for measuring the thickness of the nuclear fuel cladding oxide film comprises the following steps: acquiring a differential signal; calculating the original thickness of the oxide film according to the differential signal; acquiring a detection measurement temperature of a detection coil in the detection probe and a reference measurement temperature of a reference coil in the reference probe; calculating a compensation value according to the detection measurement temperature and the reference measurement temperature; and compensating the original thickness according to the compensation value to obtain the measured thickness of the oxide film. According to the invention, accurate compensation of the thickness of the oxide film is realized, the problem of temperature drift can be suppressed, and the accuracy and stability of measurement are effectively improved.

A long-cycle, high-burnup fuel management method based on accident-tolerant fuel, belonging to the nuclear power field, is proposed. This method targets pressurized water reactors comprising 157 fuel assemblies and includes the following steps: providing a replacement group, comprising a first-class fuel assembly and a second-class fuel assembly using accident-tolerant fuel with an enrichment of not less than 5.5%, wherein the first-class fuel assembly includes IFBA combustible poison rods, and the second-class fuel assembly includes IFBA combustible poison rods and Gd-containing combustible poison rods; during refueling, one fuel assembly and 68 remaining reserve fuel assemblies with the lowest reactivity are removed from the core center, and the remaining reserve fuel assembly with the highest reactivity from the spent fuel pool is loaded into the core center and placed into the replacement group. This method enables a long-cycle refueling of 24 months, effectively flattening core power and improving the economics of nuclear power plants.

The invention belongs to the technical field of nuclear power, and particularly relates to a spent fuel pool purification filter resin recovery method. According to the method provided by the invention, the oxidation performance of the spent fuel pool is fully utilized, the recovery of the ion exchange resin is realized, the extra operation cost and the generation amount of radioactive solid wastes caused by frequent replacement of the resin are reduced, the service life of the resin is prolonged, the addition of traditional chemical agents is avoided, and the cost is reduced. By switching purification between the boric acid storage tank and the spent fuel pool, ion exchange resin is recovered while water purification is realized, the service life of the resin is prolonged, the operation cost of the unit is saved, and the problem that the cation exchange resin in the filter fails due to purification of water in the boric acid storage tank is solved.

The utility model relates to an air pressure adjusting device for verifying a liquid level instrument in a spent fuel pool, which comprises a box body, and an internal pipeline, an air inlet pipeline, an exhaust pipeline and an air supply pipeline are arranged in the box body; the box body is provided with an air source interface capable of being connected with an external air source, an exhaust interface capable of exhausting air outwards and an instrument interface capable of supplying air outwards; one end of the air inlet pipeline is connected with the internal pipeline, and the other end of the air inlet pipeline is connected with the air source interface; one end of the exhaust pipeline is connected with the internal pipeline, and the other end of the exhaust pipeline is connected with the exhaust interface; one end of the air supply pipeline is connected with the internal pipeline, and the other end of the air supply pipeline is connected with the instrument interface; and the high-precision pressure gauge is connected with the internal pipeline and is used for monitoring and displaying an air pressure value in the internal pipeline. The air pressure adjusting device has the remarkable advantages in the aspects of flexible operation and the like, and the requirement for providing stable pressure for the liquid level instrument in the spent fuel pool for verification can be effectively met.

Leakage detection systems and methods of monitoring leakage using probes of fiber optic and / or Time Domain Reflectometry (TDR) cables that provide at least one of an electric and electromagnetic signal upon coming in contact with a liquid from the pool. An interrogator may be multiplexed with several of the cables and generate interrogation signals whose reflection determines an exact position along the cable of leakage contact. The cables may be run under a floor and / or wall of a spent fuel pool or other liquid volume in a nuclear power plant. Leakage even of deionized water near room temperature may be detectable. The cables may be a single serpentine cable or several straight cables wrapping around the volume. The cables may be small, such as 10 millimeters or less in diameter. Cables may be run behind liners at any pitch or interval, potentially between supporting structures and the liner.

