54 results about "Burnup" patented technology

The application provides an oxygen concentration interval checking method, and relates to the technical field of lead-bismuth fast reactor oxygen concentration interval checking, and the method comprises the following steps: establishing a lead-bismuth fast reactor model, obtaining an initial oxygen concentration interval of the lead-bismuth fast reactor model, and selecting a plurality of preset oxygen concentrations from the initial oxygen concentration interval; performing physical field calculation on the fuel rod under each preset oxygen concentration to obtain physical field parameters corresponding to each preset oxygen concentration; the physical field parameters comprise temperature distribution parameters, fuel rod geometric parameters, contact stress, cladding oxidation corrosion parameters and fission gas parameters; judging whether each preset oxygen concentration meets the use requirements of the fuel rod based on the physical field parameters corresponding to each preset oxygen concentration, and determining a qualified oxygen concentration interval based on each preset oxygen concentration that meets the use requirements of the fuel rod. The qualified oxygen concentration interval is obtained by checking the initial oxygen concentration interval, and the problems of heat transfer deterioration, inability to inhibit corrosion and inability to meet the long-term use requirements of the fuel rod during fuel rod burnup can be avoided.

This invention discloses a U-shaped neutron detection device for detecting the burn-up depth of spent fuel, belonging to the field of non-destructive testing technology for high-level radioactive spent fuel. It includes: two opposing arm detection components; a back detection component; each arm detection component includes: a moderator material; multiple arm He-3 counters spaced apart within the moderator material; a gamma shield disposed on the outer wall of the moderator material; and a cadmium sheet disposed on the outer wall of the other side of the moderator material; the back detection component includes: a support material, with both ends of the support material fixedly connected to the ends of the two moderator materials; multiple back He-3 counters spaced apart within the support material; a base plate disposed on the outer wall of the support material; and a cable connection assembly disposed on the side of the base plate away from the spent fuel component to be measured. This invention provides dual protection for the measurement of spent fuel burn-up depth by utilizing fission neutron information and transmitted neutron information, meeting the requirements for online detection.

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.

This invention provides a method for evaluating the properties of fuel debris containing a mixture of nuclear fuels with varying burnup levels. [Solution] The evaluation method includes the step of calculating the fissile material ratio indicating the properties of the fuel debris based on the evaluated values ​​of the Cm-244 count rate and the amount of fissile material of the fuel debris, data showing the correlation between the ratio of the Cm-244 count rate and the amount of fissile material and the fissile material ratio, and / or the evaluated values ​​of the Eu-154 count rate and the amount of fissile material of the fuel debris, data showing the correlation between the ratio of the Eu-154 count rate and the amount of fissile material and the fissile material ratio.

This invention relates to a method and system for optimizing the irradiation energy spectrum of reactors producing radioisotopes. The method employs an intelligent optimization algorithm (genetic algorithm) to construct a large number of irradiation energy spectra, builds a burnup database based on these spectra, and performs burnup calculations to determine the yield of various radioisotopes. This determines the optimal energy spectrum for producing various radioisotopes and their theoretically highest yield, thereby improving the production efficiency of reactors irradiating various radioisotopes. Compared with existing technologies, the method of this invention can determine the optimal irradiation energy spectrum and theoretically highest yield for reactors producing multiple radioisotopes, overcoming the limitations of traditional methods that struggle to determine the optimal irradiation energy spectrum for reactor irradiation production of various radioisotopes and achieve energy spectrum optimization.

This invention belongs to the field of numerical calculation technology in nuclear reactor physics. It provides a method and system for calculating the burnup of a fuel assembly that moves with control rods. First, an axial mesh is generated at the interface of the axial material segments and merged with the initial mesh to obtain the mesh before the control rod position changes. Then, an axial mesh is generated at the interface of the axial material segments and merged with the initial mesh to obtain the mesh after the control rod position changes. The mesh after the control rod position changes is offset to obtain the offset mesh. Finally, the burnup weight of each mesh in the mesh before the control rod position changes is calculated, and the burnup value of each mesh in the offset mesh is calculated based on the burnup weight and the burnup value of each mesh in the mesh before the control rod position changes. This method ensures the matching of the burnup calculation mesh and the neutronics calculation mesh, improves the accuracy of burnup calculation without reducing the computational mesh size, and guarantees computational efficiency.

