Method and device for determining thermal safety margin of reactor core

Abstract

This application provides a method and apparatus for determining the thermal safety margin of a reactor core, belonging to the field of nuclear power technology. The method includes: acquiring initial information of various fuel assemblies in the reactor core, the initial information including the geometric characteristics of the mixing grid of the fuel assemblies and test results obtained from critical heat flux density tests on the fuel assemblies; determining a target critical heat flux density relationship corresponding to the reactor core based on the initial information of the various fuel assemblies; and calculating the thermal safety margin of the reactor core based on the target critical heat flux density relationship to obtain the target thermal safety margin. This application can solve the problem of low accuracy in the calculation of thermal safety margin of reactor cores in the prior art, fully considering the characteristic differences of different types of fuel assemblies, and improving the accuracy of thermal safety margin analysis results.

Description

Technical Field

[0001] This application relates to the field of nuclear power technology, and in particular to a method and apparatus for determining the thermal safety margin of a reactor core. Background Technology

[0002] In the field of nuclear reactor thermal safety analysis, hybrid core design has attracted increasing attention due to its combination of different types of fuel assemblies to optimize performance. Current technologies for thermal safety analysis of hybrid cores are primarily based on a universal critical heat flux (CHF) formula, applicable to core systems composed of fuel assemblies with similar critical heat flux characteristics. However, when a hybrid core contains different types of fuel assemblies with significantly different critical heat fluxes, using a uniform CHF formula for safety analysis may lead to inaccurate evaluation results or a lack of sufficient conservatism.

[0003] Existing methods typically do not adequately consider the differences in critical heat flux density performance among different types of fuel assemblies, thus failing to accurately reflect actual thermal boundary conditions and resulting in low accuracy in calculating thermal safety margins. Summary of the Invention

[0004] The main objective of this application is to propose a method and apparatus for determining the thermal safety margin of a reactor core, aiming to solve the problem of low accuracy in the calculation of thermal safety margin of hybrid reactor cores in the prior art and improve the accuracy of thermal safety margin analysis results.

[0005] To achieve the above objectives, a first aspect of this application proposes a method for determining the thermal safety margin of a reactor core, the method comprising: The initial information of various fuel assemblies in the reactor core is obtained. The initial information includes the mixing grid geometry of the fuel assembly and the test results obtained by conducting a critical heat flux density test on the fuel assembly. The mixing grid geometry of the fuel assembly includes at least one of the following: whether the fuel assembly includes a mixing blade, the thickness of the fuel assembly, the height of the fuel assembly, and the pressure drop coefficient of the fuel assembly. Based on the initial information of various fuel assemblies, the target critical heat flux density relationship corresponding to the reactor core is determined; The thermal safety margin of the reactor core is calculated based on the target critical heat flux density relationship to obtain the target thermal safety margin.

[0006] In some embodiments, determining the target critical heat flux density relationship corresponding to the reactor core based on initial information of various fuel assemblies includes: If multiple fuel assemblies in the reactor core meet the first preset condition, then one critical heat flux density relationship is selected from the critical heat flux density relationships corresponding to the multiple fuel assemblies as the target critical heat flux density relationship. If multiple fuel assemblies in the reactor core do not meet the first preset condition, the test results corresponding to the multiple fuel assemblies are merged to obtain multiple sets of merged data, and the target critical heat flux density relationship is constructed based on the multiple sets of merged data. The first preset condition includes: Each of the fuel assemblies includes a turbulence wing; Each of the aforementioned fuel assemblies does not include a turbulence fin; The difference in thickness between all the fuel assemblies is less than a first preset threshold; The difference in height between all the fuel assemblies is less than a second preset threshold; The difference in pressure drop coefficients among all the fuel assemblies is less than a third preset threshold; The difference between the test results for all the fuel assemblies is less than a fourth preset threshold.

[0007] In some embodiments, merging the test results corresponding to multiple fuel assemblies to obtain multiple sets of merged data includes: Based on the grid geometry of the fuel assembly and the axial heating method of the fuel assembly, multiple fuel assembly groups are obtained. The grid geometry includes typical grids or guide tube grids, and the axial heating method includes uniform heating or non-uniform heating. For multiple fuel component groups, multiple sets of merged data are obtained based on the test results corresponding to the fuel components in the same group.

