967 results about "Control rod" patented technology

The invention provides a control rod driving mechanism operation state evaluation method fusing multi-source characteristic information, which comprises the following steps: collecting current data generated in an operation process of a control rod driving mechanism, preprocessing, intercepting characteristic parameters of each action time sequence point in a current signal, screening the characteristic parameters to obtain key characteristic parameters, and evaluating the operation state of the control rod driving mechanism according to the key characteristic parameters. Determining uncertainty parameters of the control rod driving mechanism, performing random variable analysis, defining action reliability, constructing a multi-field dynamics simulation model of the control rod driving mechanism, and inputting random variables to perform reliability evaluation on key characteristic parameters of the control rod driving mechanism based on the multi-field dynamics simulation model; and through a hierarchical structure from top to bottom, an operation state evaluation model fusing multi-source feature information is constructed, and the operation state of the control rod driving mechanism is evaluated. The control rod driving mechanism operation state management efficiency is greatly improved, and effective support is provided for comprehensive and accurate evaluation of the control rod driving mechanism operation state.

The invention belongs to the technical field of preparation of boron carbide ceramics, and particularly relates to a preparation method of high-purity porous boron carbide ceramics based on high-frequency induction heating, under the condition that a sintering aid is not added, a plunger is forced to move downwards due to volume expansion caused by CO2 generated by a melt, so that boron carbide powder is pressurized, the effect of hot pressing sintering can be achieved, and the effect of high-purity porous boron carbide ceramics can be achieved. The problems of low yield and high cost caused by the use of a hot-pressing sintering furnace can be avoided, the comprehensive performance of the prepared boron carbide ceramic can be improved, the process is easy to control, and the method is suitable for mass production of the boron carbide ceramic round rod for the high-temperature gas cooled reactor nuclear control rod; the prepared boron carbide ceramic has the advantages of pure phase, high porosity, uniform pore structure and stable mechanical property, especially has very good pore connectivity, and can ensure timely discharge of helium and improve the safety of the nuclear control rod when being used as a high-temperature gas cooled reactor nuclear control rod.

The invention discloses a pressurized water reactor rod changing method and system, electronic equipment and a computer readable storage medium, and the method comprises the steps: obtaining a plurality of rod pre-changing strategies corresponding to a first burnup point according to a plurality of rod changing interpolation relationships of a plurality of control rods when it is monitored that the current burnup point is the first burnup point, each control rod corresponds to a rod changing interpolation relation, the rod changing interpolation relation comprises a plurality of burnup points and a rod pre-changing strategy corresponding to each burnup point, and the rod pre-changing strategy is used for indicating the pre-insertion step number of the corresponding control rod; adjusting the pre-insertion step number of each control rod according to a rod pre-changing strategy corresponding to each control rod at the first burn-up point; and rod replacement processing is carried out based on the multiple adjusted control rods. According to the invention, disturbance of the pressurized water reactor can be reduced when rod sequence exchange is carried out on the control rod.

The invention discloses an electric steam generator with a waste heat recovery device, and relates to the technical field of steam generators, the electric steam generator comprises a main body, a heating cavity is fixedly connected to the inner side of the main body, a plurality of electric heating pipes are fixedly connected to the inner side of the main body and located on the periphery of the heating cavity, and a recovery unit is arranged on the outer side of the main body. The upper side of the main body is fixedly connected with a first water inlet pipe, the right side of the main body is fixedly connected with a steam discharge pipe and a gas pipeline, the recovery unit comprises a fixing frame, and the fixing frame is fixedly connected to the outer side of the main body. According to the electric steam generator with the waste heat recovery device, the main body, the heating cavity, the electric heating pipe, the smoke conveying pipe, the heat exchange pipe and the recovery barrel are arranged, generated high-temperature smoke waste heat is conveniently recovered, stirring plates on the upper side and the lower side are conveniently controlled to rotate through a control rod, a second gear, a first gear, a motor and a fixing ring, water is stirred, and the heat exchange effect is improved.

