229 results about "Burnup" patented technology

The invention discloses a method for determining metal fuel multi-phase structure irradiation microstructure evolution and performance change. Related physical property parameters are calculated through molecular dynamics and a first principle, and a diffusion coefficient is corrected by utilizing an irradiation acceleration theory to obtain a diffusion coefficient under a metal fuel irradiation condition; the method comprises the following steps: constructing a metal fuel polycrystalline multiphase organization structure evolution phase field model, defining phase field variables required in the model, and determining an evolution equation of the phase field variables; calculating to obtain a phase-field variable initial value, then solving the phase-field variable evolution equation, and obtaining a phase-field variable numerical solution; the phase field variable numerical solution is subjected to visualization processing, and a metal fuel polycrystalline multiphase structure distribution diagram is obtained; calculating the size and distribution of an irradiation structure in the metal fuel to obtain an evolution rule of the irradiation structure along with fuel consumption; and calculating the thermal conductivity, swelling and stress in the metal fuel by using the size and distribution of the irradiation structure in the metal fuel, and obtaining the change rule of the performance of the metal fuel along with fuel consumption.

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 relates to the field of theoretical calculation simulation of a nuclear reactor core state, in particular to a simulation method for dynamic reactor core power distribution measurement and a related product, and the method comprises the following steps: obtaining accurate burnup data; establishing a steady-state reactor core theoretical model based on accurate burnup; solving a three-dimensional neutron kinetic equation by taking the steady-state reactor core theoretical model as an initial state and taking a relation curve of reactor core operation power and operation time as input to obtain a dynamic reactor core theoretical model taking time as a function; inputting the power and time relation corresponding to the power and time relation corresponding to the power plant test plan into the dynamic reactor core theoretical model to obtain a database; according to the method, the steady-state theoretical model is constructed by utilizing accurate burnup data, dynamic reactor core simulation is carried out through a transient reactor core neutronics calculation model and the like, a dynamic reactor core model completely consistent with an actual test state can be generated, and actually measured data of dynamic reactor core power distribution can be directly processed and accepted.

The invention discloses a neural network-based axial power offset prediction method for a peak regulation process of a pressurized water reactor, and the method comprises the steps: 1, arranging power, fuel enrichment degree, the number of burnable poison rods in a fuel assembly, control rod positions, burnup level, axial xenon poison distribution, axial samarium poison distribution, radial xenon poison distribution, radial samarium poison distribution and the like according to a specific format; and making a standard input file. Secondly, arranging the AO (axial power offset) value of the reactor core under each target point in a specific format, and making a standard output file; thirdly, building a convolutional neural network model; and 4, putting the standard input file and the standard output file into the model for training until model parameters meet requirements, and exporting the trained model parameters. And 5, making a standard input file of a target point needing to be predicted according to the method in the first step. Sixthly, the standard input file of the target points needing to be predicted is calculated by loading model parameters, and the AO predicted value of each target point is obtained.

The invention discloses a monocrystalline silicon neutron irradiation monitoring method and system based on a vanadium self-powered detector, and belongs to the field of nuclear engineering. The monocrystalline silicon neutron irradiation monitoring method based on the vanadium self-powered detector comprises the following steps: calculating a corresponding relation between the burnup depth and nuclide components of a fuel assembly and a corresponding relation between axial burnup distribution of the fuel assembly and fuel circulation and the burnup depth; calculating to obtain a corresponding relationship between the thermal neutron fluence rate of the irradiation pore channel and the burnup cycle and the burnup depth; and taking a position where the thermal neutron fluence rate of the irradiation pore channel does not exceed a given threshold value as a detector mounting area, and calibrating a corresponding relation between a vanadium self-powered detector signal mounted in the detector mounting area and the thermal neutron fluence rate of the irradiation pore channel by using an activated foil, thereby monitoring the thermal neutron fluence rate of the monocrystalline silicon by using the vanadium self-powered detector.