[0008] In some embodiments, constructing the target critical heat flux density relationship based on multiple sets of merged data includes: Based on the variation patterns of the test results of multiple sets of combined data, an initial expression for the critical heat flux density relationship corresponding to the reactor core is determined. The variation patterns include the trend of critical heat flux density with mass flux density, steam content, pressure, or heating length. By fitting the coefficients in the initial expression, the target critical heat flux density relationship corresponding to the reactor core is obtained.

[0009] In some embodiments, after constructing the target critical heat flux density relationship based on multiple sets of merged data, and before calculating the thermal safety margin of the reactor core based on the target critical heat flux density relationship to obtain the target thermal safety margin, the method further includes: Acquire multiple sets of thermal parameters of the reactor core at multiple acquisition times during operation; Based on each set of thermal parameters, the critical heat flux density corresponding to each set of thermal parameters is calculated using the target critical heat flux density relationship. The maximum value among the obtained critical heat flux densities is taken as the upper limit of the critical heat flux density, wherein the upper limit is used to limit the critical heat flux density calculated according to the target critical heat flux density formula.

[0010] In some embodiments, the step of calculating the thermal safety margin of the reactor core based on the target critical heat flux density relationship to obtain the target thermal safety margin includes: For each of the fuel assemblies, based on the target critical heat flux density relationship, predicted critical heat flux density values ​​for the fuel assembly are calculated under various test conditions. These predicted critical heat flux density values ​​include predicted values ​​for the critical heat flux density at the minimum deviation from the nucleus-boiling ratio point and predicted values ​​for the critical heat flux density at the minimum burnout point; and... For each test condition, the following processing is performed: Based on the predicted critical heat flux density of the fuel assembly under various test conditions, the measured critical heat flux density at the minimum deviation from the nucleus-boiling ratio point and the measured critical heat flux density at the minimum burnout point in the test results corresponding to the fuel assembly, the ratio of the predicted value to the measured value at the minimum deviation from the nucleus-boiling ratio point and the ratio of the predicted value to the measured value at the minimum burnout point are obtained under the test condition; If each test condition meets the second preset condition, then the thermal safety margin of the reactor core is calculated according to the target critical heat flux density relationship. The second preset condition is that the ratio of the predicted value to the measured value of the minimum deviation from the nucleus boiling ratio point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level, and the ratio of the predicted value to the measured value of the minimum burn-out point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level.

[0011] In some embodiments, the method further includes: If any test condition does not meet the second preset condition, the target critical heat flux density relationship is adjusted according to the preset strategy to obtain a new critical heat flux density relationship. The new critical heat flux density relationship is then used as the target critical heat flux density relationship. The process then jumps to the step of calculating the critical heat flux density corresponding to each set of thermal parameters based on the target critical heat flux density relationship. The preset strategy includes at least one of the following: removing test results that deviate from the normal distribution from the test results of various fuel assemblies and changing the initial expression of the critical heat flux density relationship corresponding to the reactor core.

[0012] In some embodiments, the step of calculating the thermal safety margin of the reactor core based on the target critical heat flux density relationship to obtain the target thermal safety margin includes: Obtain the penalty value for the pressure drop coefficient of various fuel assemblies; Based on the target critical heat flux density relationship and the pressure drop coefficient penalty value, a thermal safety margin analysis is performed to obtain the target thermal safety margin.

[0013] In some embodiments, obtaining the penalty value for the pressure drop coefficient of various fuel assemblies includes: Obtain the pressure drop coefficient uncertainty value for each of the various fuel assemblies, and the pressure drop coefficient uncertainty value is used to quantify the prediction error range of the pressure drop coefficient for each of the fuel assemblies; The first value is obtained by calculating the sum of squares of the uncertainty values ​​of the pressure drop coefficient for each fuel assembly; Calculate the square root of the first value to obtain the penalty value of the pressure drop coefficient.