The invention belongs to the technical field of bolt remote operation, and discloses a reactor internal control rod guide cylinder bolt remote loosening and grabbing device which comprises a tool rotating positioning seat, and a bolt loosening and grabbing tool and an underwater vision auxiliary positioning module which are arranged on the same circumference are installed on the tool rotating positioning seat. And the distance between the circumference diameter and the axis of the control rod guide cylinder is equal to the diameter of the guide cylinder bolt. A bolt loosening and grabbing tool, an underwater vision auxiliary positioning module, guide cylinder bolts and a bolt locking cap are all arranged on the circumference of the same size and form a 90-degree position relation with one another, when the underwater vision auxiliary positioning module is aligned with a certain guide cylinder bolt through the underwater vision technology, the current circumferential position is set to be zero, and the position of the guide cylinder bolt is adjusted to be zero. When the bolts are loosened, the grabbing tool completes initial position calibration, and the positions of all the bolts can be accurately found by rotating the encoder on the positioning seat through the tool.

A molten salt nuclear reactor includes a vessel and a control blade assembly disposed within the vessel. The control blade assembly includes a drive rod and a control blade. The drive rod further includes an oblique surface wherein vertical motion of the drive rod and its oblique surface causes horizontal motion of the control blade into and out of a reactor section of the molten salt reactor.

The invention belongs to the technical field of nuclear power, and particularly relates to a device and method for remote installation of a heat insulation sleeve relieving cushion block. Comprising a gripping apparatus and a screwing tool, the gripping apparatus is used for gripping the cushion block, the bolt and the locking cap to mount and dismount the cushion block, and the screwing tool is used for positioning and mounting the relieving cushion block on the upper cover plate of the control rod guide cylinder and fastening the bolt. The clamping device has the beneficial effects that the air cylinder is adopted to drive the grabhooks to be opened and closed, clamping of the cushion block is achieved, meanwhile, it is guaranteed that the grabhooks are in an open state during air loss through the built-in springs, and the cushion block is prevented from falling off. A grabbing structure that the bolts and the locking caps are lifted along with the cushion blocks is adopted, meanwhile, the anti-falling function is achieved, and integrated operation of clamping, conveying, positioning and screwing of the cushion blocks is achieved.

The invention relates to the technical field of load switches, in particular to a load switch and a ring main unit thereof, the load switch comprises a cabinet body, a cabinet door, a support frame, a sleeve, a static contact and a moving contact, the cabinet door is hinged to the cabinet body, the support frame is fixedly connected with the inner wall of a shell, both the sleeve and the static contact are fixedly connected with the support frame, and a control rod is further hinged to the support frame; the movable contact is slidably connected to the sleeve, one end of the movable contact is hinged to the control rod, a spray head is fixedly installed in the sleeve, and a plurality of first cleaning pipelines and second cleaning pipelines are arranged in the spray head. The cleaning assembly cleans carbon deposits on the surfaces of the moving contact and the static contact, and the problems that the contacts are powered off due to carbon deposits on the contacts, an electric appliance cannot work normally, and the service life of the electric appliance is shortened are solved.

The invention relates to a control rod drop function availability automatic test method and a DCS system, and the method comprises the steps: receiving a test start instruction, entering a test mode, and in the test mode, obtaining the test sequence of each test rod group, and the measurement rod position and the counting rod position of a control rod; sending a bar moving instruction to the current test bar group; after the control rod action of the current test rod group is completed, the counting rod position is consistent with the target rod position, and the change of the measuring rod position of the control rod is detected; judging whether the change of the measurement rod position of the control rod accords with a preset threshold value; and if yes, judging that the control rod drop function availability test of the current test rod group is qualified. According to the invention, the availability test of the rod falling function of the control rod is changed from the traditional manual operation to the automatic test, so that the loss caused by human errors is effectively avoided; time consumed by the whole test is shortened, manpower is saved, and test execution efficiency is improved; and automatic checking is carried out, and a test result is given, so that human errors caused by long manual checking time are avoided.

