112 results about "Nuclear reactor core" patented technology

The invention relates to the field of reactor data processing, and particularly discloses a numerical calculation method for solving a neutron transport equation. The method comprises the following steps: 1, describing the geometry of a reactor core of a nuclear reactor by utilizing a construction entity geometric modeling method, and subdividing a curved surface grid along a curved surface structure; 2, directly expanding a neutron flux distribution function in each curved surface grid by using a multivariate polynomial; 3, deriving a neutron transport equation and an integral residual equation of a boundary condition under each curved surface grid; 4, solving geometric integral parameters of each curved surface grid; step 5, establishing an adaptive linear system; and step 6, iteratively calculating and solving a maximum characteristic value and an unknown expansion coefficient.

A rapid inference method for a neutron flux distribution change of a reactor core of a nuclear reactor. In an offline training stage, a transient neutron transport equation is solved by means of a full-order solver, so as to obtain spatial distribution data of neutron flux and a delayed neutron precursor density changing over time, a reduced-order basis and a reduced-order coefficient data set of a training data set are obtained by means of proper orthogonal decomposition, and a reduced-order model is obtained by means of least squares fitting into the reduced-order coefficient data set; and in an online prediction stage, after the reduced-order model is solved on the basis of parameters of a predicted working state, the reduced-order basis is applied to a solved reduced-order coefficient, so as to reconstruct a physical field, which includes a neutron flux distribution of a reactor core, a delayed neutron precursor density distribution, etc. changing over time. In the present invention, data of a high-fidelity full-order model is used to construct a rapid prediction model, and a high-resolution change in the neutron flux of a reactor core over time is finely and efficiently described in a neutronics simulation, which plays an important role in reactor safety analysis, reactor data assimilation and inverse problem solving.

The application relates to a neutron absorption material and application thereof to a control rod for a nuclear reactor core. The neutron absorption material provided by the application is prepared by mixing ytterbium elements and holmium elements in a certain proportion, can effectively play the high neutron capture cross-section characteristics of the ytterbium elements and the holmium elements, effectively absorb thermal neutrons, and can ensure good chemical and thermal stability, so that the neutron absorption material can maintain its structure and performance stability for a long time in a high irradiation environment, effectively avoids the conversion of the neutron absorption material into a weak neutron absorption material after absorbing neutrons, and further controls the neutron absorption characteristics of the neutron absorption material. Further, the neutron absorption material provided by the application can still ensure the neutron absorption characteristics thereof and maintain the reactivity value under high burnup depth, so as to prolong the service life of the control rod for the nuclear reactor core.

A nuclear reactor core includes a plurality of fuel elements and moderator blocks to form a nuclear reactor core inner portion and a nuclear reactor core outer portion. The nuclear reactor core inner portion includes an inner moderator matrix formed of a plurality of inner holes that include a plurality of inner fuel openings with one or more fuel elements disposed therein. The plurality of inner holes further include a plurality of inner coolant passages to flow a coolant. The nuclear reactor core outer portion includes an outer moderator matrix formed of a plurality of outer holes that include a plurality of outer fuel openings with one or more fuel elements disposed therein. The plurality of outer holes further include a plurality of outer coolant passages to flow the coolant. The outer coolant passage diameter exceeds the inner coolant passage diameter.

The application discloses a kind of structural grid methods and systems of dividing triangular arrangement rod bundle assembly with wire winding, it is related to nuclear reactor core thermal-hydraulic analysis and calculation field, by using structured hexahedral grid to divide the flow channel of triangular arrangement rod bundle with wire winding, grid quality is higher, and calculation efficiency is more optimal;When establishing the structured grid of the flow cross section of triangular arrangement rod bundle assembly with wire winding, we fully consider the geometric characteristics of wire winding, thereby improve the geometric reduction degree of grid model;According to the calculation requirement, the number and distribution of grid nodes can be flexibly adjusted, and the axial length is shortened accordingly, so as to not affect the calculation precision, reduce the number of grids, save the calculation resource, and ensure the calculation quality.

Flow channels are formed into a nuclear reactor core receptacle and provide a flow path for leaking primary coolant. The flow path redirects the leaking coolant from an upwardly vertical direction, in which the mass flow would impart a lift off force to a core assembly seated within the receptacle, to a horizontal flow path direction which maintains hydraulic balance of the core assembly. The flow channels in the nuclear reactor core receptacle provide a passive measure to allow hydraulic hold down of the core assemblies even in the absence of mechanical hold downs.