[0014] To achieve the above objectives, a second aspect of this application provides a device for determining the thermal safety margin of a reactor core, the device comprising: The acquisition module is used to acquire initial information of various fuel assemblies in the reactor core. The initial information includes the mixing grid geometry of the fuel assembly and the test results obtained by conducting a critical heat flux density test on the fuel assembly. The mixing grid geometry of the fuel assembly includes at least one of the following: whether the fuel assembly includes mixing blades, the thickness of the fuel assembly, the height of the fuel assembly, and the pressure drop coefficient of the fuel assembly. The determination module is used to determine the target critical heat flux density relationship corresponding to the reactor core based on the initial information of various fuel assemblies. The calculation module is used to calculate the thermal safety margin of the reactor core based on the target critical heat flux density relationship, and obtain the target thermal safety margin.

[0015] The method and apparatus for determining the thermal safety margin of a reactor core proposed in this application acquire initial information on various fuel assemblies in a hybrid reactor core. This initial information includes the geometric characteristics of the mixing grid of the fuel assemblies (such as whether they include mixing vanes, thickness, height, and pressure drop coefficient, at least one of these), and test results obtained from critical heat flux density tests on the fuel assemblies. Based on the initial information of the various fuel assemblies, a target critical heat flux density relationship is determined for the hybrid reactor core. The thermal safety margin of the hybrid reactor core is calculated based on the target critical heat flux density relationship to obtain the target thermal safety margin. The embodiments of this application significantly improve the accuracy and applicability of the thermal safety margin calculation for hybrid reactor cores, ensuring the conservatism of the thermal safety evaluation of hybrid reactor core fuel assemblies. It can be effectively applied to engineering practice of thermal safety margin analysis of hybrid reactor core fuel assemblies, providing reliable technical support for the safe operation of hybrid nuclear reactor cores. Attached Figure Description

[0016] Figure 1 This is a flowchart illustrating the method for determining the thermal safety margin of a reactor core provided in an embodiment of this application. Figure 2 This is a logic block diagram of the method for determining the thermal safety margin of a reactor core provided in the embodiments of this application; Figure 3 This is a schematic diagram of the structure of the device for determining the thermal safety margin of a reactor core provided in an embodiment of this application. Detailed Implementation

[0017] To make the objectives, technical solutions, and advantages of this application clearer, the following detailed description is provided in conjunction with the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are merely illustrative and not intended to limit the scope of this application.

[0018] It should be noted that although functional modules are divided in the device schematic diagram and a logical order is shown in the flowchart, in some cases, the steps shown or described may be performed in a different order than the module division in the device or the order in the flowchart. The terms "first," "second," etc., in the specification, claims, and the aforementioned drawings are used to distinguish similar objects and are not necessarily used to describe a specific order or sequence.

[0019] Unless otherwise defined, all technical and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the art to which this application belongs. The terminology used herein is for the purpose of describing embodiments of this application only and is not intended to limit this application.

[0020] First, let's analyze some of the terms used in this application: Critical Heat Flux (CHF): When the heat flux density reaches the value corresponding to the transition from nucleus boiling to transition boiling, there are numerous bubbles on the heating surface, causing many bubbles to merge and cover part of the heating surface. Due to the low heat transfer coefficient of the gas film, the temperature of the heating surface will rise rapidly, causing the heating surface to burn out. This critical point is also called the boiling critical point or critical heat flux density.

[0021] Departure from Nucleate Boiling Ratio (DNBR) is a key indicator in nuclear reactor safety design used to quantitatively prevent departure from nucleate boiling (DNB) in fuel elements. It is defined as the ratio of the critical heat flux density to the actual heat flux density. This parameter is continuously verified during the commissioning and operation of nuclear power plants to ensure that the heat removed by the coolant is always greater than the heat released by the fuel, preventing the cladding temperature from exceeding safety limits. The safety limit for DNBR must meet the 95 / 95 DNB criterion (i.e., no DNB will occur with a 95% probability and a 95% confidence level).

[0022] Burnout Point (BO): This refers to the critical point in the rod bundle channel where excessively high heat flux density leads to localized deterioration of heat transfer and a rapid increase in wall temperature. In experiments, the BO point is usually determined by thermocouple temperature measurement or visualization methods (such as high-speed photography), and it manifests as a dry surface or blockage of air bubbles on the heating rod.

