513results about "Nuclear reaction control" 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.

The invention belongs to the technical field of nuclear power, and particularly relates to a device and method for relieving remote expansion of a cushion block locking cap of a heat insulation sleeve. Comprising a locking cap expansion device and a long rod, the long rod is used for remote operation above a pool, and the locking cap expansion device comprises a positioning pin, a positioning cylinder, a clamping gripper, an aligning and screwing operation head, an expansion head and a pressing expansion operation head. The device has the beneficial effects that a long-distance long rod is adopted to operate screwing positioning alignment and pressing expansion, and remote operation of locking of the cushion block locking cap is achieved. A special device integrating positioning of an expansion tool, screwing positioning of an expansion head, clamping and fixing and pressing expansion is adopted, underwater remote expansion integrated operation of the relieving cushion block locking cap is achieved, and the process steps are simplified.

The application relates to a reactor core power distribution measurement uncertainty analysis method, device, equipment, storage medium and product. The method comprises the following steps: constructing a simulation operation state based on an actual operation state of a reactor core, determining state parameters of the reactor core in the simulation operation state, and obtaining group constants of the reactor core; performing calculation based on a material calculation subprogram and the state parameters to obtain material compositions of each fuel rod in the reactor core; performing calculation based on a power calculation subprogram and the material compositions of each fuel rod to obtain a real power distribution of the reactor core and obtain detector reaction intensities of the reactor core; determining power parameters of the reactor core in the simulation operation state based on the state parameters and the group constants, and performing flux map simulation based on the power parameters and the detector reaction intensities to obtain a simulation power distribution of the reactor core; and determining a measurement uncertainty of the reactor core power distribution based on a difference between the real power distribution and the simulation power distribution. The method can improve the accuracy and acquisition speed of the reactor core power distribution measurement uncertainty.

The invention provides a multipurpose small reactor control method, device, equipment and medium, and the method comprises the following steps: in an automatic temperature control mode, if the deviation between the actually measured temperature of a reactor outlet and the set temperature meets a preset automatic temperature control condition, determining a first control instruction of each adjusting rod of an adjusting rod group; a first control instruction is sent to an adjusting rod set executing mechanism so as to control an adjusting rod to move till the deviation between the actually measured temperature and the set temperature is within the temperature adjusting dead zone range; in the power automatic control mode, if the deviation between the set power and the actually measured power of the reactor meets a preset power automatic control condition, determining a second control instruction of the circulating pump of the intermediate loop; and a second control instruction is sent to the circulating pump executing mechanism so as to control the rotating speed of the circulating pump until the deviation between the set power and the actually measured power is within the power adjusting dead zone range. According to the invention, the temperature and the power are respectively controlled, and the control performance is optimized.

The present invention relates to a reactor construction comprising an upper compartment above a reactor compartment, said reactor compartment being separated from the upper compartment by a radiation shield, wherein a molten salt reactor (MSR) is situated in the reactor compartment, the MSR comprising a reactor vessel with a reactor vessel lid assembly comprising a reactor vessel lid shield and a reactor vessel lid wherein a lifting element is situated in the reactor construction, the lifting element comprising: at least one bar being substantially vertical and axially adjustable along its longitudinal axis by a drive mechanism situated in the upper compartment, and wherein at least one through-hole for the at least one bar is provided in the radiation shield and wherein the reactor vessel lid shield has at least one receiving structure configured to engage with and lock to the at least one bar. The invention also relates to a method of performing a maintenance operation in the reactor compartment and a power barge comprising at least one reactor construction.

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 engine system includes: a pump configured to pump a propellant; a fuel element including a set of nuclear fuel particles; a moderator configured to surround the fuel element and defining a set of moderator coolant channels configured to cool the moderator; a reflector including a neutron-reflecting material and a reflector coolant channel arranged within the reflector to cool the reflector, the reflector configured to at a first time, operate in a closed configuration to reflect neutrons to the fuel element to increase an energy flux, and at a second time, operate in an open configured to leak neutrons out of the engine system to decrease the energy flux; and a thrust nozzle configured to outlet propellant from the fuel element to produce thrust including a nozzle coolant channel arranged within a wall of the thrust nozzle configured to cool the nozzle.

The invention provides a MoAlB-SiC composite material as well as a preparation method and application thereof, and belongs to the technical field of ceramic matrix composite materials. The preparation method of the composite material comprises the following steps: mixing Mo powder, Al powder and B powder according to a molar ratio of 1: (1-2): 1, carrying out wet ball milling, drying, and carrying out vacuum pressureless sintering to obtain MoAlB powder; and S2, adding SiC particles with the mass percentage of 5%-20% into the MoAlB powder obtained in the step S1, carrying out wet ball milling, drying, and carrying out electric field assisted direct current heating rapid hot pressed sintering to obtain the MoAlB-SiC composite material. The MoAlB-SiC composite material obtained through the preparation method is high in density, secondary phases are reduced, the mechanical performance and the high-temperature oxidation resistance of the MoAlB material are enhanced by adding SiC, and the preparation method has great significance for preparing the high-strength MoAlB-based ceramic composite material.