The invention provides a pressurized water reactor core axial power offset phenomenon prediction method and electronic equipment. The prediction method comprises the following steps: firstly, constructing a prediction model of the axial power offset phenomenon of the reactor core; wherein the prediction model is used for representing a model that the axial power offset phenomenon of the reactor core changes along with oxidation corrosion sediment CRUD and fuel burn-up on the surface of a cladding, and the prediction model comprises a neutron physical module, a thermal hydraulic module and a CRUD material module which have a coupling relationship; and then, carrying out coupling calculation on the prediction model within a preset duration, and generating prediction information of the axial power offset phenomenon of the reactor core. According to the prediction mode, the reactor core axial power offset phenomenon changing along with fuel burnup and CRUD deposition in the operation process of the pressurized water reactor is described, and therefore the prediction precision of the reactor core axial power offset phenomenon is improved.

The invention discloses a burnable poison loading method for reducing tritium emission of a nuclear power plant and belongs to the field of nuclear power. The method comprises the following steps: providing a first type fuel assembly adopted by a reactor, providing a tritium-reduced burnable poison rod, and loading discrete water-through annular burnable poison or Gd-containing burnable poison in the tritium-reduced burnable poison rod; at least part of the integral burnable poison fuel rods in the first type of fuel assemblies are replaced with tritium-reduced burnable poison rods to obtain second type of fuel assemblies, iterative calculation is carried out to obtain second type of fuel assembly structures meeting safety design requirements, and the second type of fuel assemblies are used for fuel replacement. According to the method, the tritium-reducing burnable poison rod with longer burnup time is utilized, so that the requirement for soluble boron in a primary loop is reduced, and the effect of reducing tritium emission is achieved.

A computer-assisted method for determining an optimal core-loading pattern for a nuclear reactor core. Positions of nuclear fuel assemblies are tested to assign optimal positions and to load the reactor. The reactor core includes cells positioned symmetrically to axes of symmetry and a standard assembly for insertion into each cell. Standard assemblies are distributed by the number of previous production cycles. Groups of cell positions symmetric to the axes of symmetry are identified, and symmetric positions are counted. Families of standard assemblies having similar burnups are formed, wherein the standard assemblies correspond to positions in a group. The loading pattern of standard assemblies in initial positions is tested by numerical simulation, then the positions are swapped while maintaining the previously formed families of assemblies. The swapped positions loading pattern is tested by numerical simulation. This is repeated until at least one candidate pattern for loading the reactor is obtained.

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.

The invention relates to the technical field of research of reactor fuel irradiation, in particular to a fuel power calibration method for a loop irradiation test, which comprises the following steps: carrying out an out-of-reactor hydraulic test, and establishing a high-precision CFD (computational fluid dynamics) model of a fuel loop irradiation device based on a hydraulic test result; calculating the gamma heat release rate ratio of each material and the total heat release in the test section, and preliminarily determining the nuclear power value of the fuel assembly; based on the heat conduction differential equation and the convective heat transfer equation, the gamma heat release rate of the material and the nuclear power value of the fuel assembly are corrected; repeating the calculation process of S3 until the nuclear power value of the fuel assembly converges, and obtaining the final theoretical calculation value of the gamma heat release rate of each material and the nuclear power of the fuel assembly; and further verifying the calculation result to finally obtain the nuclear power of the fuel core body, and obtaining the burnup of the fuel assembly in the irradiation test through time integration. The method aims to solve the technical problem that the nuclear power of the fuel assembly calculated based on thermotechnical measurement data in a loop irradiation test has deviation compared with an actual value.

The invention relates to the technical field of combustors, in particular to an energy-saving spiral-flow type all-directional spraying combustor for a decomposing furnace, which comprises a furnace body, an installation and maintenance mechanism, an ignition mechanism and a rapid cut-off mechanism, the middle upper part of the furnace body is connected with a second pipeline through the installation and maintenance mechanism, and the inner wall of the bottom end of the first pipeline is fixedly connected with a flow dividing fan; first flow dividing openings are evenly formed in the bottom end of the first pipeline, and flow dividing plates are evenly and fixedly connected between the first pipeline and the third pipeline. According to the device, a user can conveniently conduct overhaul and maintenance work, air or fuel can be rapidly cut off, the use safety of the device can be improved, meanwhile, it can be guaranteed that the fuel is sprayed and combusted in all directions without dead corners, local high temperature in the furnace body is avoided, the fuel effect and the product yield can be improved, and the production cost is reduced. And meanwhile, the fuel consumption can be effectively reduced, so that the effects of saving energy, reducing consumption, reducing cost and improving efficiency are achieved.