The utility model belongs to the technical field of preparation of carbon nanotubes, and particularly relates to a carbon nanotube discharging device. The discharging device is arranged at the discharging end of the tubular heating furnace and comprises a collecting bin, and a collecting cavity communicating with a discharging opening of the tubular heating furnace is formed in the collecting bin. An operation opening is formed in the collecting bin, a flexible pipeline is connected to the operation opening, and the operation opening is connected with the near end of the flexible pipeline in a sealed mode; at least one operating rod penetrates through the flexible pipeline and extends into the collecting cavity, and the far end of the flexible pipeline is in sealed connection with the operating rod; a discharging connector is arranged at the bottom of the collecting bin. The collecting bin arranged at the discharging end is beneficial to timely removal of flocculent single-walled carbon nanotubes formed in the growth process, furnace tube blockage is prevented, production continuity is guaranteed, and safety is improved. And a flexible pipeline and an operating rod are arranged on the collecting bin, so that an operator can collect the carbon nanotubes under the condition that the normal operation of the heating furnace is not influenced. The winding roller is beneficial to orderly collection of carbon nanotube floccules.

The invention discloses a decoupling control system and method for a supercritical carbon dioxide direct circulation nuclear power plant. A compressor inlet temperature constant control strategy and a compressor inlet pressure constant control strategy are adopted. A compressor inlet temperature control system adjusts the opening degree of a cooler cold side valve, and a compressor inlet pressure control system adjusts the working medium filling and discharging of a capacity box. In order to guarantee the power supply quality, a transmission shaft rotating speed constant control strategy is adopted, and transmission shaft auxiliary torque is adjusted through a transmission shaft rotating speed control system. In order to ensure the system efficiency, a reactor core outlet temperature constant control strategy is adopted, and a reactor core outlet temperature control system is used for adjusting the opening degree of a control valve; in order to achieve the reactor following operation target, a reactor core power matching demand power control strategy is adopted, the control rod speed is adjusted through a power control system, and based on the neuron PID decoupling control technology, the control target that each output parameter in the system is only influenced by a certain specific input parameter is achieved.

The invention discloses a three-dimensional fluid-solid coupling simulation method and device for a rod falling behavior of a control rod drive line, and relates to the technical field of simulation of the rod falling behavior of the control rod drive line. Fluid boundary updating in the rod falling process of the control rod is processed through a structural grid and dynamic grid layer laying technology on the basis of computational fluid dynamics and multi-body dynamics theories; a modal synthesis method and a Hertz contact model are adopted to solve the problems of flexibility and contact friction of a large-length-diameter-ratio moving part, key factors including fluid resistance and mechanical friction resistance influencing rod falling simulation accuracy are calculated, fluid-solid coupling is achieved through one-way data transmission, three-dimensional fluid-solid coupling of the rod falling behavior of the control rod drive wire is achieved, and the control rod drive wire rod falling simulation accuracy is improved. The accuracy of a simulation result is improved, the degree of dependence on a prototype test is reduced, and a technical basis is provided for researching the rod falling behavior of the reactor control rod drive wire.

The invention provides a temperature early warning system and method for a rod control and rod position system cabinet. The system comprises a sensor, a temperature prediction module, a dynamic threshold module and a judgment module. The temperature prediction module is internally provided with a rod-control rod-position system cabinet temperature prediction model, and is used for obtaining temperature and wind speed data from the sensor, and carrying out the real-time cabinet temperature prediction through the rod-control rod-position system cabinet temperature prediction model. And a POT model is arranged in the dynamic threshold module, and the dynamic threshold module is used for acquiring temperature and wind speed data and generating a real-time threshold by using the POT model. And the judgment module is used for comparing and judging the predicted temperature value and the alarm threshold temperature value and giving an alarm. According to the method and the device, the rod-control rod-position system cabinet temperature prediction model, the POT model algorithm and the cabinet temperature real-time monitoring technology are combined, so that the problem that the accuracy of abnormity judgment performed by manually setting a fixed alarm threshold value and the predicted temperature is very low when the rod-control rod-position system cabinet temperature early warning is performed in the prior art is solved.