The invention discloses a reactor control rod position control method and a related device, and the method comprises the steps: obtaining an original rod position sequence of a reactor control rod, decomposing the original rod position sequence, and obtaining M IMF components; inputting the M IMF components into a trained long short-term memory network model to obtain predicted values X1, X2,..., XM of the M IMF components, establishing an error evaluation index D-SMAPE according to the predicted values X1, X2,..., XM of the M IMF components and actual values thereof, and obtaining error coefficients delta Y1, delta Y2,..., delta YM corresponding to SWT-LMD subsequences; inputting the error coefficients delta Y1, delta Y2,..., and delta YM into a trained ResNet prediction model to obtain final IMF component prediction values Y1, Y2,..., and YM of the original rod position sequence; and controlling the control rod to act according to the final IMF component predicted values Y1, Y2,..., YM of the original rod position sequence. According to the method and the related device, the rod position of the reactor control rod can be accurately controlled.

The application discloses a nuclear power plant control rod guide tube, which comprises a protective cover and a full-range guide assembly installed in the protective cover; the full-range guide assembly comprises at least one full-range continuous guide element, and a full-range continuous guide channel is arranged in the full-range continuous guide element; the full-range continuous guide channel continuously guides at least one control rod in a single control rod assembly in a full range of the guide tube stroke. Compared with the prior art, the nuclear power plant control rod guide tube changes the non-continuous guide segment structure of the interval arranged guide panel into the full-range continuous guide segment formed by the long double-pipe structure, optimizes the control rod stroke protection function, reduces the risk of the control rod bending deformation caused by the transverse impact of the upper chamber fluid on the control rod, and guarantees the integrity of the control rod structure; meanwhile, the wear resistance of the specific position of the control rod guide tube is improved, so that the service life of the control rod guide tube is greatly prolonged, and the later operation cost is reduced.

The invention provides a rapid power reduction method and system based on an adjusting rod group and a pressurized water reactor, and relates to the technical field of nuclear reactors. According to the method and the device, the target rod group for rapid power reduction is selected from the grey rod group and the at least part of the black rod group in the adjusting rod group according to the target sequence, and the target sequence is the rod falling sequence determined based on the insertion sequence of the grey rod group and the at least part of the black rod group, so that rapid power reduction based on the adjusting rod group is realized; and the disturbance on the reactor core is small after the rod is dropped, and the influence on the shutdown capacity of a unit and the power distribution of the reactor core is reduced.

The application belongs to the technical field of control rod drive mechanism, and particularly relates to a heat insulation sleeve assembly of a magnetic lifting type control rod drive mechanism. The heat insulation sleeve assembly comprises a pipe base, a guide sleeve and a guide cover. The upper part of the pipe base is a first end, and the lower part is a second end. The radial dimension of the first end is larger than that of the second end. The lower part of the second end is provided with a first connecting end, and the guide sleeve is fixedly connected with the first connecting end of the pipe base. The upper part of the guide cover is provided with a fourth connecting end, and the lower part is provided with a horn cover. The fourth connecting end of the guide cover is fixedly connected with the second fixed end of the guide sleeve. The heat insulation sleeve assembly of the application is in a fixed state, and the problem of active wear of the heat insulation sleeve assembly is solved from the root.

To provide a nuclear power plant and a method for generating power in a nuclear power plant which can improve the load following property of a nuclear power generation plant and can further stabilize the output of a nuclear reactor.SOLUTION: The nuclear power plant having a nuclear reactor includes: an output control unit 70 for controlling the opening degree of an air extraction valve 13 according to a load request; and an energy storage unit for storing renewable energy. The output control unit 70 controls the temperature of water to supply to the nuclear reactor on the basis of the opening degree of the air extraction valve 13 and the amount of stored energy in the energy storage unit.SELECTED DRAWING: Figure 1

The invention relates to a reactor shutdown method and system, and the method comprises the steps: a reactor protection system receives a monitoring signal and carries out the preprocessing of the monitoring signal, and obtains a preprocessing signal; the reactor protection system is designed by adopting a plurality of safety sequences; the reactor protection system judges whether an emergency shutdown condition is met based on the pre-processing signal; if yes, the reactor protection system outputs a local shutdown signal; if not, not outputting the local shutdown signal; the control rod shutdown subsystem receives the local shutdown signal and cuts off a power supply of a control rod driving system according to the local shutdown signal; and the shutdown rod shutdown subsystem receives the local shutdown signal and cuts off a power supply of the shutdown rod driving system according to the local shutdown signal. According to the invention, the emergency shutdown of the reactor can be rapidly and effectively realized under the accident condition, the influence of common cause failure on the shutdown function can be effectively avoided, and a diversified instrument control system does not need to be designed to cope with the influence of the common cause failure of the reactor protection system.