The application discloses a method for selecting blast furnace charging grade, comprising the following steps: (1) fitting the relationship between blast furnace charging grade and ore consumption and fuel consumption; fitting the relationship between iron ore grade and price; (2) determining the proportion of furnace burden structure, the grade and price of pellet and lump ore, the ratio of coal, and the price of pulverized coal and coke; (3) based on the set blast furnace charging grade, calculating the grade of sinter, and the corresponding grade of mixed ore, calculating the price of mixed ore according to the relationship between iron ore grade and price, and then calculating the price of sinter; (4) calculating the ore consumption and fuel consumption according to the relationship between blast furnace charging grade and ore consumption and fuel consumption, and calculating the molten iron cost corresponding to the current blast furnace charging grade; (5) changing the blast furnace charging grade, executing steps (3)-(5), until the blast furnace charging grade TFe corresponding to the lowest molten iron cost T is found in the limited blast furnace charging grade change range, which is the optimal blast furnace charging grade.

The invention discloses a nuclear reactor protection setting value correction method and a computer readable storage medium. The method comprises the following steps: S1, establishing a nuclear reactor model and an intermediate range detector model; s2, a fuel assembly in a reactor core of the nuclear reactor model is selected for neutron transport simulation calculation, and an intermediate range response factor obtained after fission neutrons of the fuel assembly are transmitted to the position of the intermediate range detector model from a reactor core active area is obtained; s3, calculating an intermediate range theoretical current value at the beginning of the life period and an intermediate range theoretical current value at the current burnup point according to the intermediate range response factor, and obtaining a correction coefficient at the current burnup point; and S4, according to the correction coefficient under the current burn-up point, correcting the protection constant value of the middle range set at the beginning of the life period to obtain the protection constant value of the middle range under the current burn-up point. The method guarantees the accuracy of the protection constant value of the nuclear reactor, reduces the risk of mistaken shutdown of the nuclear reactor and the risk of protection failure of the reactor, and gives consideration to the safety of a unit.

The application discloses a kind of determination methods and program products of core reactivity state, the method includes: the key state parameter in the process of core reaction is collected, key state parameter includes at least one of core power distribution parameter, coolant temperature, coolant density, control rod position, boric acid concentration and burnup;Determine the burnup step of fuel, the evolution process of fissile nuclide and fission product in core is modeled based on the burnup step of fuel and burnup model, to obtain the reactivity parameter related to the burnup of fuel, the burnup model is used to describe the nuclide transmutation and consumption logic of fuel during the operation of core;Corresponding digital twin is constructed with the reactivity state of the core, the reactivity state of the core is mapped by the digital twin based on the reactivity parameter and the key state parameter driving, can realize high-precision, real-time, full life cycle dynamic digital mapping and intelligent detection of core reactivity state.

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.

The invention discloses a sensitivity coefficient calculation and fitting method for a self-powered neutron detector of a nuclear reactor, which comprises the following steps of: 1, performing component calculation under different burn-up depths and working condition parameter conditions by using a nuclear reactor component program to obtain neutron flux density and fuel component linear power density at the self-powered neutron detector under corresponding conditions; secondly, current signals generated by the self-powered neutron detector under different working condition parameters are calculated through a Monte Carlo program; thirdly, the relation between the linear power density of the fuel assembly and a current signal generated by the self-powered neutron detector is established, and the sensitivity coefficient of the self-powered neutron detector is obtained; fourthly, performing function fitting on the sensitivity coefficient of the self-powered neutron detector with respect to the burn-up depth and the working condition parameters to obtain a fitting coefficient; according to the method, the mapping relation between the current signal of the self-powered neutron detector of the nuclear reactor and the linear power density of the fuel assembly is effectively established, and key technical support is provided for online monitoring of the nuclear reactor.