The invention relates to a spent fuel pool liquid level measuring system and method.The system comprises a measuring pipe inserted into a spent fuel pool, the lower end of a verification measuring pipe is inserted into the bottom of the verification spent fuel pool and communicated with the verification spent fuel pool, and the upper end of the verification measuring pipe is exposed out of the top of the verification spent fuel pool and can be connected with a gas source device in a sealed mode; the verification air source device can inflate or exhaust air for the verification measuring tube to realize pressurization or decompression; a heating element which extends up and down and a measuring element which is adjacent to the verification heating element and extends up and down are arranged in the verification measuring tube; the heating control unit is connected with the verification heating element, and the signal processing unit is connected with the verification measuring element and can calculate the actual value of the current liquid level. Actual measurement of the liquid level of the spent fuel pool is achieved, and accuracy of water level monitoring of the spent fuel pool is ensured.

The invention belongs to the technical field of nuclear power plant aging life extension, and particularly relates to a performance supervision sample tree for a neutron absorber of a quadrilateral spent fuel storage grillwork. An upper pipe seat is arranged at the upper end of the quadrangular main body, a lower pipe seat is arranged at the lower end of the quadrangular main body, sample plate shelves are arranged on four sides of the quadrangular main body, 16 groups of sample plates are respectively placed on two groups of opposite sides, eight boron-aluminum plate sample plates are placed on eight groups on one side of each group of opposite sides, and eight groups of boron-aluminum coated sample plates are placed on the other side of each group of opposite sides; meanwhile, water-permeable holes are formed in the lower part of the sample plate shelf. The aging condition of a spent fuel storage grillwork neutron absorber material in a spent fuel pool is monitored, the actual performance of the neutron absorber material is evaluated, and the neutron absorber sample plate is put into a spent fuel storage grillwork and can be taken out regularly for detection.

Leakage detection systems and methods of monitoring leakage using probes of fiber optic and / or Time Domain Reflectometry (TDR) cables that provide at least one of an electric and electromagnetic signal upon coming in contact with a liquid from the pool. An interrogator may be multiplexed with several of the cables and generate interrogation signals whose reflection determines an exact position along the cable of leakage contact. The cables may be run under a floor and / or wall of a spent fuel pool or other liquid volume in a nuclear power plant. Leakage even of deionized water near room temperature may be detectable. The cables may be a single serpentine cable or several straight cables wrapping around the volume. The cables may be small, such as 10 millimeters or less in diameter. Cables may be run behind liners at any pitch or interval, potentially between supporting structures and the liner.

The present invention relates to a flexible connection device for the fuel transfer channel of a pool-type heating reactor. By using the device provided by the present invention, the fuel transfer channel can be ensured to be connected, and the rupture of the fuel transfer channel caused by the uncoordinated deformation of the reactor water pool and the spent fuel pool can be avoided. The inner sealing ring and the outer sealing ring of the device provided by the present invention are both made of rubber and adopt a corrugated telescopic structure, which can adapt to the uncoordinated deformation of the reactor water pool and the spent fuel pool and can avoid the rupture of the fuel transfer channel; the inner sealing ring and the outer sealing ring form a double-channel seal, and when the inner sealing ring is damaged, the outer sealing ring can still ensure that the pool water does not leak into the surrounding environment. Moreover, a leak detector for detecting leaked pool water is installed on the outer sealing ring of the device provided by the present invention. Once the pool water leaks from the inner sealing ring, an alarm signal can be issued in time; in addition, a water collecting pipe is also installed on the outer sealing ring, which can collect the leaked pool water and guide it to a designated area.

The invention relates to the field of nuclear power unit overhaul, in particular to a spent fuel pool temperature control system and method during nuclear power unit overhaul, and the system comprises a pressure container which is arranged in a refueling pool, and the refueling pool is connected with a fuel transfer cabin through a spent fuel transfer trolley track. The fuel transfer cabin is connected with the spent fuel pool through a sluice gate; two ends of the normal waste heat discharge pipeline A and the normal waste heat discharge pipeline B are respectively connected with the pressure container; two ends of the spent fuel pool cooling and purifying pipeline A are respectively connected with the spent fuel pool, one end of the spent fuel pool cooling and purifying pipeline B is connected with the refueling pool, and the other end of the spent fuel pool cooling and purifying pipeline B is connected with the spent fuel pool. The system adopts two rows of spent fuel cooling and purifying pipelines and two rows of normal residual heat removal pipelines to cool the spent fuel pool, so that the decay heat exporting capability of the system on the spent fuel assembly of the spent fuel pool is greatly improved, and the temperature of the spent fuel pool during overhaul is reduced.