The utility model relates to the technical field of extrusion molding equipment, and discloses a quick color changing device for secondary coating of an optical cable, which comprises a main machine hopper, a support plate is fixedly connected in the main machine hopper, mounting ports are symmetrically arranged on the support plate, a first containing hopper and a second containing hopper are arranged at the mounting ports, and the first containing hopper and the second containing hopper are arranged in the main machine hopper. A color changing assembly is arranged on the supporting plate and comprises a control rod rotationally installed on the supporting plate, the upper end of the control rod is fixedly connected with a color changing knob, a color changing piece is fixed to the lower end of the control rod, and the color changing piece is in a cross shape. According to the color changing device, the color changing component is arranged, the color changing knob can be rotated to realize the function of releasing the color masterbatch, so that the natural color base material and the color masterbatch can be conveniently mixed, the function of quickly changing the color is realized, and the color changing device is compact in overall structure, does not need to additionally occupy the space at the upper part of the screw extruder and is convenient to use.

The invention belongs to the technical field of special maintenance of reactor bodies, and particularly relates to an electric spark machining head for cutting a sleeve flange in a narrow space. Comprising an expansion mechanism, a fine adjustment jacking mechanism, a rotating mechanism, an electrode and a radial servo feeding mechanism, the rotating mechanism provides rotary motion of the electrode and drives the electrode to conduct rotary machining, the radial servo feeding mechanism provides radial feeding motion of the electrode, and the expansion mechanism plays a role in fixing and positioning. The fine adjustment jacking mechanism drives the rotating mechanism, and the radial servo feeding mechanism drives the electrode to do up-down, rotating and radial feeding movement. The method has the beneficial effects that the damage to the pressure-bearing boundary of the primary loop of the nuclear power unit can be avoided under the condition that the heat sleeve is worn out of limits and needs to be replaced aiming at the common problem of wear settlement of the heat sleeve of the control rod driving mechanism of the nuclear power unit, and the whole dismantling work has relatively high safety, reliability and economical efficiency; and greater benefits can be created for the operation of a power plant.

The invention discloses a high-temperature gas cooled reactor control rod reactivity value measurement method and system, and relates to the technical field of high-temperature gas cooled reactor debugging and operation, and the method comprises the following steps: arranging measurement equipment to lift a to-be-measured control rod group to an upper limit, maintaining the reactor power stable operation, and carrying out initial data acquisition; calculating the theoretical reactivity value and neutron kinetic parameters of the control rod in a specific state, quickly inserting the control rod group to be measured to a lower limit, and recording data change; the reactivity value of the control rod is calculated, and the reactivity value obtained through calculation is corrected according to the initial power state of the reactor; and carrying out uncertainty analysis on the corrected reactivity value, and comparing a measurement result with a theoretical calculation value to optimize a measurement strategy. According to the invention, the time required for measurement can be reduced, the measurement precision is improved, the accuracy of the measurement result is enhanced, the measurement strategy is more perfect through continuous optimization, and the measurement result is more reliable.

The present invention relates to a nuclear fuel assembly suitable for a small modular reactor (SMR), the nuclear fuel assembly comprising: a spacer grid (110) having 17x17 grid cells; a plurality of fuel rods (120) arranged in the grid cells of the spacer grid (110); guide tubes (130) arranged and fixed in the grid cells of the spacer grid (110); and an upper end fitting (140) and a lower end fitting (150) fixed to the upper end and the lower end of the guide tubes (130), respectively, wherein the guide tubes (130) are characterized in that eight guide tubes are arranged in the main axis directions (Ixx)(Iyy) with respect to the central cell of the spacer grid (110), four guide tubes are arranged in the diagonal direction (Ixy), and sixteen guide tubes are arranged in the non-diagonal directions (Ixxy)(Ixyy).