The application provides a yield optimization method, system and medium for isotope irradiation production in a research reactor, and relates to the technical field of nuclear research reactors. The yield optimization method comprises the following steps: obtaining design data of the research reactor, irradiation hole usage data range and total required number of at least one type of isotope target; constructing a research reactor Monte Carlo model associated with the research reactor according to the design data; determining a flux influence factor and a multi-group microscopic cross section corresponding to the at least one type of isotope target, and a target-free neutron flux field function according to the research reactor Monte Carlo model; and taking, as a yield optimization result, layout data of each irradiation hole corresponding to a preset isotope yield target through burnup calculation according to the irradiation hole usage data range, the total required number of irradiation holes, the flux influence factor and the multi-group microscopic cross section corresponding to the at least one type of isotope target, and the target-free neutron flux field function.

The present invention discloses a burnup tracking calculation method, system and medium for multi-cycle fuel management; it relates to the technical field of reactor burnup calculation; after the first-cycle core modeling, geometric grid division is performed on each burnup zone unit in the full-core model. During the cycle of the full-core model, the replacement of all reactor internal components in the full-core model is achieved by exchanging the positions of the corresponding spatial numbers in the repeated geometric grid. Combining with the multi-cycle burnup tracking calculation process, the dynamic changes of fuel burnup, control rods and beryllium followers during reactor operation and their coupling relationship with the particle transport process are accurately considered, and the simulation of the actual refueling process of the research reactor based on the Monte Carlo method is completed, realizing high-precision and high-efficiency core calculation and analysis for high-flux research reactors, and providing theoretical support for aspects such as high-quality testing of nuclear fuels and materials, optimization of core loading and irradiation resources of research reactors, and improvement of the operation safety of research reactor reactors.

To solve such a problem that fuel balls are blocked when a fuel element runs in a dust and chipping environment is not fully considered.SOLUTION: A pressure-bearing unit comprises a pressure-bearing box body and an end flange, an allocation space is formed in the pressure-bearing box body, and the end flange is connected to the opening side of the pressure-bearing box body; a rotor unit comprises a rotating shaft, a material taking disc and a material distributing disc, and the material taking disc and the material distributing disc are connected to the rotating shaft in a sleeving mode and installed in the allocation space; a bush unit comprises a first bush and a second bush, and the first bush and the second bush are connected to the material taking disc and the material distributing disc in a sleeving mode. A non-return unit is sleeved on the rotating shaft and is connected with the end flange; the bearing assembly comprises a first bearing and a second bearing, and the first bearing and the second bearing are arranged in the end flange and the pressure bearing box body; a driving unit comprises a servo motor, a speed reducer, a coupler and a magnetic synchronizer which are connected in sequence.SELECTED DRAWING: Figure 1

The invention discloses a fuel loading structure of a reactor core balance cycle, a fuel management method and system and a computer storage medium, the fuel loading structure of the reactor core balance cycle comprises N1 first new fuel assemblies with the enrichment degree of 4.95% and N2 second new fuel assemblies with the enrichment degree of 4.45%, and N1 and N2 are related to the cycle length grade of the balance cycle; in the radial direction of the reactor core, the N1 first new fuel assemblies, the N2 second new fuel assemblies and the old fuel assemblies are mutually crossed and permeated, and the N1 first new fuel assemblies are arranged on the peripheries of the N2 second new fuel assemblies; the N1 first new fuel assemblies are in central symmetry in pairs relative to the original point of the reactor core; and the N2 second new fuel assemblies are in central symmetry in pairs relative to the original point of the reactor core. According to the technical scheme, not only is the circulation length flexible enough, but also the number of the new fuel assemblies entering the pile can be reduced, the discharging burnup is improved, and the economical efficiency is improved.

The invention discloses a conveyor belt type fuel ball transmission system and method for a high-temperature gas cooled reactor, a bottom outlet of a reactor core is communicated with an inlet of a broken ball separation device, a broken ball outlet of the broken ball separation device is communicated with a broken ball storage tank, a complete fuel ball outlet of the broken ball separation device is communicated with an inlet of a burn-up measurement device, and the burn-up measurement device is communicated with an outlet of the reactor core. A spent fuel outlet of the burnup measuring device is communicated with the spent fuel storage tank, a recycled fuel outlet of the burnup measuring device is communicated with an inlet of the discharging pipe, a plurality of conveying belts are arranged on the conveying pipeline, and when any conveying belt moves to the lowest position of the conveying pipeline, an outlet of the discharging pipe directly faces the conveying belt; and when any conveying belt moves to the top of the conveying pipeline, the conveying belt is over against the charging opening of the reactor core, and the system is high in operation stability.