This application discloses a method, apparatus, electronic device, and storage medium for calculating fuel sphere burnup, relating to the field of nuclear reactor physics and thermal calculation technology. The method includes: dividing fuel spheres into multiple batches based on their distribution; establishing neutron transport equations for each batch of fuel spheres; calculating the self-shielding factor of each batch of fuel spheres; calculating the neutron flux density of each region of the reactor core using the homogenized cross section of the pebble bed region and the homogenized cross section of the reflector layer region; calculating the power density of each region of the pebble bed based on the neutron flux density of each region of the reactor core; calculating the neutron flux density and power density of each batch of fuel spheres at the current reactor core power level based on the neutron flux density of the pebble bed region and the self-shielding factor of each batch of fuel spheres; and calculating the burnup depth of the corresponding batch of fuel spheres based on the power density of each batch of fuel spheres at the current reactor core power level, thereby improving the technical accuracy of fuel sphere burnup calculation in non-uniformly packed pebble bed reactors.

This invention relates to a method for determining the assumed uranium-plutonium isotope composition of a reprocessing plant solution system. Based on the reprocessing target, the method determines the component parameters and irradiation history parameters of the target, and then determines a conservative combination of nuclide composition calculation parameters for criticality safety analysis. Nuclide composition calculations are performed for different control rod insertion states, and a nuclide composition database is established. Combining the control rod insertion range in the irradiation history of different reactor types' spent fuel assemblies and the conservative axial burnout distribution of the target components, the method determines the average burnout limit for spent fuel assemblies with different initial enrichment levels that satisfies the uranium-plutonium isotope composition assumption. After comparing and adjusting the iterative analysis with the range and distribution of average burnout of spent fuel assemblies with different initial enrichment levels, the method ensures that, under the determined uranium-plutonium isotope composition assumption and its corresponding average burnout limit for spent fuel assemblies, the reprocessing plant can handle the vast majority of spent fuel assemblies, thereby maximizing the economic efficiency of criticality safety design.

The application provides a precise calculation method and an automatic calculation system for the upper limit of a reactor control rod bank of a third-generation pressurized water reactor nuclear power plant. The precise calculation method can obtain the upper limit of the control rod bank under the conditions of any burnup, any power and any boron concentration, considering the actual operation of the reactor, and effectively supports the accurate control of the reactor by the nuclear power plant operators. In order to more conveniently and automatically calculate the upper limit of the control rod bank, the automatic calculation system is developed on the basis of the precise calculation method, thereby overcoming the shortcomings of long time consumption and high error rate of manual calculation, and further ensuring the accurate control of the upper limit of the control rod bank by the nuclear power plant operators. The automatic calculation system is repeatedly debugged and actually verified, proving the practicability of the upper limit of the control rod bank fast calculation system, and having a good industrial application prospect.

This invention discloses a simulation method for differential refueling in a pebble bed reactor based on a semi-Lagrange fixed-grid backtracking algorithm, relating to the field of fuel cycle calculation technology for high-temperature gas-cooled reactors. The method includes: calculating the macroscopic burnup depth and poison nucleus density of fuel spheres in each grid of a homogenized pebble bed using a core diffusion program; calculating the backtracking distance of each grid based on the flow velocity in the channel where each grid is located and the time step used in each refueling step, and determining the position before backtracking; calculating weighting coefficients based on this backtracking position, and updating the macroscopic burnup depth and poison nucleus density of fuel spheres in the calculation grid using a cubic spline interpolation function; and finally correcting the calculated macroscopic burnup depth and poison nucleus density using a correction equation. This invention solves the numerical dispersion and oscillation problems existing in traditional methods, significantly improving the smoothness and stability of core physics field calculations under differential refueling conditions.