A fuel element (100) for a nuclear reactor, the fuel element comprising non-fissile particles (110) dispersed throughout fissile fuel (120), the non-fissile particles comprising a core comprising neutron absorbing material, wherein the neutron absorbing material is positioned at an innermost radial portion of the core, wherein the core is configured to remain stable at a temperature no less than 1000 °C; and an outer layer surrounding the core, the outer layer comprising silicon carbide.

The application provides a nuclear reactor first cycle starting method and device, an electronic device and a storage medium, and belongs to the technical field of nuclear power. The application obtains neutron intensity change data of a secondary neutron source assembly, the neutron intensity change data including time steps and neutron release intensity of the time steps, selects the smallest neutron release intensity from the neutron intensity change data as a target neutron intensity, obtains the time step of the target neutron intensity if the target neutron intensity is greater than or equal to a preset neutron intensity threshold, obtains a target time, determines a target reactor entry time period of the secondary neutron source assembly according to the target time and a pre-obtained reactor starting time of the target nuclear reactor, and performs first cycle starting on the target nuclear reactor according to the target reactor entry time period and the secondary neutron source assembly. The application can allocate appropriate reactor entry time for the secondary neutron source assembly, and realizes first cycle starting of the nuclear reactor using the secondary source assembly to replace the primary source.

The invention relates to the technical field of nuclear reactor design, in particular to a safety assembly, a lead-cooled fast neutron reactor and a passive negative reactivity feedback method. The safety assembly comprises a first positioning piece, a second positioning piece and a containing piece. The first positioning piece and the second positioning piece are respectively connected with the accommodating piece so as to position the two ends of the accommodating piece in the linear direction, the accommodating piece is provided with an accommodating cavity for accommodating a coolant, and one end of the accommodating cavity in the length direction is communicated with the outside of the accommodating piece. The safety assembly is applied to the lead-cooled fast neutron reactor to serve as a safety barrier on the periphery of the reactor core, passive negative reactivity can be rapidly introduced, the structure is simple, implementation is convenient and fast, the nuclear reactor design specification and related technical requirements can be better met, and safety and reliability are further improved.

The invention discloses a neutron source assembly which comprises a neutron source rod body, a lower end plug, a composite upper end plug and a guide cylinder, and the neutron source rod body comprises a neutron source rod end plug, a neutron source rod cladding tube, an upper reflecting layer, a lower reflecting layer, a source box and a matched end plug. The neutron source assembly adopts the rod-in-rod design, the possibility of abrasion of the neutron source rod body in the transportation process and the like can be reduced, the neutron source leakage risk is reduced, the counterweight design of the composite upper end plug can counteract the buoyancy of a coolant in a reactor, and it is ensured that the neutron source assembly smoothly falls to the bottom and keeps stable; the neutron source rod body can be used in a pressurized water reactor and can also be used in a liquid heavy metal reactor. Meanwhile, the sealing packaging structure of the neutron source rod body can reduce the risk of neutron source leakage caused by abrasion in the transportation and in-reactor operation process, the arrangement of the guide cylinder provides guidance for the assembly of the neutron source rod body, and the probability of collision damage of the neutron source rod body in the assembly process is reduced.

A control rod drive mechanism having a torque tube, a control rod assembly including a connecting rod disposed within the torque tube and including an annular collar defining a key slot, an elongated key that is slidably receivable within the key slot, a lock cam assembly rotatably secured to a top end of the connecting rod and including a locking cam, and a holdout collar disposed non-rotatably within the torque tube, the holdout collar including a locking recess with an entry slot, wherein in a first axial position the lock cam assembly is rotatable with respect to the torque tube, and a second axial position the lock cam assembly is non-rotatable with respect to the torque tube.

The invention relates to a control method and device for a high-temperature gas cooled reactor, and the method comprises the steps: responding to a received power automatic control instruction inputted by a user, and obtaining a power automatic control parameter; and performing rod position control on each control rod in a control rod group of the high-temperature gas cooled reactor based on the automatic power control parameters, and acquiring and displaying at least one parameter of the high-temperature gas cooled reactor in the rod position control process.

A nuclear fission power plant (10) configured for extra-terrestrial use comprising a gas-cooled moderated nuclear reactor core (20) with a metallic fuel (30), a moderator (60), and a containment vessel (40) with a containment vessel inlet (42) and a containment vessel outlet (44). A neutron reflector (70) is disposed around a periphery of the gas-cooled moderated nuclear reactor core, and a plurality of neutronic control elements (50) comprising rotatable control drums disposed around the periphery of the gas-cooled moderated nuclear reactor core. A fluid circuit (90) is configured to receive hot gas from the containment vessel outlet and deliver cooled gas to the containment vessel inlet. An electricity generating system (91) is coupled to the fluid circuit so as to generate an electrical current.