The utility model discloses a nuclear power station spent fuel pool liquidometer pressure verification device, which comprises an installation box and a gas pipeline, the installation box comprises a box body and a box cover which are matched with each other, the box body is provided with a first side plate and a second side plate which are arranged in parallel at an interval, the first end of the gas pipeline is arranged on the first side plate, and the second end of the gas pipeline is arranged on the second side plate. The first end of the gas pipeline is used for being connected with a gas source device, the second end of the gas pipeline is located on the second side plate, and the second end of the gas pipeline is used for being connected with a connector of a measuring barrel in a spent fuel pool of the nuclear power station; the gas pipeline is sequentially provided with a safety exhaust valve, an isolating valve, an adjustable pressure reducing valve and a pressure gauge from the first end to the second end of the gas pipeline. The nuclear power station spent fuel pool liquid level meter pressing verification device is relatively simple in structure, high in use safety, convenient to manufacture and popularize, relatively compact in structure, convenient to transport and store and high in practicability.

This invention belongs to the technical field of nuclear facility safety equipment, and relates to a passive spent fuel pool cooling and replenishment system and waste heat utilization system. The system includes a spent fuel pool, a cooling pool, a spent fuel decay waste heat cooling system, a waste heat utilization system, and a passive replenishment system. The spent fuel pool is filled with coolant to cool the spent fuel assemblies, and the cooling pool is located above the spent fuel pool. The spent fuel decay waste heat cooling system relies on natural circulation to remove the decay waste heat generated by the spent fuel assemblies within the spent fuel pool. The waste heat utilization system is used to convert and utilize the decay waste heat energy of the spent fuel assemblies. The passive replenishment system ensures timely replenishment of the spent fuel pool under both normal and accident conditions. The system proposed in this invention can passively replenish the spent fuel pool under normal and accident conditions during power loss accidents, and solves the problem of low economic efficiency associated with spent fuel pools.

The invention discloses a spent fuel pool underwater additive repairing device and method, and belongs to the field of additive manufacturing. The underwater additive repairing device comprises a radiation shielding device, an underwater laser additive device and a sealing device, wherein the underwater laser additive device and the sealing device are arranged on the inner side and the outer side of the radiation shielding device respectively. The radiation shielding device realizes dynamic sealing and radiation protection through a shielding top cover, a shielding cylinder and a split type shielding bottom cover; the sealing device adopts a pneumatic motion assembly and a vacuum suction technology, and is combined with a semi-ring-shaped sealing belt and a strip-shaped sealing belt, so that the sealing performance of the device in an underwater environment is ensured. The underwater laser additive material device and high-pressure inert gas form a local dry cavity through the drainage cover, and an ideal environment is provided for laser additive material repairing. Through non-contact remote operation, high-precision laser repair and multiple radiation protection measures, the radiation risk of personnel is remarkably reduced, the repair efficiency and quality are improved, the service life of the spent fuel pool is prolonged, and the in-situ maintenance system is suitable for in-situ maintenance of key facilities of the nuclear power station.

The invention discloses a liquid level monitoring instrument and monitoring method for severe accidents of a nuclear power plant, the instrument comprises a mounting plate, the mounting plate is provided with a movable monitoring box, and the monitoring box is provided with a liquid level display screen; a monitoring pipe and a fluid pipe which can communicate with each other penetrate through the bottom of the monitoring box, and a distance measuring generator is arranged at the upper end of the inner wall of the monitoring pipe; the method comprises the following steps: S1, starting liquid conveying and supplying liquid for monitoring; S2, monitoring the displacement of a floating block and calculating the height of the liquid level; S3, discharging the liquid in a monitoring pipe and realizing resetting of the floating block; S4, starting air conveying and cleaning and maintaining the monitoring pipe; the liquid level monitoring system effectively overcomes the defects of traditional nuclear power plant spent fuel pool liquid level monitoring equipment in precision, adaptability, safety and practicability, can stably and accurately complete a liquid level monitoring task in a serious accident scene, and provides powerful guarantee for safe operation of a nuclear power plant.