The invention relates to the technical field of nuclear reactors, and provides a high-temperature gas cooled reactor system. The system comprises: a reactor; the combined converter is connected to an inlet of the reactor; the burnup measuring device is connected to an outlet of the reactor; the sipping analysis device is connected between the burnup measuring device and the combined converter, a detection cavity is formed in the sipping analysis device, and the sipping analysis device is used for receiving the fuel balls from the burnup measuring device and can establish a negative pressure environment in the detection cavity so as to detect the activity value of target radionuclide in the gas in the detection cavity; the control module is electrically connected with the combined converter, the burn-up measuring device and the sipping analysis device; when the burnup value is smaller than a set burnup threshold value and the activity value is smaller than an activity constraint value, the control module controls the combined converter to return the fuel balls to the reactor for recycling; when the burnup value is larger than or equal to a set burnup threshold value or the activity value is larger than or equal to an activity constraint value, the control module controls the combined converter to unload the fuel balls.

Some embodiments of the invention disclose a modeling method for a fission gas swelling behavior of UO2 fuel under a high-burn-up condition, and the rule of irradiation swelling caused by irradiation bubbles in a high-burn-up structure in the UO2 fuel is accurately determined by coupling grain refinement and bubble evolution. The method comprises the steps of constructing a total free energy density function; based on a recrystallization grain nucleation theory, establishing a grain refinement nucleation model; based on an Allen-Cahn equation, a phase field dynamics evolution equation of high-burn-up structure evolution is constructed; solving a total free energy density function and a phase field dynamics evolution equation, and carrying out visualization processing on a numerical solution; and based on the final morphology of the high-burnup structure obtained after visualization processing, determining the effective thermal conductivity, porosity and grain size in the high-burnup structure in the UO2 fuel, and determining the rule of irradiation swelling caused by irradiation bubbles in the high-burnup structure in the UO2 fuel.

The invention discloses an optimal selection method for the charging grade of a blast furnace. The optimal selection method comprises the following steps: (1) fitting a relational expression among the charging grade of the blast furnace, ore consumption and fuel consumption; fitting a relational expression of iron ore grade and price; (2) determining a furnace charge structure proportion, grades and prices of pellets and lump ores, a coal ratio and prices of pulverized coal and coke; (3) calculating the sinter grade and the corresponding blended ore grade based on the set blast furnace charging grade, calculating the price of the blended ore according to the relational expression of the iron ore grade and the price, and further calculating the sinter price; (4) calculating the ore consumption and the fuel consumption according to the relation between the blast furnace charging grade and the ore consumption and the fuel consumption, and calculating the molten iron cost corresponding to the current blast furnace charging grade; and (5) changing the charging grade of the blast furnace, and executing the steps (3)-(5) until the charging grade TFe of the blast furnace corresponding to the lowest molten iron cost T is found in the limited charging grade change range of the blast furnace, namely the preferable charging grade of the blast furnace.

A multi-physics coupling system and method for a microreactor includes: a neutron physics simulation module, a thermal-hydraulic simulation module, a power conversion module, a main function module, a burnup nuclide update module, and a control drum position update module. The present invention dynamically adjusts the position of the control drum via the control drum position update module during the burnup coupling calculation process. In conventional multi-physics coupling methods, the position of the control drum is fixed, so the effective proliferation factor of the core is always greater than 1.0, and the results of the coupling calculation cannot reflect the actual state of the reactor during operation. When the multi-physics coupling method of the present invention is used for calculation, by continuously adjusting the control drum position, the effective proliferation factor of the reactor is always near 1.0, which is closer to the operating conditions of a real reactor and can more accurately simulate the state of the reactor.

