61 results about "Spent fuel pool" patented technology

A long-term fuel management method based on accident-tolerant fuel, comprising the following steps: providing replacement fuel assemblies with a first enrichment degree, a second enrichment degree, and a third enrichment degree, wherein 5% < the first enrichment degree < the second enrichment degree < the third enrichment degree, and the replacement fuel assemblies adopt accident-tolerant fuel; taking out 69 groups of fuel assemblies in the reactor and putting them into the spent fuel pool, selecting a group of fuel assemblies with the highest reserve reactivity from the spent fuel pool and setting it at the center of the reactor, and taking 68 groups of replacement fuel assemblies and the remaining fuel assemblies in the reactor as a refueling group; calculating the installation sequence of the refueling group according to the reactor design limit and safety certification conditions and installing it; repeating refueling every 24 months. This method can extend the fuel cycle period to 24 months, effectively improve the fuel utilization rate, and improve the economy of nuclear power plants.

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.

The present invention discloses a target component, belonging to the technical field of the core of a nuclear reactor, which includes a first end plug, a clamping tube, a second end plug, a connecting tube, a third end plug and a target rod. The first end plug is used to be connected with a pressing system, and the connection between the first end plug and the clamping tube is a movable connection. A limiting member is provided inside the connecting tube. The limiting member can move axially to be connected with the first end plug so that the first end plug is connected to the clamping tube in position, and the limiting member can move axially away from the first end plug so that the first end plug and the clamping tube are detachably connected. In the present invention, the connection between the first end plug and the clamping tube is a detachable connection, and locking and unlocking are achieved through a mechanical clamping structure, which is convenient for installation and disassembly, enables structural components such as the pressing system to be reused, reduces the cost of the radiation target, reduces the generation of radioactive waste, and at the same time, no debris or falling parts will be generated during the underwater disassembly process of the target component, and it will not affect the water in the spent fuel pool.

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.

The present invention belongs to the technical field of dry storage of spent nuclear fuel in nuclear power plants, and particularly relates to an interference inspection tool for loading a dry storage container for spent nuclear fuel. Before loading the spent nuclear fuel into the sealed storage container, interference inspection can be carried out by using an interference inspection tool for loading a dry storage container for spent nuclear fuel provided by the embodiments of the present application, which can avoid the collision between the loading manipulator and the fixing screw of the storage grid inside the storage container after entering the spent fuel pool and prevent the occurrence of safety accidents.

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 method and system for commissioning management of a spent fuel pool cooling system. The method for commissioning management of the cooling system includes: obtaining the three-part decay heat power per unit mass of metallic uranium for each group of spent fuel assemblies; obtaining the mass of each group of spent fuel assemblies and calculating the total decay heat power of the spent fuel pool by using a VBA tool; obtaining a pre-established heat exchange model of the spent fuel pool and determining the inlet temperature limit of the cooling water of the currently commissioned cooling system according to the heat exchange model of the spent fuel pool and the total decay heat power of the spent fuel pool; measuring the inlet temperature of the cooling water of the currently commissioned cooling system to obtain an inlet temperature measurement value, and determining a commissioning plan for the cooling system according to the inlet temperature measurement value and the inlet temperature limit. Implementing the technical solution of the present invention can not only save manpower and time, but also provide scientific opinions on the commissioning of the cooling system for operators.

Underwater radiation-resistant magnetic induction positioning and communication device and signal processing method. S1. Place the receiving and transmitting coils in water. The hardware circuits of the transmitting end and the receiving end are respectively connected to the upper computer and the lower computer through serial ports, and the device is started. S2. Generate a 100 kHz high-frequency sine wave through the DDS module, and use the serial port signal to control the analog switch to modulate the signal and switch between the two working modes of communication and positioning. S3. Control the signal selection module to select different signal transmission links. After being amplified by two-stage operational amplifiers and a high-frequency transformer, an alternating magnetic field is generated in space. S5. The receiving coil converts the alternating magnetic field in space into a voltage signal, and shifts the original 100 kHz signal to 455 kHz for filtering by means of spectrum shifting. S6. Use a single-chip microcomputer to cooperate with a programmable amplifier to amplify the signal. S7. Control the sampling frequency, and restore the original signal by using the spectrum aliasing phenomenon caused by undersampling. The system of the present invention can operate stably in the spent fuel pool environment.

