113 results about "Burnup" patented technology

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 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 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 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 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 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.

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.

The application discloses a fusion reactor liquid blanket tritium breeding rate online regulation system simulation method and device, wherein the simulation device comprises a fusion blanket neutron transport module, a blanket breeding agent Monte Carlo-activation coupling burnup simulation module, a blanket coolant feeding module, a blanket tritium extraction module, a blanket coolant purification treatment module, a blanket tritium breeding rate calculation module and a blanket tritium breeding agent circulation system; the simulation device realizes the circulation of the blanket coolant nuclide based on the mass conservation equation and the burnup equation. The application is based on the Boltzmann neutron transport equation, the neutron flux of the blanket coolant is simulated by a Monte Carlo method, the time and space variation law of the chemical composition of the system material is obtained by solving the burnup equation, the replenishment demand of the blanket coolant system is calculated, and the influence of the replenishment composition on the blanket tritium breeding rate is simulated, so that the application can be widely applied to the fusion reactor liquid lithium lead blanket.

The application discloses a metal fuel irradiation test physical thermal analysis method, device, equipment and medium, relates to the field of nuclear reactor physics and thermal engineering, and comprises the following steps: determining the burnup depth information of the metal fuel according to the measured thermal power, inputting the burnup depth information into a pre-configured structure swelling model for analysis, obtaining the volume swelling rate of the metal fuel in the axial direction and the radial direction, updating the structure size of the metal fuel according to the volume swelling rate, and updating the thermal conductivity of the metal fuel according to the updated structure size; according to the burnup depth information, the updated structure size of the metal fuel and the burnup distribution situation calculated by the burnup management calculation model based on the determinism method, the burnup distribution curve of the metal fuel at the corresponding irradiation time is refitted, the heat release calculation is carried out according to the burnup distribution curve, and the heat release distribution information of the metal fuel is obtained; the heat release distribution information of the metal fuel, the updated structure size and the thermal conductivity of the metal fuel are input into a sub-channel analysis model for thermal analysis, and the key thermal parameters of the metal fuel are obtained.

The invention provides a reactor control method based on a rapid power reduction system, electronic equipment and a medium, and relates to the technical field of nuclear reactors. The rapid power reduction system comprises a plurality of control rods for rapid power reduction, each control rod comprises a grey rod group, and the reactor control method based on the rapid power reduction system comprises the following steps: acquiring a corresponding relation between an insertion depth limit value of the control rod and reactor core burnup; and controlling the operation rod position of the control rod not to exceed the insertion depth limit value corresponding to the current burn-up step according to the corresponding relation. By limiting the insertion depth of the operation rod position of the control rod for rapid power reduction, it is ensured that the rapid power reduction system has enough rod value to fall into the reactor core, the operation complexity is reduced, and the operation safety of the nuclear power plant is improved.

The present disclosure relates to the technical field of reactor fuel management, and particularly relates to a nuclear fuel management method of a multi-module pebble bed high temperature gas cooled reactor. The method comprises: obtaining first fuel management information in a first module initial loading and a transition process according to first input information corresponding to the multi-module pebble bed high temperature gas cooled reactor; obtaining second fuel management information in a second module initial loading and a transition process according to low enrichment shallow burnup spent fuel element information corresponding to the first module and second input information corresponding to the second module; obtaining third fuel management information in a fourth module initial loading and a transition process according to low enrichment shallow burnup spent fuel element information corresponding to at least one third module and second input information corresponding to the fourth module; and executing the fuel management information set according to an evaluation result of the fuel management information set in the process of executing the fuel management information set. The present disclosure can improve the accuracy of nuclear fuel management and improve the utilization rate of fuel elements.

The invention is applicable to the technical field of nuclear reactors, and provides a xenon reactivity calculation method, device, equipment and product for a nuclear reactor. The method comprises the following steps: acquiring current set parameters for processing furnace charge in the nuclear reactor and multiple groups of optimal xenon information corresponding to the nuclear reactor; the furnace charge in the nuclear reactor generates xenon in the fission reaction process; according to the multiple sets of optimal xenon information, target xenon parameters of the nuclear reactor processed under the current furnace charge circulation parameters and the current furnace charge burnup parameters are determined; the xenon reactivity generated by the charge in the nuclear reactor is calculated based on the target xenon parameter. According to multiple groups of optimal xenon information corresponding to the nuclear reactor, the target xenon parameters processed by the nuclear reactor under the current furnace charge circulation parameters and the current furnace charge burnup parameters are determined, and the xenon reactivity is calculated according to the target xenon parameters, so that the accuracy of the xenon reactivity is improved; and an effective reference basis is provided for an operator to manage the nuclear reactor.

A long-cycle, high-burnup fuel management method based on accident-tolerant fuel, belonging to the nuclear power field, is proposed. This method targets pressurized water reactors comprising 157 fuel assemblies and includes the following steps: providing a replacement group, comprising a first-class fuel assembly and a second-class fuel assembly using accident-tolerant fuel with an enrichment of not less than 5.5%, wherein the first-class fuel assembly includes IFBA combustible poison rods, and the second-class fuel assembly includes IFBA combustible poison rods and Gd-containing combustible poison rods; during refueling, one fuel assembly and 68 remaining reserve fuel assemblies with the lowest reactivity are removed from the core center, and the remaining reserve fuel assembly with the highest reactivity from the spent fuel pool is loaded into the core center and placed into the replacement group. This method enables a long-cycle refueling of 24 months, effectively flattening core power and improving the economics of nuclear power plants.