This invention discloses a method, system, and medium for data mapping between a solid geometric model and an unstructured mesh, relating to the field of nuclear reactor core coupling programming technology. By converting the CSG geometric entity into a coupling intermediate mesh (unstructured mesh), and using the coupling intermediate mesh as an intermediary between the CSG geometric entity and other professional unstructured meshes, high-precision coupled data transmission and coupled calculation between the CSG geometric entity and other professional unstructured meshes can be achieved through mesh mapping and other methods, based on the geometric correspondence between the CSG geometric entity and the coupling intermediate mesh.

A cutting device and method for cutting a plurality of parallel fingers of at least one core component in a plane orthogonal to the longitudinal axis of the fingers, the fingers being adapted to be introduced into the fuel assembly of a nuclear reactor core. The cutting device includes: a first bearing plate including a plurality of abutments respectively for a finger to be cut. The cutting device also includes a rotatable cutting plate including a corresponding cutting blade for each abutment in the first bearing plate, the rotatable cutting plate being rotatable about a rotating axis between a starting position and a final position in a circumferential direction, the first bearing plate and the cutting plate being parallel to each other, the cutting blades including a plurality of groups of cutting blades. For each group of cutting blades the angle between the cutting blades and the corresponding abutments is different to an angle between the cutting blades and the corresponding abutments of any other group of cutting blades.

The application provides a nuclear reactor core device and a preparation method thereof, and belongs to the technical field of material welding. The preparation method of the nuclear reactor core device comprises the following steps: preparing a rib plate assembly, a W-shaped plate assembly and a C-shaped plate assembly; assembling multiple W-shaped plate assemblies and C-shaped plate assemblies into a whole circle and fixing the same by using inner and outer support tools, performing automatic welding by using a laser symmetrical welding method, and forming a W / C-shaped plate assembly; synchronously and alternately welding the W / C-shaped plate assembly on the inner and outer sides by using two automatic wire filling welding guns by using a double-sided layered symmetrical welding method, and realizing the connection of the W / C-shaped plate assembly with a bottom plate and a top plate; performing automatic butt welding of an outer half-ring plate by using laser welding and laser wire filling welding; and finally, welding internal fillet welds and performing nondestructive testing. By the cooperation of laser welding, automatic MIG welding, special tools and a symmetrical welding sequence, high-quality and high-efficiency automatic welding of the core containment vessel is realized, and welding deformation is effectively controlled.

The reactor core fuel assembly transfer device comprises the reactor core fuel assembly transfer trolley, the reactor core fuel assembly transfer trolley comprises a trolley body, walking wheels and guide wheels, the walking wheels and the guide wheels are arranged on the trolley body, and the trolley body comprises a bottom plate, a rear side wall, a front side wall, a left side wall and a right side wall. The vehicle body is connected with a driving mechanism, and one end, far away from the driving mechanism, of the bottom plate is provided with a hole for a loader to turn over and pass through; the travelling wheels and the guide wheels are respectively mounted on the left side wall and the right side wall; the reactor core fuel assembly transfer trolley realizes cross-rail travelling along the transfer rail through the travelling wheels and the guide wheels; the left side wall and the right side wall of the vehicle body are concentrically provided with hinge holes used for being connected with a loader respectively. No power alternation, no clamping stagnation and no power failure risk exist in the whole working process, and the transfer efficiency and safety are improved.

The invention belongs to the nuclear industry and can be used for the production of fuel pellets from uranium-molybdenum metal powder enriched with uranium-235 to 7%, which can be used as fuel elements for nuclear reactors. The sintering of the pellets is carried out in an inert argon atmosphere at a temperature in the range of 1100°C to 1155°C, as the initial powder, a particle size of 160 μm is used, with a molybdenum content of 9.0 to 10.5 wt.%. The powder is preheated at a temperature of 500°C for 10-20 hours (in an argon medium), and then cold-pressed into pellets in a mold at a pressure of up to 950 MPa. According to another method, when preparing uranium-molybdenum pellets using a binder (plasticizer), the pellets are heated in argon from 300°C to 450°C for a duration of 2-4 hours before sintering, in order to remove the binder. The invention allows to increase the uranium content in the fuel, to reduce the heat accumulated in the core of the nuclear reactor, and to reduce the energy release in the event of a violation of the normal operating conditions of the nuclear reactor, which will improve its safety and emergency stability.