[0023] In the field of nuclear reactor core thermal-hydraulic safety analysis, especially for hybrid core designs containing multiple fuel assembly types, accurate prediction and assessment of the critical heat flux (CHF) is a crucial step in ensuring the safe operation of the core. Existing technologies primarily focus on using universal CHF relationships for thermal safety analysis of hybrid cores. These universal relationships are typically developed based on specific or standardized fuel assembly geometry (such as rod diameter, grid spacing, and positioning grid design) and operating conditions. Their applicability is good when analyzing hybrid cores composed of fuel assemblies with similar or minor differences in CHF characteristics, providing relatively reliable safety assessment results.

[0024] However, with the diversification of advanced fuel assembly designs and core layouts, hybrid cores composed of different types of fuel assemblies with significantly different critical heat flux density characteristics (e.g., assemblies with different rod bundle structures, positioning grid designs, or cladding materials) are increasingly appearing in practical engineering. For such cores, existing analytical methods relying on a single universal CHF relationship have significant limitations.

[0025] Universal formulas are insufficient to accurately capture the significantly different CHF behaviors exhibited by different types of fuel assemblies under specific geometric characteristics (especially the influence of the complex structure of the mixing grid on flow and heat transfer) and operating parameters. Directly applying universal formulas may lead to CHF predictions that are too high or too low for certain assembly regions, failing to accurately reflect local thermal-hydraulic conditions.

[0026] Due to biases in CHF predictions, the Deviation Nucleus Boiling Ratio (DNBR) calculated based on the general formula may not accurately reflect the true margin at the most dangerous points in the core. This could lead to insufficient DNBR margin assessment for certain high-risk areas (such as components with poor CHF characteristics) or overly conservative assessment for low-risk areas in safety analyses, affecting the accuracy of core economic evaluations. Current technology lacks a reasonable method for determining DNBR design limits for components with varying CHF characteristics in hybrid cores.

[0027] Existing technologies have failed to fully consider the impact of differences in CHF characteristics of different components within a hybrid reactor core on the overall safety assessment, and have also failed to provide an effective and systematic processing strategy to integrate this difference information in order to ensure the conservatism and reliability of the final safety assessment results.

[0028] Based on this, embodiments of this application provide a method and apparatus for determining the thermal safety margin of a reactor core, aiming to provide a method that can effectively address the thermal safety analysis of mixed reactor cores composed of fuel assemblies with significantly different critical heat flux density characteristics. Embodiments of this application can fully consider the characteristic differences of different types of fuel assemblies, reasonably determine the DNBR design limit, and formulate scientific processing strategies to ensure the accuracy and conservatism of the thermal safety evaluation of mixed reactor cores.

[0029] The method and apparatus for determining the thermal safety margin of a reactor core provided in this application are specifically described through the following embodiments. First, the method for determining the thermal safety margin of a reactor core in this application is described.

[0030] The method for determining the thermal safety margin of a reactor core provided in this application relates to the field of nuclear power technology. This method can be applied to a terminal, a server, or software running on either a terminal or a server. In some embodiments, the terminal can be a smartphone, tablet, laptop, desktop computer, etc.; the server can be configured as an independent physical server, a server cluster or distributed system composed of multiple physical servers, or a cloud server providing basic cloud computing services such as cloud services, cloud databases, cloud computing, cloud functions, cloud storage, network services, cloud communication, middleware services, domain name services, security services, CDN, and big data and artificial intelligence platforms; the software can be an application implementing the method for determining the thermal safety margin of a reactor core, but is not limited to the above forms.

[0031] This application can be used in a wide variety of general-purpose or special-purpose computer system environments or configurations. Examples include: personal computers, server computers, handheld or portable devices, tablet devices, multiprocessor systems, microprocessor-based systems, set-top boxes, programmable consumer electronics, network PCs, minicomputers, mainframe computers, and distributed computing environments including any of the above systems or devices. This application can be described in the general context of computer-executable instructions executed by a computer, such as program modules. Generally, program modules include routines, programs, objects, components, data structures, etc., that perform specific tasks or implement specific abstract data types. This application can also be practiced in distributed computing environments where tasks are performed by remote processing devices connected via a communication network. In distributed computing environments, program modules can reside in local and remote computer storage media, including storage devices.