The invention discloses a heating device for an insulating oil micro-water meter, and the heating device comprises a basic part which comprises a basic assembly, a supporting assembly disposed on the basic assembly, and a heating assembly disposed on the supporting assembly. The limiting component comprises a moving assembly arranged on the basic assembly, a sliding assembly arranged on the moving assembly and a control assembly arranged on the basic assembly. According to the heating device for the insulating oil micro-water meter, an insulating oil container filled with a solution is placed on a supporting plate in a placement groove, and the insulating oil container can downwards extrude the supporting plate, so that the supporting plate drives a moving block to downwards move through an L-shaped rod, and a long rod moves along with the supporting plate; and when the long rod moves downwards, the control rod moves in the chute, so that the arc-shaped plate is controlled by the long rod to move towards the middle to clamp the insulating oil container, and the insulating oil container body is prevented from moving.

The invention belongs to the technical field of processing of nuclear reactor control rod neutron absorber materials, and particularly relates to a preparation method of a nuclear-grade hafnium cast ingot. The method comprises the following steps: preparing sponge hafnium into an electrode block; performing vacuum electron beam melting on the electrode block for at least two times to obtain a finished cast ingot; and processing the finished cast ingot to prepare the nuclear-grade hafnium cast ingot which can be used for processing the control rod profile for the nuclear reactor assembly. According to the preparation method of the nuclear-grade hafnium cast ingot, the tensile strength, the welding performance and the creep performance of a nuclear-grade hafnium profile can be effectively improved; meanwhile, the technological process is simplified, and the technological economy and the production efficiency are remarkably improved; in addition, sponge hafnium is directly adopted for vacuum electron beam melting, so that the cost of raw materials is greatly reduced. According to the method, the application performance of the nuclear-grade hafnium profile is improved, meanwhile, the process is simplified, the production efficiency is improved, the production cost is reduced, and the economical efficiency of nuclear-grade hafnium profile manufacturing is greatly improved.

The present invention belongs to the technical field of ceramic preparation, and relates to a preparation method and device for a high-purity porous boron carbide nuclear control rod. The present invention uses boron trichloride gas as the boron source, methane as the carbon source, argon as the carrier gas and protective gas, and hydrogen as the catalytic gas and reducing gas to prepare a high-purity porous boron carbide nuclear control rod. The method of increasing the deposition temperature of boron carbide step by step and region by region from the back to the front is adopted, which can not only improve the deposition rate of boron carbide, but also ensure that the physical and mechanical properties of each part of the prepared high-purity porous boron carbide nuclear control rod are consistent. By changing the particle size of the boron carbide powder for preparing the boron carbide blank and coordinating the adjustment of temperature, flow rate, pressure and deposition time, the porosity and pore size of the prepared high-purity porous boron carbide can be regulated and a high total boron and carbon content can be obtained, so as to meet various actual requirements.

The invention belongs to the technical field of maintenance of upper reactor internals of a reactor, and particularly relates to a process and a device for maintaining a bolt locking cap of a control rod guide cylinder of an upper reactor internals. Comprising a bolt torque reinspection tool and a locking cap automatic bending tool. The bolt torque reinspection tool is matched with a torque wrench to realize reinspection of bolt torque; the automatic bending tool for the locking cap is used for controlling the bulging claw to correspond to a groove of a locking bolt head and driving the bulging claw to press downwards to bend the locking cap, so that looseness prevention of the bolt is achieved. The device has the beneficial effects that a bolt torque reinspection tool and an automatic locking cap bending tool are conveyed to a working area through a ring crane, and the field environment is not changed; the bolt torque reinspection tool and the locking cap automatic bending tool are both of a long-rod-shaped structure as a whole, an operator can perform remote operation on an operation platform, and whether the torque of the guide cylinder bolt meets the locking requirement or not can be inspected in cooperation with a torque spanner.

The invention discloses an online protection method and device for a reactor core, a medium and equipment, and the method comprises the steps: obtaining the rod position information of each control rod in a reactor in a current state, and obtaining a whole reactor power reconstruction coefficient in the current state according to the rod position information and a preset rod state database; based on the whole reactor power reconstruction coefficient, the detector current signal and the power plant working condition signal in the current state, the fuel assembly line power density and the deviation nucleation boiling ratio of each channel and the fuel assembly line power density and the deviation nucleation boiling ratio of the whole reactor are obtained through calculation; and comparing the linear power density of the fuel assembly of each channel, the linear power density of the fuel assembly of the whole reactor with a preset linear power density threshold value, the deviation nucleation boiling ratio of each channel, the deviation nucleation boiling ratio of the whole reactor and a preset boiling ratio threshold value, and determining and outputting a pre-protection signal and protection information. The reactor power reconstruction coefficient participates in protection parameter calculation, so that the accuracy and reliability of a protection signal are improved, and the safety of a power plant is improved.