The application provides a reactor control method based on a fast power reduction system, an electronic device and a medium, and relates to the technical field of nuclear reactors. The fast power reduction system comprises a plurality of control rods for fast power reduction, and the control rod comprises a gray rod group. The reactor control method based on the fast power reduction system comprises the following steps: obtaining a corresponding relationship between an insertion depth limit value of the control rod and a core burnup; and controlling an operation rod position of the control rod to be not more than the insertion depth limit value corresponding to a current burnup step according to the corresponding relationship. The application limits the insertion depth of the operation rod position of the control rod for fast power reduction, ensures that the fast power reduction system has sufficient rod value that can fall into the core, reduces the complexity of operation, and improves the safety of nuclear power plant operation.

The invention discloses a gadolinium-loaded metal hydride fuel short-life micro-reactor core design method and a reactor core, the reactor core is loaded with gadolinium-loaded metal hydride fuel with high enrichment degree and low uranium mass percent, and the existence form of gadolinium in the reactor core is set, and the uranium mass percent is matched with the gadolinium content; the reactor core adopts a rotating strand matched with a gadolinium burnable poison to control the residual reactivity of the reactor core, the rotating strand performs reactor core power adjustment, and the gadolinium adjusts the reactor core reactivity fluctuation in the life period. The reactor core power is 20 MWt, the reactor core life period is 50 EFPD, 2407 fuel rods are arranged in the reactor core, and the fuel rods comprise 847 gadolinium-loaded metal hydride fuel rods with low uranium mass fraction and 1560 gadolinium-loaded metal hydride fuel rods with high uranium mass fraction. According to the present invention, the strong absorption characteristic and the high burn-up rate of the gadolinium burnable poison are utilized, the existence form and the content of the gadolinium in the reactor core are adjusted, and the rotating strand is matched with the gadolinium burnable poison to control the residual reactivity of the reactor core so as to achieve the minimum fluctuation design of the reactivity of the reactor core in the whole life period, and improve the performance of the reactor core.

This invention belongs to the field of modular small modular reactor (SMR) core technology, specifically relating to a long-short cycle method for SMR core balancing. It includes: Step 1: The fuel enrichment for core balancing is 4.4%~5.0%; Step 2: The core balancing uses a partial low-leakage loading, three-batch refueling, and long-short cycle strategy; Step 3: A long-cycle scheme with three-batch refueling fuel assemblies is implemented; Step 4: In the three-batch refueling fuel assemblies, 12 new fuel assemblies are arranged on the periphery of the core, and 12 or 13 new fuel assemblies are arranged inside the core, with new and old fuel assemblies interleaved in different areas of the core; Step 5: The enrichment of the internal new fuel assemblies is 4.45%, with 8~24 gadolinium-loaded fuel rods; the enrichment of the external new fuel assemblies is 4.95%, with 0~8 gadolinium-loaded fuel rods. The beneficial effects of this invention are: effectively reducing the maximum burnup of fuel rods and supporting the improvement of overall core performance.

A method for analyzing the bubbling behavior of high-fuel-consumption fuel under a transient high-temperature condition relates to the technical field of reactor safety analysis, and comprises the following steps: establishing a three-dimensional model; subdividing the three-dimensional model to obtain a computational domain, and performing physical property assignment on each unit grid of the computational domain; an irradiation test process is simulated, and the fission rate of fuel of each unit grid is calculated; within the irradiation time, the fuel consumption is calculated and updated, and irradiation swelling and creep behaviors of the fuel core are calculated; calculating the total amount of fission gas released by the fuel core body; adding a virtual bubbling core in the three-dimensional model; based on the volume expansion increment of the fuel core body and the true stress of the cladding, simulating and calculating the swelling behavior of the fuel core body, the creep behavior of the cladding and the plastic deformation process of the cladding in the gradual heating process of the fuel by setting external heating boundary conditions; the method is used for solving the problem that a traditional method cannot accurately analyze bubbling changes in a fuel assembly.