[0032] Figure 1 This is a flowchart of the method for determining the thermal safety margin of a reactor core provided in an embodiment of this application. Figure 1 The method may include, but is not limited to, steps S100 to S300.

[0033] Step S100: Obtain initial information of multiple fuel assemblies in the hybrid reactor core. The initial information includes the mixing grid geometry of the fuel assembly and the test results obtained from the critical heat flux density test of the fuel assembly. The mixing grid geometry of the fuel assembly includes at least one of the following: whether the fuel assembly includes a mixing blade, the thickness of the fuel assembly, the height of the fuel assembly, and the pressure drop coefficient of the fuel assembly.

[0034] In this embodiment, the hybrid reactor core consists of multiple fuel assemblies. The structures (mixing grids) and thermal response (CHF) of different assemblies may differ. Directly applying a single CHF formula would lead to CHF prediction bias, thereby affecting the accuracy of thermal safety margin calculation. To analyze the safety margin of the hybrid reactor core, it is necessary to obtain the mixing grid geometry characteristics and critical heat flux density test results for each fuel assembly.

[0035] Specifically, the mixing grid is a key structural component of the fuel assembly. Its main function is to disrupt the flow of coolant, promoting the mixing of hot and cold fluids, thereby significantly improving heat transfer efficiency and preventing localized overheating. Its geometric characteristics directly affect its mixing capability, which in turn affects the flow state of the coolant (such as disturbance intensity and flow channel resistance) and heat transfer efficiency, thus altering the CHF value. The geometric characteristics of the mixing grid include whether it integrates mixing blades, its thickness, its installation height within the fuel assembly, and its pressure drop coefficient. Among them, the turbulence blades are specially designed blades on the grid to generate vortices and enhance the mixing effect. Components with turbulence blades have stronger coolant disturbance and higher heat transfer efficiency. The thickness of the grid affects the flow resistance and the intensity of mixing. Different installation heights of the mixing grid in the fuel assembly will result in different axial disturbance ranges of the coolant. If the height difference is large, the heat transfer conditions (such as coolant vapor content and temperature) at different axial positions of each component will have low matching degree, and the CHF distribution pattern will also be different. The pressure drop coefficient reflects the magnitude of the resistance encountered by the coolant when flowing through the fuel assembly. Different coefficients mean that under the same flow rate, there are differences in coolant flow velocity and pressure loss in each component, and flow velocity is a key parameter affecting CHF.

[0036] Critical heat flux (CHF) is the most critical parameter for thermal safety in nuclear reactors. When the surface heat flux of the fuel rods exceeds CHF, heat transfer deteriorates drastically, causing the fuel rod temperature to spike and posing a risk of burnout. Therefore, a higher CHF value is better, indicating a larger safety margin. CHF testing: Each newly designed fuel assembly must undergo CHF testing (such as a 5x5 rod bundle test in a high-temperature, high-pressure loop) to measure its CHF value under different operating conditions (such as different pressures, flow rates, and temperatures). The critical heat flux relationship is an empirical formula established based on a large amount of experimental data and through mathematical fitting methods, which can predict the CHF value of that fuel assembly under any given operating condition. Each type of fuel assembly typically has its own proprietary, validated CHF relationship.

[0037] By comparing the geometric characteristics of the mixing grid and the CHF test results, it can be determined whether the hydraulic and thermal safety characteristics of different fuel assemblies in the mixed core are similar. When the geometric characteristics of the mixing grid and the CHF test results of all fuel assemblies in the mixed core are basically the same, the CHF relationship of any fuel assembly can be used as the general CHF relationship for subsequent thermal safety margin analysis. When the geometric characteristics of the mixing grid and the geometric characteristics of the CHF test results of all fuel assemblies in the mixed core are significantly different, the CHF relationship needs to be refitted.

[0038] Step S200: Determine the target critical heat flux density relationship corresponding to the hybrid reactor core based on the initial information of various fuel assemblies.