The present invention provides a nuclear reactor core operation analysis method and system, which includes: establishing a solid core geometric model; meshing the geometric model and setting physical parameters and boundary conditions of the solid core; establishing a neutron nuclear database, performing neutron physics calculation based on the database and the solid core geometric model to obtain the power size and power distribution results of the solid core; performing thermal analysis calculation based on the solid core power size and power distribution results to obtain the temperature distribution results of the solid core; performing mechanical analysis calculation based on the core temperature distribution results of the solid core to obtain deformation geometry calculation results; iteratively performing neutron physics calculation based on the core temperature distribution results of the solid core and the deformation geometry calculation results until the results of the iterative calculation meet specific convergence conditions; and obtaining neutron physics parameters, temperature distribution parameters and stress-strain distribution parameters of the fixed core under steady-state conditions based on the results of the iterative calculation.

This application discloses a core power prediction method and apparatus, a core monitoring method and system, relating to the field of nuclear industry technology. The core power prediction method includes: acquiring target power detector data for a target core; the target power detector data includes values ​​from multiple power detectors; processing the target power detector data using a target power prediction model to obtain a predicted power value for the target core; wherein the target power prediction model employs a data grouping processing method, trained using historical power detector training data and their corresponding first historical power values. According to embodiments of this application, this core prediction method can monitor core power using a small number of detectors while ensuring the accuracy of core power prediction.

The invention relates to the technical field of nuclear reactor cores, and provides a reactor core adopting different hydrides for combined application, the reactor core comprises a plurality of fuel assemblies which are sequentially arranged in a zoning manner along the radial direction from the center to the periphery; part of the fuel assemblies simultaneously comprise uranium zirconium hydride fuel rods filled with uranium zirconium hydride fuel and uranium yttrium hydride fuel rods filled with uranium yttrium hydride fuel. Metal hydride fuel rods are introduced into the moderator-free reactor core, the reactor core neutron energy spectrum is slowed down, the reactor core fission reaction rate is increased, the reactor core critical mass is reduced, meanwhile, the equivalent nuclear density of uranium and hydrogen in different areas of the reactor core is adjusted in the mode that different hydride types are used in a combined mode, and reactor core power distribution is further flattened.

The invention discloses a visual test device for simulating high-temperature and high-pressure working conditions of a nuclear reactor core rod bundle channel prototype, the device is provided with a core test layer and a multi-stage depressurization layer, and a core test layer window on the side surface of the core test layer is sealed and insulated by using a sealing ring and an insulating spacer; an electric heating rod bundle is arranged in the core test layer; the multi-stage pressure reduction layer is formed by coaxially arranging a plurality of cylindrical pressure reduction layer windows outside the core test layer; circulating cooling water with stable pressure is respectively introduced into the core test layer and each pressure reduction layer, the pressure of each path of circulating cooling water is different, and the pressure difference of the two sides of each window is within the window material pressure bearing range; circulating cooling water in each pressure reduction layer is kept at a single-phase low temperature, so that visual observation is facilitated, and the temperature of the sealing material is ensured to be lower than the temperature resistance limit. According to the invention, the key technical problems of material strength limitation and thermal expansion mismatch in the high-temperature and high-pressure visualization test are solved, and a reliable test platform is provided for the high-temperature and high-pressure visualization of the reactor core rod bundle channel prototype.

The application provides a fuel and a method for increasing the neutron source intensity of a nuclear reactor, wherein the abundance of O in the fuel is 1% to 50% 18 O is 1% to 50%, and alpha particles and the enriched 18 O can significantly enhance the initial neutron source intensity of a nuclear reactor core in a pressurized water nuclear power plant, and help to improve the start-up performance and reduce the dependence on external neutron sources. In the passive start-up mode without using an external neutron source, the application improves the isotope abundance of O in the fuel, and improves the (alpha, n) reaction neutron source in the nuclear reactor core in the pressurized water nuclear power plant. 18 O in the fuel, and improves the isotope abundance of O in the fuel, and improves the (alpha, n) reaction neutron source in the nuclear reactor core in the pressurized water nuclear power plant.