The invention belongs to the technical field of nuclear reactor physics, and particularly relates to a fast and fine burnup calculation method for dispersive fuel based on reaction rate equivalence. Comprising the following steps: step 1, merging multi-group parameters; step 2, equivalently homogenizing the region; step 3, calculating burnup; and step 4, restarting calculation. The dispersion type fuel burn-up calculation method based on the reaction rate equivalence has the advantages that the efficiency of dispersion type fuel burn-up calculation is remarkably improved, meanwhile, high calculation precision is kept, and burn-up information of all layers in fuel particles is kept. The method provides an efficient and accurate calculation tool for design and safety analysis of the dispersion type fuel assembly.

The present invention discloses a low-temperature steel electrode for the base material SA-350Gr.LF3 of a high-burnup spent fuel storage and transportation container and a preparation method thereof. The electrode is obtained by wrapping a welding core with a powder. The composition of the powder is as follows in parts by weight: 4.5-10 parts of electrolytic manganese, 1-5 parts of titanium-boron alloy, 3-12 parts of sodium fluoride, 6-12 parts of nickel powder, 15-30 parts of calcium fluoride, 4-10 parts of silica fume, 10-15 parts of iron powder, 1-2.5 parts of silicon-zirconium alloy, and 15-35 parts of marble. The composition of the welding core is as follows in weight percentage: 0.02%-0.2% of carbon, 0.1%-0.4% of manganese, 0.01%-0.03% of silicon, 99.5%-99.9% of iron, and the balance is inevitable impurities. The electrode of the present invention has good welding process, and the mechanical properties, impact properties and crack resistance of the weld are good.

The invention discloses a control test method and system for a high-temperature gas cooled reactor burn-up measurement system, and the method comprises the steps: adjusting an energy scale coefficient in a current environment through employing a self-adaptive energy scale algorithm after the measurement system is started; an edge cloud architecture is established to perform data acquisition and processing, measurement points independently acquire burnup data, key parameters are extracted and then uploaded to a central system, and pattern recognition and trend prediction are performed; and carrying out online calibration on the measurement system, recording all measurement and calibration data, and generating a detailed report. According to the method, the energy scale coefficient is adjusted through the self-adaptive energy scale algorithm, data acquisition and processing of the edge cloud architecture and online calibration and report generation are performed, so that comprehensive control and test of the high-temperature gas cooled reactor burnup measurement system are realized, the accuracy and reliability of the system are improved, real-time monitoring and analysis of burnup data are realized, and the real-time monitoring and analysis of the burnup data are realized. And important data support is provided for safe operation and maintenance of the nuclear power station.

A fuel design for high burnup in water-cooled reactors is disclosed. A nuclear fuel rod comprising an outer zone of annular fuel pellets composed of actinides and fissile material and an interior comprising a low density, high open porosity ceramic foam of zirconia or magnesia is used. Use of annular pellets lowers fuel temperature, reduces fission gas release and fuel swelling. Foam with low crushing strength is used to accommodate fuel swelling. Fresh fuel reactivity is controlled by burnable absorbers such as gadolinia, incorporated as sintered bodies within the foam or as part of the fuel composition or both. The placement of burnable absorber in the foam reduces the contact area with surrounding media, mitigating diffusional transport and preserving its self-shielding characteristic. A novel two-piece foam holder is proposed to insert the ceramic foam.

To enable a weight and degree of burning of a nuclear fuel contained in an object to be measured appropriately even when advance information on the object is a few.SOLUTION: A nuclear fuel measurement system has: a muon trajectory detector 11 that is provided in a facing position across an object 1, and measures a muon μ to detect a muon trajectory; a radiation detector 12 that is installed in the vicinity of the object, and measures a radioactive ray such as a gamma ray and the like to be radiated from the object; a trajectory determination unit 26 that reconstructs the muon trajectory detected by the muon trajectory detector, and determines whether the muon trajectory transmits the radiation detector; a nuclear fuel weight determination unit 27 that selects an event in which the muon trajectory does not transmit the radiation detector, and derives a weight of a nuclear fuel in the object on the basis of a measurement value of the muon trajectory detector; and a degree-of-burning determination unit 28 that selects an event in which the muon trajectory does not transmit the radiation detector, and derives a degree of burning of the nuclear fuel in the object on the basis of each measurement value of the muon trajectory detector and radiation detector.SELECTED DRAWING: Figure 1