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.

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

Thorium-based fuel bundles are used in existing PHWR reactors (e.g., Indian 220 MWe PHWR, Indian 540 MWe PHWR, Indian 700 MWe PHWR, CANDU 300 / 600 / 900) in place of conventional uranium-based fuel bundles, with little or no modifications to the reactor. The fuel composition of such bundles is 60+ wt % thorium, with the balance of fuel provided by low-enriched uranium (LEU), which has been enriched to 13-19.95% 235U. According to various embodiments, the use of such thorium-based fuel bundles provides (1) 100% of the nominal power over the entire life cycle of the core, (2) high burnup, and (3) non-proliferative spent fuel bundles having a total isotopic uranium concentration of less than 12 wt%. Reprocessing of spent fuel bundles is also avoided.

The application discloses a kind of special-shaped fuel geometric nested neutron transport and burnup calculation method and system, method includes: based on geometric nested model, the neutron transport calculation of special-shaped reactor is realized;Geometric nested model is obtained by constructing as follows: using the modeling of conventional assembly to construct entity geometry method, using direct acceleration geometry Monte Carlo method to model special-shaped fuel assembly, the geometry system space of direct acceleration geometry Monte Carlo geometry is embedded into the geometry system space of construct entity geometry, complete geometric nested modeling;After neutron transport calculation is completed, the burnup of geometric nested model is calculated, and the burnup calculation of special-shaped fuel reactor is realized.The application can improve the problem that Monte Carlo calculation cannot accurately and efficiently establish special-shaped fuel model using CSG method, improve the calculation efficiency using DAGMC method, and accurately and efficiently complete special-shaped fuel reactor core modeling and neutron transport calculation and burnup calculation.

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 invention provides a Gaussian-Seidel-based MMPA burnup equation set accelerated solving method, belongs to the field of reactor physical calculation, and aims to solve the technical problems of high solving complexity, long calculation time and low efficiency of an existing MMPA burnup equation set. The method comprises the following steps: firstly, based on current reactor burnup working condition parameters, accurately constructing a burnup matrix under the current burnup condition through nuclear data processing and burnup chain modeling; the method comprises the steps that firstly, an MMPA burnup equation is constructed, an MMPA burnup equation set suitable for Gaussian Seidel iteration solution is obtained through derivation, finally, an iteration convergence criterion is set, and the Gaussian Seidel iteration method is adopted for conducting iteration solution on the MMPA burnup equation set till the convergence condition is met; and finally, outputting a nuclide density distribution result at a corresponding moment. According to the method, the calculation efficiency can be greatly improved on the premise that the calculation precision is guaranteed, and the problems that a traditional MMPA burnup equation is high in solving complexity and long in calculation time consumption are solved.

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.

The application discloses a kind of thermal-mechanical-fission product diffusion coupling methods of coated particle dispersion fuel pellets, steps as follows:1, set up and establish random distribution of coated particle dispersion fuel three-dimensional model;2, determine and set respectively applicable to complete coated particle and failure coated particle heat transfer module, mechanics module, diffusion module and auxiliary model;3, carry out the multi-physics field coupling performance analysis calculation of fuel pellet;4, extract and store the stress of all coated particles in fuel pellet and determine the failure state of coated particle;5, count the number of failure coated particles and change the failure coated particle calculation module;6, repeat steps 3-5 to perform fuel performance analysis calculation until the end of burnup.The application can accurately obtain the temperature field, stress-strain field, fission product concentration distribution, fission product release amount and failure particle number of dispersion fuel pellets.

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 invention discloses a model training method and device, an analysis method and device, related equipment, a storage medium and a computer program product. The method comprises the steps that multiple sets of sample data are obtained, an initial model is constructed, and each set of sample data in the multiple sets of sample data is related to burnup, initial uranium concentration, cooling time, curium-244 content and plutonium content corresponding to a spent fuel assembly; the initial model is trained through the multiple sets of sample data, a first model is obtained, and the first model is at least used for determining the plutonium content corresponding to the spent fuel assembly based on the burnup, the initial uranium concentration, the cooling time and the curium-244 content corresponding to the spent fuel assembly. The first model obtained through training can be used for accurately determining the plutonium content in the spent fuel assembly.

This invention pertains to fuel elements for nuclear engineering and the VVER-1200 type water-hydrodynamic reactor, and relates to improvements in the structure of VVER-1000 type fuel elements. Essentially, it modifies the structure of the bottom plug, increases the length of the fuel elements, and the length and mass of the fuel column, thereby increasing the reactor's thermal power to 3300 MW. The reactor core, composed of fuel assemblies, is manufactured taking into account the increased thermal power of the reactor and the increased height of the fuel elements within the fuel assemblies, and can be used within a reactor vessel with a geometry similar to that of a typical VVER-1000 type reactor vessel. As a result, fuel burnup is increased, thereby increasing the reactor's thermal power to 3300 MW, while maintaining the size of the VVER-1000 type reactor vessel and previous safety levels, while simultaneously increasing the overall power generation of the reactor.