The application belongs to the technical field of nuclear power, and particularly relates to a boron and water supplement system and method for a spent fuel pool of a nuclear power plant. In view of the problems in the prior art, the application proposes, in the event of an extreme accident of complete loss of in-plant power and borate solution leakage, supplementing the borate solution by using a spare liquid storage space in the existing facilities of the nuclear power plant through gravity flow, which does not require power supply, and supplementing water by combining with external power equipment. The fatal defect that the prior art cannot supplement boron reserves in the extreme accident is effectively solved, and the risk of spent fuel returning to criticality due to the decrease of borate concentration caused by water supplementing alone is avoided. Without the need of large-scale modification of the existing equipment of the nuclear power plant, the modification can be realized only by optimizing the connection mode and operation logic, and the modification cost is low and the implementation period is short.

The utility model relates to a reactor core fuel assembly transfer device of a nuclear reactor and a transfer track thereof, the transfer device comprises a transfer track, the transfer track comprises a KX side transfer track, a KX side straight track and a KX side turnover track, the KX side straight track is located in a spent fuel pool of a fuel plant; the KX side overturning track is arranged above the KX side straight track so as to realize overturning of the loader; the RX side transfer track comprises an RX side straight track and an RX side turnover track, and the RX side straight track is located in a refueling pool of a reactor building; the RX side overturning track is arranged above the RX side straight track so as to realize overturning of the loader; the transfer channel is installed in a containment wall between the fuel plant and the reactor plant in a penetrating mode, an inner rail is installed in the transfer channel, and the two ends of the inner rail are arranged corresponding to the KX side straight rail and the RX side straight rail respectively; and the rail transfer mechanism is mounted at the bottom of the RX side straight rail corresponding to the RX side turnover rail, and is matched with the transfer trolley to realize high-low switching of the inlet position of the RX side turnover rail.

This invention relates to the field of nuclear power waste heat utilization technology, specifically a waste heat utilization system for spent fuel pools in nuclear power plants. It includes: a nuclear power turbine thermal system, which provides steam, which is condensed in a condenser to form condensate, which exchanges heat with the spent fuel pool waste heat exchange system; the spent fuel pool waste heat exchange system utilizes the condensate flowing through the spent fuel pool waste heat exchanger to exchange heat with the heat generated by the decay of spent fuel in the spent fuel pool, transferring heat to the condensate and increasing its temperature; a spent fuel pool cooling system transfers the waste heat from the spent fuel pool to the equipment cooling water via a heat exchanger; and the equipment cooling water loop system transfers the heat absorbed during equipment cooling to seawater. Its advantages are: using condensate as the main cooling water source for the spent fuel pool, realizing the utilization of spent fuel decay waste heat, increasing condensate temperature, and reducing the heat consumption rate of the turbine thermal system.

A method of handling spent nuclear fuel assemblies immerses the spent nuclear fuel assemblies in water in a relatively short time period when compared to traditional methods. A spent nuclear fuel assembly is removed from a nuclear reactor core, inserted into a sodium removal machine having a receiver, a cleaning vessel, and an elevator. A cleaning fluid is applied to the cleaning vessel and fuel assembly, and the fuel assembly is flushed with water while in the cleaning vessel. The cleaning vessel is at least partially submerged in the spent fuel pool during cleaning to provide passive heat removal. The cleaning vessel is lowered by an elevator into the spent fuel pool. The fuel assembly may then be loaded into a rack and / or a cask for long-term storage.

A nuclear fuel decay heat utilization system usable for space heating in one embodiment comprises a nuclear generation plant building housing a spent fuel pool containing submerged fuel assemblies which emit decay heat that heats the pool. Plural fluidly isolated but thermally coupled heat removal systems comprising pumped flow loops operate in tandem to absorb thermal energy from the heated pool water, and transfer the thermal energy through the systems in a cascading manner form one to the next to a final external heat sink outside the plant building from which the heat is rejected to the ambient environment. A programmable controller operably regulates the intake and flowrate of water from the heat sink into the heat removal systems and monitors ambient air temperature inside to building. The flowrate is regulated to maintain a preprogrammed building setpoint air temperature by increasing fuel pool water temperature to a maximum permissible limit.