The invention belongs to the technical field of nuclear reactor cores, and particularly relates to a fast thermal coupling metal hydride reactor core controlled by a rotating drum and a rotating drum control method, and the reactor core comprises the rotating drum, a reactor core active area, a coolant flow channel, a beryllium oxide reflecting layer and a thermal insulation shielding layer; the coolant flow channel is composed of a coolant flow channel inner barrel and a coolant flow channel outer barrel; the reactor core active area is wrapped by the coolant flow channel inner barrel, and oxide fuel rods and metal hydride fuel rods are arranged in the reactor core active area; the beryllium oxide reflecting layer is arranged on the coolant flow channel outer barrel in a sleeving manner, and the heat preservation shielding layer is arranged on the beryllium oxide reflecting layer in a sleeving manner; a plurality of rotating drums are arranged between the beryllium oxide reflecting layer and the heat preservation shielding layer in the circumferential direction. The oxide fuel rods and the metal hydride fuel rods are arranged in the reactor core active area at the same time, so that the fast thermal coupling design of a reactor core neutron energy spectrum is realized, and the reactivity control value of the rotating drum is improved.

The present application belongs to the technical field of nuclear reactor core thermal safety analysis, and particularly relates to a thermal safety envelope curve generation method for a candu reactor core irradiation target. The method comprises the following steps: step 1, data collection and arrangement; step 2, calculation of irradiation heat release rate; step 3, establishment of a CFD analysis model; step 4, preparation of temperature limit criteria; step 5, preparation of a thermal calculation sequence; step 6, two-dimensional large-scale calculation screening; step 7, three-dimensional detailed calculation statistics; step 8, generation of an envelope curve; and step 9, envelope curve-based evaluation. The CFD calculation model used is more accurate than the theoretical calculation or numerical calculation method used in conventional methods, and fully considers external fluid flow heat transfer, non-uniform power distribution on the circumference of the VFD component outer surface, and internal asymmetric heat transfer processes.

A nuclear reactor comprises a pressure vessel containing a nuclear reactor core. A reactor core cooling system comprises a standpipe including a plurality of orifices in fluid communication with a refueling water storage tank (RWST) to drain water from the RWST into the standpipe, and an injection line configured to drain water from the standpipe to the pressure vessel. In some embodiments the standpipe is disposed in the RWST, while in other embodiments the standpipe is disposed outside of the RWST and cross-connection pipes connect the plurality of orifices with the RWST. The reactor core cooling system may further comprise a valve configured to control flow through one orifice of the plurality of orifices in fluid communication with the RWST based on water level in the standpipe. The valve may comprise a float valve having its float disposed in the standpipe.

A metal hydride fuel burn-up self-sustaining reactor core controlled by burnable poison relates to the technical field of nuclear reactor cores, and comprises a low-enrichment-degree control rod-free fuel assembly located in the center of the reactor core, and a middle-enrichment-degree control rod-free fuel assembly arranged around the periphery of the low-enrichment-degree control rod-free fuel assembly. A middle-enrichment-degree safety rod-containing fuel assembly is arranged between the fuel tank and the fuel tank; a high-enrichment-degree control rod-free fuel assembly is arranged around the periphery of the medium-enrichment-degree control rod-free fuel assembly, and a medium-enrichment-degree compensation adjusting rod-containing fuel assembly is arranged between the medium-enrichment-degree control rod-free fuel assembly and the high-enrichment-degree control rod-free The periphery of the high-enrichment-degree control rod-free fuel assembly is filled with lead bismuth; the reactor core is used for solving the problem that a traditional reactor core is difficult to reach expected burnup depth and even causes short service life.

The utility model provides a control rod assembly, and belongs to the technical field of nuclear reactor core control rod assemblys.The control rod assembly comprises a cladding, a first end plug, a second end plug and an absorber, the first end plug and the second end plug block openings in the two ends of the cladding respectively, and the absorber is limited in the cladding through the first end plug and the second end plug; the variable-diameter section is arranged on the cladding, so that the wall thickness of the variable-diameter section is increased from the first end to the second end, the wear resistance of the cladding is improved under the condition that an external structure is not increased, and the control rod cladding is prevented from continuously abrading with a reactor internal component guide structure under the action of flow-induced vibration when the control rod stays at a fixed axial position for a long time; therefore, the control rod is prevented from wear-through damage and the structural integrity of the control rod is ensured.