The invention relates to the technical field of reactor core control and safety, in particular to a method and a device for predicting the boron concentration of a nuclear reactor. The method comprises the following steps: determining burnup calculation deviation of critical boron concentration before shutdown of a reactor; obtaining the re-critical theoretical boron concentration of the reactor when the reactor is critical; determining a critical boron concentration calculation deviation; and according to the critical boron concentration calculation deviation, the burn-up calculation deviation and the re-critical theoretical boron concentration, determining the target re-critical boron concentration. According to the method, boron concentration correction is carried out only through the deviation in the burn-up correction process and the deviation in the power feedback process under the condition that additional program calculation does not need to be carried out, and it is guaranteed that the deviation between the corrected theoretical re-critical boron concentration and the corrected actual re-critical boron concentration is greatly reduced; the problem that the deviation between the predicted value and the measured value of the re-critical boron concentration in the nuclear life period is too large is effectively solved.

The invention provides a high-temperature gas cooled reactor fuel loading and unloading system which comprises a pressure container, a surge bin and a broken ball separator, and a fuel supply bin is arranged above the pressure container and connected with a vacuum pump. The surge bin is connected to the supply pipeline and located between the pressure container and the fuel supply bin in the vertical direction, and a gas inlet of the surge bin is connected with a helium gas source through a gas inlet pipe. The broken ball separator is located below the pressure container, and a burnup detection device is arranged in the broken ball separator. The new fuel balls are directly conveyed into the reactor in a gravity conveying mode, the situation that the fuel balls are damaged due to the fact that airflow carries the fuel balls to impact and collide with the conveying pipe is avoided, and the integrity of the fuel balls is improved. New fuel balls can be safely and stably supplied, and the problems of high graphite dust content and frequent ball clamping in the conveying process of the fuel balls can be effectively solved.

This invention discloses a method and apparatus for analyzing the burnup lifetime of a self-powered neutron detector emitter, comprising: performing generational processing on the total neutron fluence absorbed by the self-powered neutron detector throughout its service life, and simultaneously performing layered processing on the emitter of the self-powered neutron detector; performing burnup analysis calculations on the neutron fluence of each generation and on each emitter layer; during the calculation of the neutron fluence of each generation, iteratively calculating the atomic nucleus density of each emitter layer for different service periods using the burnup calculation formula, thereby obtaining the burnup distribution information of the emitter under the total neutron fluence condition during the service period; wherein, during the iterative calculation process, for a certain iteration step, the atomic nucleus density of each emitter layer and the incident neutron fluence considering the self-shielding effect are given by the previous iteration calculation process. This invention can greatly improve the accuracy of burnup analysis of self-powered neutron detectors, providing more substantial and accurate data support for the design, maintenance, and replacement strategies of self-powered neutron detectors.

This invention discloses a method for operating a high-burnup spent fuel transfer container with graded pressure monitoring and depressurization, belonging to the field of nuclear fuel cycle technology, and includes three stages: loading, transfer, and unloading. In the loading stage, the container is connected to the loading well through-hole and filled with water to balance the water level before loading fuel. After completion, leakage testing and radiation monitoring are performed. In the transfer stage, a qualified container is transported as a whole to the target facility using a tractor. In the unloading stage, before opening the lid, the pressure in the lid chamber and inner chamber is detected via a quick-connect coupling. If overpressure occurs, the pressure is controlled by connecting to the facility's exhaust gas system. After fuel assembly damage monitoring, the container's inner cavity is cooled and cleaned. Subsequent steps include precise positioning, inner lid disassembly and cleaning for protection, and closed-loop recording. This method solves the problems of low positioning efficiency, high high-altitude risks, high personnel radiation dose, and equipment vulnerability in existing operations. It achieves graded pressure control and systematic pollution prevention and control, improving the safety, accuracy, and efficiency of operations, and has strong compatibility with existing facilities.