The invention provides an underwater radioactive foreign matter identification method and device and a controller, and the method comprises the steps: controlling an underwater vehicle to cruise according to a preset cruise path, and obtaining the counting rates of four gamma radiation detectors; calculating a cruise detection limit in real time according to the counting rate and cruise local data; calculating a total count measurement value according to the count rate, and comparing the total count measurement value with a cruise detection limit; when the total count measured value is lower than the cruise detection limit, it shows that no radioactive foreign matter is found, cruise continues to be conducted according to the cruise path, and the latest count rate serves as cruise background data to be updated; and when the total count measured value is higher than the cruise detection limit, the radioactive foreign matter exists, and the orientation of the radioactive foreign matter is identified according to the counting rates of the four gamma radiation detectors and the radioactive foreign matter moves. The invention provides an efficient, safe and intelligent underwater radioactive foreign matter recognition and treatment scheme, the overall level of radioactive foreign matter management in the spent fuel pool is remarkably improved, and the method has wide application prospects and popularization value.

Leakage detection systems and methods of monitoring leakage using probes of fiber optic and / or Time Domain Reflectometry (TDR) cables that provide at least one of an electric and electromagnetic signal upon coming in contact with a liquid from the pool. An interrogator may be multiplexed with several of the cables and generate interrogation signals whose reflection determines an exact position along the cable of leakage contact. The cables may be run under a floor and / or wall of a spent fuel pool or other liquid volume in a nuclear power plant. Leakage even of deionized water near room temperature may be detectable. The cables may be a single serpentine cable or several straight cables wrapping around the volume. The cables may be small, such as 10 millimeters or less in diameter. Cables may be run behind liners at any pitch or interval, potentially between supporting structures and the liner.

The application provides a personnel intervenable time calculation method, system and medium under a spent pool accident scenario, and relates to the field of nuclear power. The personnel intervenable time calculation method comprises the following steps: obtaining refueling data of a nuclear power plant, and calculating total decay heat power of the spent pool before an accident according to the refueling data; obtaining spent pool data of the spent fuel pool and the number of newly added spent fuel assemblies corresponding to each accident time within the total duration of the accident; calculating a boiling time through time step iteration according to the number of newly added spent fuel assemblies, the spent pool data and the total decay heat power of the spent pool before the accident; and calculating a water supply starting time, a water supply ending time and a total required water supply amount through time step iteration according to the boiling time, the number of newly added spent fuel assemblies, the spent pool data and the total decay heat power of the spent pool before the accident.

The invention discloses a dirt self-separation ultrasonic cleaning and descaling device for a pressurized water reactor fuel assembly. The dirt self-separation ultrasonic cleaning and descaling device comprises a spent fuel pool, an assembly cleaning mechanism, a dirt collecting tank, a dirt depositing tank, a first water circulating pipeline, a first filter assembly, a first circulating pump, a second water circulating pipeline, a second filter assembly and a second circulating pump. The dirt collecting tank is arranged on the bottom side of the assembly cleaning mechanism. The dirt deposition tank is arranged below the dirt collection tank; one end of the first water circulating pipeline is connected to the top of the dirt collecting tank, and the other end is communicated with the spent fuel pool; the first filter assembly and the first circulating pump are arranged on the first water circulating pipeline; one end of the second water circulating pipeline is connected to the dirt collecting tank, and the other end is communicated with the first water circulating pipeline; and the second filter assembly and the second circulating pump are arranged on the second water circulating pipeline. The problem that after ultrasonic descaling is carried out on the fuel assembly, deposited scales are discharged into a spent fuel pool, and the radioactive level is increased can be solved.