The invention discloses a nuclear reactor neutron fluence determination method and device, and relates to the technical field of nuclear reactor core design and safety. Comprising the following steps: acquiring a first neutron fluence rate of a first position in an axial active section of a fuel assembly in a nuclear reactor; acquiring a correction coefficient of a second position outside the axial active section of the fuel assembly in the nuclear reactor; the correction coefficient comprises at least one of a space correction coefficient and an energy group correction coefficient, the space correction coefficient is used for correcting attenuation of the neutron fluence rate from the first position to the second position, and the energy group correction coefficient is used for correcting energy group structure deviation between a first energy group structure of the first neutron fluence rate and a second energy group structure of the second position; based on the correction coefficient, correcting the first neutron fluence rate to obtain a second neutron fluence rate of the second position; and determining the neutron fluence of the second position based on the second neutron fluence rate of the second position. According to the invention, the accuracy of neutron fluence prediction outside the axial active section of the fuel assembly can be improved.

The invention relates to an experimental device capable of being used for researching thermal hydraulic characteristics of a reactor core rod bundle of a nuclear reactor, and relates to the technical field of reactors. The experimental device capable of being used for researching the thermal hydraulic characteristics of the reactor core rod bundle of the nuclear reactor comprises a pressure-bearing shell, a lower sealing head, a measuring assembly, a flow channel, a simulation rod array and an upper sealing head, wherein a containing cavity is formed in the pressure-bearing shell; the lower sealing head is mounted at the lower end of the pressure-bearing shell; the measuring assembly is mounted at one end of the pressure-bearing shell away from the lower sealing head; the measuring assembly comprises a mounting flange and a screen module, the mounting flange is connected with the pressure-bearing shell, and the screen module is detachably mounted on the side, away from the pressure-bearing shell, of the mounting flange; the runner is positioned in the accommodating cavity, and one end is connected with the mounting flange; the simulation rod array comprises at least one simulation rod, one end of each simulation rod is fixed to the measuring assembly, and the other end of each simulation rod penetrates through the flow channel. The upper sealing head is installed on the side, away from the pressure-bearing shell, of the measuring assembly. The experimental device can meet the requirements of heat transfer characteristic experiments after water level swelling and drying in the rod bundle at the same time.

The invention discloses a nuclear reactor core power capability verification method and device, and the method comprises the steps: obtaining first xenon atomic nucleus density distribution data of a reactor core according to a target reactor core working condition point corresponding to a typical burnup condition for any typical burnup condition of a plurality of typical burnup conditions of the reactor core; performing order reduction processing on the first xenon atomic nucleus density distribution data to obtain a plurality of low-order primary functions corresponding to the typical burnup condition and a target range of expansion coefficients corresponding to the plurality of low-order primary functions; obtaining second xenon atomic nucleus density distribution data corresponding to the typical burnup condition according to a plurality of recovery expansion coefficients corresponding to the typical burnup condition and a plurality of low-order basis functions; determining reactor core power distribution data corresponding to the typical burnup condition according to the second xenon atomic nucleus density distribution data; and performing power limit value verification on the reactor core power distribution data corresponding to each typical burnup condition to obtain a power capability verification result. According to the invention, the power capability verification calculation efficiency can be improved.

The application relates to a kind of passive residual heat removal systems of fuel salt discharge tank, heat transfer working medium is solid at room temperature and is liquid after heating, the passive residual heat removal system includes fuel salt discharge tank, salt discharge pipe, natural circulation loop, air cooler and circulation maintaining heat exchange device, fuel salt discharge tank is communicated with nuclear reactor core by salt discharge pipe, fuel salt discharge tank is communicated with air cooler and circulation maintaining heat exchange device respectively by natural circulation loop;Wherein, circulation maintaining heat exchange device is in thermal contact with molten salt reactor main loop pipeline, to compensate the heat loss of passive residual heat removal system by the heat energy of normal working loop.According to the passive residual heat removal system of the application, heat transfer working medium is not water, in the normal operation of reactor, it is standby state, maintain natural circulation fluid in molten liquid state and slowly natural circulation, once accident occurs, the fuel salt of reactor core is discharged into discharge tank, and the natural circulation fluid of molten liquid state can quickly establish natural circulation to remove decay heat.