The invention discloses a xenon oscillation automatic suppression auxiliary early warning prompting method based on a 3AO method, and belongs to the technical field of nuclear power reactor core monitoring. Firstly, various high-precision sensors are arranged at key parts of a nuclear reactor, occurrence signs of xenon oscillation are captured in time, and the system can accurately identify early signals of xenon oscillation by analyzing data of the sensors, so that a basis is provided for formulation of a subsequent control strategy. The core of the control strategy is fusion of a 3AO method and a deep Q network algorithm, the algorithm comprehensively considers the real-time state of the reactor core, and a control decision is optimized by learning historical data, so that the optimal control rod position and boron concentration adjustment scheme is calculated. And then, the system automatically prompts an operator to adjust the control rod and the boron concentration according to a control instruction generated by an intelligent algorithm so as to respond to the change requirement of the nuclear reactor in real time, and an efficient closed-loop control system is formed by feeding back an adjustment effect in real time, so that the accuracy and the reaction speed of a control strategy can be continuously improved.

The embodiment of the invention provides a reactor control method and device, electronic equipment and a storage medium, and belongs to the technical field of nuclear power. The method comprises the steps that the type of a steam turbine and a control strategy of a nuclear power unit are obtained, and the nuclear power unit comprises a reactor; according to the type of the steam turbine and the control strategy, an energy balance signal and a heat signal are calculated, the energy balance signal is used for indicating the minimum energy needed by operation of the steam turbine, and the heat signal is used for indicating the energy generated by the reactor; obtaining a feedforward signal according to the energy balance signal and the heat signal; and controlling the rod position action of the control rod according to the feedforward signal. According to the embodiment of the invention, the response speed and the adjustment capability of the nuclear power unit in the case of rapid load change can be improved.

To securely and easily disassemble and assemble a rack gear and a pinion gear. A drive mechanism 5 includes a rack gear 5A, a pinion gear 5B engaging with the rack gear 5A, a positioning member 5D fixed to a portion in the rack gear 5A where rack teeth 5Ab are not included, and a plurality of abutting members 5E disposed along a rotation direction of the pinion gear 5B. In a case in which at least one of the rack gear 5A and the pinion gear 5B is moved in a relative movement direction at the time when the rack gear 5A engages with the pinion gear 5B, abutment between the positioning member 5D and the abutting member 5E defines a rotational position of pinion teeth 5Bb of the pinion gear 5B to be an engaging position with the rack teeth 5Ab of the rack gear 5A.

The embodiment of the invention provides a self-adjusting shielding heat preservation device for a micro reactor and an operation method of the self-adjusting shielding heat preservation device, and relates to the field of nuclear reactors. The reactor comprises an outer shell, an inner shell, a control rod and a spiral flow channel, the inner shell is sleeved with the outer shell, an outer air inlet hole and an outer air outlet hole are formed in the outer shell, an inner air inlet hole and an inner air outlet hole are formed in the inner shell, a reaction cavity is formed in the inner side of the inner shell, and the spiral flow channels are arranged on the outer side of the inner shell and the inner side of the outer shell. The control rod is rotatably arranged on the side wall of the inner shell in a penetrating mode and located on the inner side of the outer shell, the control rod is provided with a control area and a non-control area, and the spiral runner is arranged between the inner shell and the outer shell; the outer gas inlet hole is used for inputting cooling gas, so that the cooling gas sequentially passes through the inner gas inlet hole, the reaction chamber, the inner gas outlet hole and the spiral flow channel and then is discharged from the outer gas outlet hole, and the cooling gas is used for flowing along the spiral flow channel and enabling the control rod to rotate so as to drive the control region to rotate into or out of the reaction chamber.