The invention discloses a long-period high-fuel-consumption fuel management method based on accident fault-tolerant fuel, and belongs to the field of nuclear power. The method is directed to a pressurized water reactor comprising 157 fuel assemblies and comprises the steps of: providing a replacement group comprising a first type of fuel assemblies and a second type of fuel assemblies employing an accident tolerant fuel with an enrichment of not less than 5.5%, where the first type of fuel assemblies comprise IFBA burnable poison rods, and the second type of fuel assemblies comprise IFBA burnable poison rods, and the IFBA burnable poison rods comprise IFBA burnable poison rods and IFBA burnable poison rods; the second fuel assembly comprises an IFBA burnable poison rod and a Gd-containing burnable poison rod; during refueling, one fuel assembly in the center of the reactor core and 68 remaining fuel assemblies with the lowest backup reactivity are unloaded, one fuel assembly with the highest backup reactivity in the spent fuel pool is loaded in the center of the reactor core, and a replacement group is loaded. According to the method, 24-month long-period refueling can be realized, the reactor core power is effectively flattened, and the economical efficiency of the nuclear power plant is improved.

Provided is a long-cycle high-burnup fuel management method, which comprises the following steps: S1. determining the number of fuel assemblies to be unloaded, wherein the fuel assemblies to be unloaded include a group of fuel assemblies located at the center of the core; S2. loading fuel assemblies, wherein the fuel assemblies include a group of old fuel assemblies that can be reused in the spent fuel pool and the remaining groups of brand-new fuel assemblies; S3. replacing the fuel assemblies located at the center of the core with the old fuel assemblies and loading brand-new fuel assemblies, so that the loaded fuel assemblies and the fuel assemblies that are not unloaded are in a chessboard distribution, the loaded new fuel assemblies are the same enrichment fuel assemblies or different enrichment fuel assemblies, and the enrichment of the new fuel assemblies is greater than 5%; and S4. repeating steps S1 to S3 at an interval of a 24-month refueling cycle. The above method significantly increases the cycle length on the basis of using fuel with a nuclear fuel enrichment of more than 5.0%, and achieves the long-cycle refueling target.

The invention belongs to the technical field of nuclear power station spent pool liquid level detection, and particularly relates to a high-reliability compensation type cable liquid level meter which comprises an on-site measurement assembly and an on-site instrument box, and the on-site measurement assembly is connected with the on-site instrument box through a coaxial cable. The field measurement assembly is used for measuring the liquid level of the spent fuel pool in real time, generating a high-frequency electromagnetic wave analog signal and transmitting the high-frequency electromagnetic wave analog signal to the on-site instrument box; and the in-situ instrument box is used for receiving the high-frequency electromagnetic wave analog signal, converting the analog signal into a digital signal, processing the digital signal to obtain liquid level information of the spent fuel pool, and outputting the liquid level information of the spent fuel pool to the main control room. The liquid level of the spent pool in the nuclear power station can be accurately monitored.

The invention discloses a natural circulation and power generation circulation combined spent fuel pool cooling system, which comprises a spent fuel pool, a natural circulation cooling system and a power generation circulation system, and is characterized in that the natural circulation cooling system is partially arranged in the spent fuel pool; the other part of the natural circulation cooling system is arranged outside the spent fuel pool; the power generation circulating system is arranged on the other part; under the condition that no external input exists when an accident happens, under the cooperation of the natural circulation cooling system and the power generation circulation system, it is guaranteed that high-temperature heat can be discharged in time after the accident happens to the spent fuel pool, and a long enough time window is provided for later accident treatment.

This invention discloses a refueling system and method for a small nuclear power reactor. The system, located within the reactor building, includes a reactor pool, a spent fuel pool, and a fuel transfer channel. The spent fuel pool comprises a spent fuel storage area and other storage areas separated by walls lower than the spent fuel pool itself. The fuel transfer channel is a passage enclosed by walls and extends from the bottom of the spent fuel pool to the reactor pool. During refueling, the water in the reactor pool and spent fuel pool is at the refueling level, which is higher than the walls. The reactor pool is connected to the spent fuel pool via the fuel transfer channel, and the spent fuel storage area is connected to the other storage areas. After refueling, the water in the reactor pool and fuel transfer channel is drained, and the spent fuel pool is drained to the storage level, which is lower than the walls. The reactor pool and spent fuel pool, as well as the spent fuel storage area and other storage areas, are now isolated. This invention improves the reliability of the refueling system.