The invention provides a transport-burnup coupling calculation method based on adaptive burnup step size division, and belongs to the field of reactor physical computation.The method comprises the steps that an explicit relation model of average neutron flux density and burnup step size is established, and convergence criteria are set to achieve optimal step size estimation; and constructing an iterative solution model based on a strict neutron-flux density expression, and determining an optimal burn-up step length through a convergence criterion. Finally, convergence verification is adopted to ensure the constancy of neutronics parameters in the current step length, and the numerical error is controlled within a permissible range; and by comparing the burnup results of the optimal step length and the initial step length, the influence of the step length on the calculation precision is quantitatively evaluated. The method has the core advantages that on the premise that calculation precision is guaranteed, self-adaptive division of the burnup step length in the whole life period is achieved, the number of times of transport-burnup coupling calculation is reduced, only one-time transport calculation is needed in each burnup step length, the transport-burnup coupling iteration frequency is effectively reduced, and the overall calculation efficiency is remarkably improved.

The invention discloses a global-local coupling refined burnup calculation method for a high-flux reactor irradiation duct. The method comprises the following steps: 1, global reactor core calculation: carrying out component few-group homogenization section calculation and reactor core steady-state simulation by using a deterministic theory program to obtain neutron flux distribution; 2, local target transportation: establishing a fine local model by using a Monte Carlo program with reactor core flux distribution as a boundary condition, and carrying out high-precision transportation calculation to obtain fine neutron flux and cross section of the target; and 3, target burnup calculation: carrying out target burnup analysis by utilizing Monte Carlo output. And 4, burnup of other components: calculating burnup of other components of the non-irradiation target piece through a deterministic theory program. And 5, coupling iteration: determining a target section of a theoretical program by using Monte Carlo high-precision section correction, and restarting reactor core calculation to realize global-local burnup coupling of the whole reactor core. And 6, circularly judging: if the target burnup is not reached, repeating the steps 1-5, otherwise, ending. The method can solve the problem of section deviation caused by difficulty in obtaining local fine neutron-flux density and distortion of boundary conditions by a deterministic theory program, improves the target burnup calculation precision and efficiency, and is suitable for high-flux reactor irradiation target analysis.

The invention relates to the technical field of nuclear fuel burnup detection, and particularly discloses a micro-zone burnup test method based on 98Mo quantitative analysis aiming at the problem of low accuracy of burnup detection of different micro-zones of nuclear fuel in an existing nuclear fuel burnup measurement method. The micro-zone burnup test method comprises the following steps: S1, taking a nuclear fuel sample, and determining the Mo content of the sample by adopting EPMA; s2, performing isotope test on fission nuclide 98Mo and isotope thereof in the sample by adopting SIMS; s3, the content C < 98Mo > of the 98Mo is calculated; s4, calculating a micro-area fuel consumption value Bu of the nuclear fuel according to the following formula; wherein B is a microcell burnup value, 9.380 is a coefficient, m < 98Mo > is the relative atomic weight of 98Mo, m < UO2 > is the relative atomic weight of UO2, and FY is the yield of nuclide obtained by fission of 235U into 98Mo. According to the method, the damage to the nuclear fuel sample is small, the micro-zone burnup in the micrometer size range of the nuclear fuel can be efficiently detected, the micro-zone burnup value of the nuclear fuel and the micro-zone burnup distribution at different positions of the fuel are obtained, the measurement range is more accurate, and the accuracy is higher.

This invention discloses a device for neutron spectrum shift control and a nuclear reactor, relating to the field of nuclear reactor technology. The device includes one or more water-squeezing rods with neutron spectrum shift function. Each water-squeezing rod includes a first shell and multiple partitions. The first shell defines a receiving cavity. The partitions are spaced apart within the first shell, dividing the receiving cavity into multiple closed isolation cavities and at least one starting cavity. The isolation cavities and the starting cavity are arranged along the length of the first shell. The first shell has a communication hole connecting the starting cavity to the external environment, allowing coolant to enter the starting cavity. The partitions are configured to form liquid inlet channels connecting adjacent chambers after a preset contact time with the coolant, allowing coolant to enter. Through precise design of the coolant response time by the partitions, this device can achieve dynamic matching between the water inlet sequence and the nuclear reactor burnup process, making reactivity compensation smoother and more controllable.