[0039] In this embodiment, the target critical heat flux density relationship is determined based on initial information, depending on the specific circumstances. If the mixing grid geometry and critical heat flux density test results of various fuel assemblies are basically the same, the critical heat flux density relationship of any fuel assembly is selected as the target critical heat flux density relationship, and a pressure drop coefficient penalty factor is introduced to handle pressure drop differences. If there are significant differences between the mixing grid geometry and critical heat flux density test results of various fuel assemblies, the test data of each fuel assembly are grouped according to grid type (e.g., typical grid, guide tube grid) and heating method (uniform heating, non-uniform heating). Based on the grouped data, a general critical heat flux density relationship is refitted as the target critical heat flux density relationship, and the critical heat flux density design limit (upper limit) is determined.

[0040] Step S300: Calculate the thermal safety margin of the hybrid reactor core according to the target critical heat flux density relationship to obtain the target thermal safety margin.

[0041] In this embodiment, key thermal parameters under actual operation or accident conditions of the hybrid reactor core are obtained, including but not limited to: coolant mass flux density, local steam content, system pressure, and fuel assembly heating length. These parameters are substituted into the target CHF relationship to calculate the predicted CHF value P for each fuel assembly within the hybrid reactor core. Combined with the actual heat flux density Q of each assembly, the deviation from the nucleus boiling ratio at each location is calculated using the formula DNBR = P / Q. The DNBR values ​​at all locations are iterated to find the global minimum DNBR value (i.e., the DNBR value at the minimum DNBR point), which is the target thermal safety margin of the hybrid reactor core. If the target thermal safety margin is ≥1.0 (meeting the safety redundancy requirements of the engineering design), the hybrid reactor core is deemed thermally safe; if it is less than 1.0, the relationship is refitted through data removal, regrouping, or adjustment of the relationship form.

[0042] The logic block diagram of this embodiment is as follows: Figure 2As shown, initial information on various fuel assemblies in the hybrid reactor core is first obtained, including the geometric characteristics of the mixing grid of the fuel assemblies (whether they contain mixing blades, thickness, height, and pressure drop coefficient) and the test results of critical heat flux density (CHF). Based on the initial information, it is determined whether the first preset condition of small difference between the geometric characteristics and CHF test results is met. If it is met, one CHF relation specific to each fuel assembly is selected as the target CHF relation. If not, the test data is grouped according to "grid type + heating method," and a general CHF relation applicable to the hybrid reactor core is fitted, which is then used as the target CHF relation. Based on multiple sets of thermal parameters, the predicted CHF values ​​P of the target CHF relation at the minimum DNBR point and minimum BO point under each operating condition are calculated. These are compared with the experimentally measured values ​​M to verify whether the double 95% criterion (second preset condition) of "95% probability M / P < CHF limit at a 95% confidence level" is met. If not, the relation is optimized by a preset strategy of eliminating data deviating from a normal distribution, regrouping, and adjusting the relation form, and the verification is repeated until it is met. The pressure drop coefficient penalty value of each fuel assembly is obtained, and it is combined with the verified target CHF relationship to carry out thermal safety margin analysis, and finally the target thermal safety margin of the hybrid reactor core is calculated.

[0043] This embodiment combines the geometric features of the mixing grid with CHF test data grouping to ensure that the target CHF relationship matches the actual heat transfer characteristics of the mixed core. At the same time, it introduces a pressure drop factor penalty to quantify the impact of pressure drop differences on CHF. It is applicable to mixed cores composed of fuel assemblies with basically the same critical heat flux density, as well as mixed cores composed of fuel assemblies with large differences in critical heat flux density. By introducing a pressure drop factor penalty or refitting the general relationship, the conservatism and accuracy of thermal safety margin analysis are improved.

[0044] In some embodiments, step S200 may include, but is not limited to, steps S210 to S230: Step S210: If multiple fuel assemblies in the hybrid reactor core meet the first preset condition, then select one critical heat flux density relationship from the critical heat flux density relationships corresponding to the multiple fuel assemblies as the target critical heat flux density relationship. Step S220: If multiple fuel assemblies in the hybrid reactor core do not meet the first preset condition, the test results corresponding to the multiple fuel assemblies are merged to obtain multiple sets of merged data, and the target critical heat flux density relationship is constructed based on the multiple sets of merged data.