The invention discloses a nuclear rod power generation device, the nuclear rod power generation device comprises a nuclear fuel device, a neutron source device, a control assembly and a water pipe device, the nuclear fuel device is arranged in the water pipe device, and the water pipe device comprises a water inlet and a water outlet; the nuclear fuel device comprises an inner sleeve and an outer sleeve, the inner sleeve is arranged in the outer sleeve, the two ends of the inner sleeve and the two ends of the outer sleeve are connected in a sealed mode, a vacuum cavity is formed between the inner sleeve and the outer sleeve, and nuclear fuel coatings are arranged on the outer side of the inner sleeve and the inner side of the outer sleeve. The neutron source device is arranged on the inner side of the inner sleeve through a guide pipe, and the control assembly adjusts the position of the neutron source device in the guide pipe. The invention provides a nuclear rod power generation device which adopts a safe and miniaturized nuclear reaction technology and can replace a huge nuclear reactor.

To enable reactivity control using a simplified configuration.SOLUTION: A nuclear reactor 201 is provided, comprising: a core section 11 not including nuclear fuel; and a plurality of fuel drums (fuel units) 251 arranged so as to surround the core section 11, and each having a neutron absorbing part 55 for absorbing neutrons and a nuclear fuel part 56, where positions of the neutron absorbing part 55 and the nuclear fuel part 56 are adjustable relative to the core section 11.SELECTED DRAWING: Figure 3

To provide a lower component attaching / detaching device capable of easily attaching a spool piece to a main body of a control rod drive mechanism.SOLUTION: And a push-up mechanism 14 that pushes up and holds the spool piece 32a to a height at which the spool piece 32a can be attached to the drive mechanism body 31 of the control rod drive mechanism 3 in a case where the spool piece 32a that can be included in the control rod drive mechanism 3 that drives the control rod cannot be held by the lifting and lowering mechanism 11 due to the expansion and contraction of the plurality of fluid pressure cylinders 11a.SELECTED DRAWING: Figure 10

A method for transitioning a nuclear reactor during initial cycle startup to a power generating state is disclosed. The method includes placing the reactor in a zero power state, omitting a low power physical test (LPPT) for the current cycle of the reactor based on a predetermined set of criteria, and transitioning the reactor to the power generating state without conducting the LPPT based on a cross-check. The cross-checking includes predicting a first set of values ​​for elements of the LPPT using a first design code, developing empirical equations for the elements of the LPPT using data from past cycles of the reactor, predicting a second set of values ​​for the elements of the LPPT using the empirical equations, and cross-checking the first values ​​against the second values.

A high temperature control rod (22) for a nuclear fuel assembly (10) is described herein that includes a neutron absorbing material (24) having a melting point greater than 1500 °C that does not form a eutectic with a melting point less than 1500 °C, and may further include a cladding material (30) having a melting point greater than 1500 °C. The cladding material (30) is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material (24) is selected from the group consisting of Gd2O3, Ir, B4C, Re, and Hf. The metal cladding (30) or the absorbing material (24) may be coated with an anti-oxidation coating (26) of Cr with or without a Nb intermediate layer (28).

A measuring device for determining the power level of the reactor core is provided, which includes an in-vessel detection assembly. The in-vessel detection assembly comprises a rotatable housing that defines a cavity and a detection element positioned at a first angular portion of the cavity within the housing. A control drum for the reactor core and a system for monitoring the reactor power level are also provided.

The invention relates to the technical field of nuclear reactors, in particular to a control rod, a lead-based fast neutron high-flux research reactor and an experimental method thereof. Wherein the research reactor comprises a reactor container, a supporting platform, a shielding layer, a reactor lower chamber, a guide pipe and a control rod; a fuel assembly is arranged in the reactor container; one end of the guide pipe is located in the reactor container and is of a sealed structure, the length direction of the guide pipe is parallel to the length direction of the fuel assembly, the other end of the guide pipe sequentially penetrates through the reactor container, the supporting platform, the shielding layer and the under-reactor chamber and is located in the under-reactor chamber, and a driving mechanism matched with the guide pipe is arranged in the under-reactor chamber; the control rod is located in the guide pipe and can be driven by the driving mechanism to move in the length direction of the guide pipe. In a working state, the height of one end of the sealing structure of the guide pipe is not lower than that of the fuel assembly. According to the invention, the control rod can be installed, enough top space is reserved for the reactor vessel, and the reactor vessel has inherent safety characteristics.