The utility model provides a reactor structure for realizing a starting-free neutron source in subsequent circulation, which comprises a reactor body and a signal amplification device, and the signal amplification device is of a cavity shell structure and is arranged between a fuel assembly in the reactor and a neutron detector outside the reactor. And the neutron detector is located on a neutron radiation path between the fuel assembly and the neutron detector. According to the utility model, by utilizing the arranged signal amplification device, the thickness of a shielding water layer between the spent fuel assembly and the reactor external source range neutron detector is reduced during the subsequent circulating reactor core loading period, and the number of neutrons reaching the detector is increased, so that the reactor core can be loaded with the spent fuel assembly which does not reach the minimum burnup. The detector can obtain effective counting in the loading process without depending on a neutron source, so that the problem that the neutron fluence rate level in a reactor core is difficult to accurately measure by the out-of-core neutron detector under the condition that the signal-to-noise ratio is large enough is solved.

The application relates to a coal injection amount adjusting method and system, electronic equipment and a storage medium, the coal injection amount adjusting method comprising the following steps: acquiring average batch number of discharging per unit time, batch material volume information, volume between the interface of a furnace shaft and a furnace waist and a tuyere center line plane, and smelting cycle time length information; confirming a coal powder thermal lag time length according to the average batch number of discharging per unit time, the batch material volume information and the volume between the interface of the furnace shaft and the furnace waist and the tuyere center line plane; and confirming a coal injection adjustment starting time according to the coal powder thermal lag time length, the smelting cycle time length information and first batch of furnace charge entering a furnace time information after coke load changes. The application comprehensively considers single and multiple factors to confirm the coal injection amount adjustment starting time, improves the punctuality of the coal injection amount adjustment starting time, can accurately calculate the coal injection amount adjustment starting time when the coke load of the entering furnace charge changes, can be used for guiding blast furnace operators to adjust the furnace condition, and avoids the fluctuation of the furnace temperature and the waste of the fuel consumption.

The invention discloses a fuel assembly burnup limit value optimization method and system, and the method comprises the steps: measuring the boundary dimension of each discharged fuel assembly, so as to obtain dimension measurement information; determining reactor cavity minimum gaps corresponding to the multiple measured fuel assemblies according to the fuel consumption values and the size measurement information of the multiple measured fuel assemblies; obtaining a new fuel consumption limit value by increasing the initial fuel consumption limit value of the fuel assembly, and calculating the corresponding irradiation growth amount of the fuel assembly from the inflection fuel consumption value to the new fuel consumption limit value according to a pre-obtained maximum design limit value curve; according to the reactor cavity minimum gap and the calculated irradiation growth amount, whether the new fuel consumption limit value meets a preset feasibility condition or not is judged, and when the new fuel consumption limit value meets the preset feasibility condition, the new fuel consumption limit value serves as the optimized fuel consumption limit value. According to the technical scheme, the utilization efficiency of the fuel assembly can be improved on the premise that safety is guaranteed.

The invention discloses a nuclear reactor fuel rod damage positioning method based on fine nuclide component information, and belongs to the technical field of nuclear power station fuel assembly safety monitoring. The method comprises the following steps of: firstly, accurately simulating and acquiring nuclide components of each fuel rod in a reactor core in a specific operation history by utilizing physical-fuel coupling calculation of the reactor core; secondly, fission product release and migration models are integrated, activity concentrations of various nuclides released into a loop coolant after each fuel rod is damaged are calculated, and a damaged nuclide database containing high-dimensional feature vectors is constructed; and finally, performing probability matching on the actually monitored primary loop nuclide activity data and the feature vectors in the damaged nuclide database by applying a deep metric learning network, and outputting a position with the highest probability as a damage positioning result. The method breaks through the strong dependence of a traditional method on a simplified nuclide model and historical data of a specific reactor type, can achieve the online, precise and universal positioning of the position of the damaged fuel rod, and provides reliable decision support for the safe operation and fuel management of a nuclear power station.