The invention provides a triaxial test soil sample preparation device capable of simulating sedimentary stratification distribution and a preparation method. The preparation device comprises a rack, a sample compaction cylinder, a stratification positioning assembly and a pressurization assembly, the sample compaction cylinder is provided with a through chamber for accommodating a sample; the bedding positioning assembly comprises a bedding thickness control rod, a first pressing die and a second pressing die. The first pressing die and the second pressing die are provided with extrusion faces matched with each other. The first pressing die and the second pressing die at least comprise a plurality of replaceable pressing dies with different inclination angles and are used for controlling different bedding angles; in the state that the sample bedding is pressed, the bedding thickness control rod and the pressurizing assembly are installed on the two sides of the cylinder body respectively and fixedly arranged with the rack, the first pressing die is fixedly installed at the tail end of the bedding thickness control rod, and the second pressing die is installed at the end, facing the pressurizing assembly, of the cylinder body. Or the second pressing die is fixedly installed at the movable end of the pressurizing assembly, and at least one of the first pressing die and the second pressing die can stretch into the cylinder body.

The present invention provides a temperature field analysis method and system for natural air cooling of a control rod drive mechanism. The analysis method comprises: establishing a control rod drive mechanism model, and setting different external air temperatures to perform simulation to obtain corresponding surface heat fluxes and first surface temperatures of the control rod drive mechanism model under different thermal loads; establishing a natural air cooling model, and performing simulation on the basis of the surface heat fluxes to obtain second surface temperatures of the control rod drive mechanism model; and on the basis of an interpolation method, determining a balance point temperature between the first surface temperatures and the second surface temperatures, wherein the control rod drive mechanism model comprises an internal flow channel, and the natural air cooling model is an integrated head package natural air cooling model. On the basis of the principles of sensitivity analysis, thermosyphon of a single drive mechanism model and natural air cooling of a full-field model are subjected to two-way coupling analysis, and finally, a drive mechanism surface heat flux and temperature field calculation result is obtained.

The invention discloses a high-temperature-resistant electromagnetic coil for a control rod driving mechanism of a nuclear power station, which comprises a shell internally provided with a high-temperature electromagnetic coil, a lead end of the high-temperature electromagnetic coil upwards penetrates out of the shell, an electrical penetration piece is arranged at the joint of the lead end and the shell, the electrical penetration piece is arranged on the upper end surface of the shell, and the high-temperature-resistant electromagnetic coil is arranged in the shell. The lead wire end upwards penetrates out of the electrical penetration piece; the outer side of the electrical penetration piece is sleeved with an insulating cloth sleeve, the lower end of the insulating cloth sleeve is arranged on the outer side of the electrical penetration piece in a sleeving mode, and the upper end of the insulating cloth sleeve is arranged on the outer side of the lead wire end in a sleeving mode. Through the collaborative design of the shell, the electrical penetration assembly and the insulating cloth sleeve, the problems of difficult maintenance and poor heat dissipation of a traditional resin encapsulation process are solved, the dustproof, insulating and high-temperature-resistant performances are considered at the same time, and a more efficient and reliable high-temperature electromagnetic coil solution is provided for a control rod driving mechanism of a nuclear power station.

The present invention relates to a method for producing a nuclear design core model accounting for load following, the method comprising the steps of: receiving load-following requirements for a corresponding cycle, wherein the load-following requirements include the amount of power reduction, the number of load-following operations, and at least one of the time or duration of operation at reduced power; determining the amount of insertion of a control rod on the basis of the load-following requirements; determining the duration of insertion of the control rod on the basis of the amount of insertion of the control rod and the maximum insertion limit thereof; determining the time of insertion of the control rod by simulating core combustion by reflecting the duration of insertion of the control rod in a nuclear design code at each of a plurality of time points of the corresponding cycle; and producing a nuclear design core model accounting for load following on the basis of the determined amount of insertion of the control rod, the determined duration of insertion of the control rod, and the determined time of insertion of the control rod.