The invention belongs to the technical field of nuclear power, and particularly relates to a nuclear reactor core isothermal temperature coefficient estimation method and device, a computer readable storage medium, electronic equipment and a computer program product. The method comprises the following steps: acquiring isothermal temperature coefficient measurement data of a nuclear reactor core; segmenting the isothermal temperature coefficient measurement data to obtain data of each segment; performing linear fitting on each piece of segmented data to obtain a fitting slope of each piece of segmented data; and determining an isothermal temperature coefficient estimation result of the nuclear reactor core according to the fitting slope of each piece of segmented data. According to the method and the device, the isothermal temperature coefficient measurement data are not used as a whole for simple linear fitting any more, the isothermal temperature coefficient measurement data are segmented, linear fitting is respectively carried out on each segmented data, and then the fitting slope of each segmented data is comprehensively considered to determine the final isothermal temperature coefficient estimation result. Therefore, the precision of the estimation result is effectively improved.

The present application relates to a nuclear reactor control rod assembly, comprising a control rod and a starting neutron source. The control rod is inserted into a nuclear reactor core to shut down the nuclear reactor, and the starting neutron source is disposed on the control rod, which is used to control the startup of the nuclear reactor. Thus, inserting the control rod with the starting neutron source into the nuclear reactor core can start the nuclear reactor. After the nuclear reactor is started, i.e., when the nuclear reactor is in an operating state, the control rod with the starting neutron source is pulled out of the nuclear reactor core so that the control rod is located above the nuclear reactor core. At this time, when the nuclear reactor is in an operating state, the control rod is located above the nuclear reactor core, preventing the starting neutron source from reacting with neutrons in the nuclear reactor core, thereby reducing the consumption of the starting neutron source and extending the service life of the starting neutron source, thereby meeting the startup and monitoring requirements of long-life advanced nuclear reactors.

The application discloses a reactor neutron detector calibration method, system and computer equipment, relates to the nuclear reactor core monitoring technical field, and is based on the absolute neutron flux at a position of a reactor core measured by an activation method, combines the neutron flux distribution relation inside and outside the reactor, maps the absolute neutron flux at the position of the neutron detector outside the reactor, obtains the corresponding relation between the neutron detector measurement result and the absolute neutron flux at the neutron detector, and based on the corresponding relation and the neutron detector measurement result, the absolute neutron flux at the detector can be quickly and accurately obtained. The application only uses the activation method to calibrate the conversion factor of the neutron detector once, and under various conditions such as changes in the core arrangement, changes in the core control rod state, changes in the core power and the like, the absolute neutron flux at the detector can be directly obtained through the measurement result of the neutron detector, and multiple irradiation experiments are not needed, so that the measurement efficiency is improved, and the cost is saved.

Provided in the present application are a method and apparatus for determining a nuclear fuel loading and unloading mode, and a device and a storage medium. The method for determining a nuclear fuel loading and unloading mode comprises: acquiring a unit number of an adjacent core unit corresponding to a target core unit, the unit number indicating a positional relationship between the adjacent core unit and the target core unit; on the basis of the unit number, acquiring a nuclear fuel storage state of the adjacent core unit; on the basis of the nuclear fuel storage state and the unit number, determining a target core offset mode and a target loading auxiliary tool guiding mode for the target core unit when same performs nuclear fuel loading and unloading; and on the basis of the target core offset mode and the target loading auxiliary tool guiding mode, determining a target loading and unloading mode corresponding to the target core unit. By means of the embodiments of the present application, the efficiency and accuracy of determining a nuclear fuel loading and unloading mode for a nuclear reactor core can be improved.

To reduce monitoring cost in a nuclear system.SOLUTION: A nuclear system comprises: a nuclear reactor core; a heat conductor that conducts heat which is generated by nuclear reaction of nuclear fuel in the nuclear reactor core to outside by means of a cooling medium; a reactivity control device that has a plurality of control drums in which a neutron absorption part arranged outside the nuclear reactor core to absorb neutrons is provided; a state detector that detects an operational state of a nuclear reactor; and an analysis unit that reproduces the operation state of the nuclear reactor in a virtual space using a virtual model of the nuclear reactor on the basis of a detection result of the state detector.SELECTED DRAWING: Figure 3