The invention discloses a multi-physics coupling simulation method, device, equipment and medium for the combination of a research reactor deterministic theory and a probability theory. The method comprises the following steps: completing multi-cycle burnup calculation of a research reactor core under a preset boundary condition; calculating fuel fission heat release and gamma heat release of each target area of the research reactor core; determining coolant inlet flow velocity of each fuel element of the research reactor core; the minimum DNBR of the fuel element is determined and used for thermotechnical safety margin evaluation; feeding back a thermotechnical calculation result of the fuel element test loop to the physical model through a physical-thermotechnical parameter transmission channel, and obtaining a corresponding temperature calculation section based on the square root of the existing temperature section database and an adjacent temperature weighted interpolation; and iterating the section parameters and the material thermophysical property parameters to obtain a multi-physical field coupling analogue simulation result of the research reactor core. The invention belongs to the physical and thermal field of reactors. According to the method, the deviation of reactor physical design and target irradiation calculation can be reduced.

The invention discloses a high-temperature gas cooled reactor on-line burnup measurement system and a control method thereof, and the system comprises a feeding pipe which is provided with a feeding channel for fuel balls to pass through, a plurality of selective distributors, a positioning distributor which is communicated with the feeding channel and a plurality of positioning distributors, and a burnup detector. The selective distributor is used for distributing fuel balls entering the selective distributor from the feeding channel to any one of the multiple positioning distributors, the number of the burnup detectors is the same as that of the positioning distributors, the burnup detectors and the positioning distributors are in one-to-one correspondence, and the burnup detectors are used for detecting the burnup values of the fuel balls in the positioning distributors. The positioning distributor is used for distributing the fuel balls in the positioning distributor to one of the combustion circulation channel and the spent fuel storage channel according to the fuel consumption value detected by the fuel consumption detector. According to the online burnup measurement system for the high-temperature gas cooled reactor, burnup value detection can be carried out on a plurality of fuel balls at the same time, the material changing speed is increased, the fuel circulation rate is increased, and the operation stability of the pebble-bed type high-temperature gas cooled reactor is effectively guaranteed.

A method for replacement operation of control rod assemblies, according to one embodiment of the present invention, comprises the steps of: calculating a minimum degree of burnup of control rod assemblies required per cycle for each bank of a nuclear reactor based on an operation DB of a nuclear power plant; determining and classifying, as control rod assemblies to be disposed, control rod assemblies if the degree of burnup thereof is less than the product of the minimum degree of burnup and a margin of a set value; determining and classifying, as replaceable spare control rod assemblies, control rod assemblies having a degree of burnup greater than or equal to a difference between 100% and the minimum degree of burnup; determining and classifying, as control rod assemblies to be replaced, control rod assemblies having a degree of burnup that is less than the difference between 100% and the minimum degree of burnup and greater than or equal to a predetermined multiple of the product of the minimum degree of burnup and the margin of the set value; replacing, with the spare control rod assemblies, the control rod assemblies to be replaced; and managing, as history for an N+1 cycle, the remaining control rod assemblies from among the control rod assemblies to be replaced.

The application discloses a kind of single crystal silicon neutron irradiation monitoring method and system based on vanadium self-powered detector, belong to nuclear engineering field.Therein, single crystal silicon neutron irradiation monitoring method based on vanadium self-powered detector includes the following steps: the corresponding relationship of burnup depth and fuel assembly nuclear composition and the corresponding relationship of axial burnup distribution of fuel assembly and fuel cycle, burnup depth are calculated, and then the corresponding relationship of irradiation channel thermal neutron fluence and burnup cycle, burnup depth is calculated;With the position of irradiation channel thermal neutron fluence not more than given threshold as detector installation area, the corresponding relationship of vanadium self-powered detector signal and irradiation channel thermal neutron fluence installed in detector installation area is calibrated using activated foil, so as to monitor single crystal silicon thermal neutron fluence using vanadium self-powered detector.