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 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 used for realizing high-low switching of the inlet position of the RX side turnover rail.

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 present invention provides a nuclear fuel charging / discharging and position automatic tracking system and method. The system comprises: fuel assemblies, provided with first identification information capable of uniquely identifying the fuel assemblies; inserts, arranged in the fuel assemblies and provided with second identification information capable of uniquely identifying the inserts; a system database, storing reactor core arrangement information and spent fuel pool arrangement information; and a processing device, which, according to the current reactor core arrangement information and the spent fuel pool arrangement information, establishes a correspondence relationship with preset reactor core arrangement information of a next cycle, and configures shuffling movement information of each fuel assembly and a corresponding insert thereof, so as to generate a shuffling scheme. The present invention automatically generates a shuffling scheme according to the reactor core arrangement information and the spent fuel pool arrangement information in the system database of its own, so as to simplify the data processing and conversion operation during charging / discharging processes, thus improving the working efficiency, and reducing the risk of reactor core charging / discharging operation errors, and ensuring the operation safety of reactor cores.

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.

Disclosed in the present invention are a stainless steel cladding mounting structure and a mounting method therefor, and a spent fuel pool. The stainless steel cladding mounting structure comprises a cladding panel, at least one engagement rail arranged on a first surface of the cladding panel facing a concrete inner wall face, and an engagement groove engaged with the engagement rail and configured to be fixed to the concrete inner wall face, wherein the engagement rail corresponds to the side of a weld seam of the cladding panel, and a leakage detection groove enclosing the outer side of the weld seam is formed inside the engagement rail; in a pool, the cladding panel is vertically arranged and is fitted against the outer surface of the engagement groove, and the engagement rail is engaged in the engagement groove. In the present invention, a mounting structure is formed by mounting the cladding panel on faces such as concrete through the engagement of the engagement rail and the engagement groove, thereby realizing mounting of the cladding panel in the pool. The cladding panel and components such as the engagement rail and the engagement groove, and the fixation of the engagement rail on the cladding panel can all be prefabricated in a workshop, thereby achieving modularization, which in turn reduces the workload on a construction site and reduces on-site construction safety risks, achieving the aim of optimizing the construction progress and improving the construction quality.

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.

The invention discloses a device and a method for measuring torsional deformation in a nuclear fuel assembly unloading process, and belongs to the field of underwater nuclear fuel assembly measurement. The device comprises a measurement equipment main body and a data acquisition and processing system. The measuring equipment body is located in the transmission pool through the upper boss, the ten sets of line laser measuring modules are installed in the measuring equipment body, and the assembly supporting seat is located above a bottom plate of the measuring equipment body. In the unloading process, the unloaded fuel assembly is hoisted into the assembly supporting base from the fuel basket through the spent crane, the ten sets of line laser measuring modules collect contour data of a strip, an upper pipe base and a lower pipe base on the fuel assembly at the same time, and then the data collecting and processing system calculates and analyzes the torsional deformation amount of the fuel assembly. And after the measurement is finished, the fuel assembly is hoisted into the spent fuel pool grillwork through the sluice gate channel by the spent crane.

The invention relates to the technical field of power plant reactor building arrangement, and discloses an arrangement structure for arranging a passive safe water tank in a reactor building, which is characterized in that the safe water tank is arranged in the reactor building, and a plurality of partition plates are arranged in an inner cavity of the safe water tank; a plurality of partition plates are arranged in the safe water tank and divide an inner cavity of the safe water tank into a first storage cavity, a second storage cavity and a third storage cavity, the first storage cavity, the second storage cavity and the third storage cavity are mutually communicated or blocked through a plurality of communicating pieces, a power pump is arranged on the communicating piece between the first storage cavity and the third storage cavity, and the first storage cavity is used for being communicated with a spent fuel pool. The second storage cavity is communicated with the inner cavity of the pressure container, and the third storage cavity is communicated with the containing space. The invention relates to an arrangement structure for arranging a passive safety water tank in a reactor building, aiming at solving or improving the problems that the structure of the safety water tank is complicated and the volume of the reactor building is large.

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.