[0045] The first preset condition includes: Each of the fuel assemblies includes a turbulence wing; Each of the aforementioned fuel assemblies does not include a turbulence fin; The difference in thickness between all the fuel assemblies is less than a first preset threshold; The difference in height between all the fuel assemblies is less than a second preset threshold; The difference in pressure drop coefficients among all the fuel assemblies is less than a third preset threshold; The difference between the test results for all the fuel assemblies is less than a fourth preset threshold.

[0046] In this embodiment, it is determined whether the various fuel assemblies in the hybrid reactor core meet a first preset condition. If the first preset condition is met, it indicates that the various fuel assemblies are highly similar in terms of mixing ability and critical heat flux density characteristics. In this case, one of the critical heat flux density relationships (e.g., CHF-A, CHF-B, CHF-C) corresponding to each of the various fuel assemblies can be directly selected as the target critical heat flux density relationship. If the first preset condition is not met, it indicates that there are significant differences between the various fuel assemblies, and their CHF performance is determined to be not directly equivalent. It is necessary to merge the critical heat flux density test results corresponding to the various fuel assemblies to obtain multiple sets of merged data, and reconstruct a general target critical heat flux density relationship (CHF) based on the multiple sets of merged data. 混合堆芯 ).

[0047] Specifically, the first precondition is the core criterion for determining whether multiple fuel assemblies can share the same CHF relationship, and the following conditions must be met simultaneously.

[0048] The mixing fin feature is consistent: each of the fuel assemblies includes a mixing fin; or, none of the fuel assemblies include a mixing fin. That is, the mixing grids of all assemblies must be completely identical in terms of the core feature of having or not having a mixing fin.

[0049] Geometric similarity: The maximum difference in thickness between all fuel assembly mixing grids is less than a first preset threshold (the first preset threshold can be set based on engineering experience, such as 0.5 mm, and the specific value needs to be determined through sensitivity analysis to ensure that the effect of thickness difference on CHF is negligible).

[0050] Installation position alignment: The maximum difference between the installation heights of all fuel assembly mixing grids in the reactor core is less than the second preset threshold (the second preset threshold can be set according to engineering experience, such as 50mm, to ensure that the height difference has a negligible impact on the axial flow and heat transfer of the coolant).

[0051] Similar hydraulic characteristics: The maximum difference between the pressure drop coefficients of all fuel assembly mixing grids is less than the third preset threshold (the third preset threshold can be set according to engineering experience, for example, 0.05, to ensure that the effect of resistance difference on coolant flow distribution is negligible).

[0052] Thermal performance convergence: The difference between the critical heat flux density test results for all fuel assemblies is less than a fourth preset threshold (the fourth preset threshold can be set based on engineering experience, for example, 5%, determined by calculating the mean deviation or standard deviation of test data for different assemblies to ensure the consistency of CHF characteristics). Critical heat flux density test results can include parameters such as coolant mass flux density, local coolant vapor content, system pressure, and fuel assembly heating length.

[0053] This embodiment simplifies the analysis by directly selecting existing CHF relationships and adding pressure drop penalties when the geometric characteristics are similar; when the geometric characteristics are significantly different, it reconstructs general CHF relationships by merging experimental data. This ensures a simplified process when the characteristics are similar and guarantees accuracy through data fusion when the characteristics are significantly different, thereby comprehensively improving the accuracy and reliability of the thermal safety margin analysis of hybrid reactor cores.

[0054] In some embodiments, step S220 may include, but is not limited to, steps S221 to S224: Step S221: Group the various fuel assemblies according to the grid geometry of the fuel assembly and the axial heating method of the fuel assembly to obtain multiple fuel assembly groups. The grid geometry includes typical grids or guide tube grids, and the axial heating method includes uniform heating or non-uniform heating. Step S222: For multiple fuel assembly groups, multiple sets of merged data are obtained based on the test results corresponding to the fuel assemblies in the same group; Step S223: Based on the variation pattern of the test results of multiple sets of merged data, determine the initial expression of the critical heat flux density relationship corresponding to the hybrid core. The variation pattern includes the trend of critical heat flux density with mass flux density, steam content, pressure or heating length. Step S224: Fit the coefficients in the initial expression to obtain the target critical heat flux density relationship corresponding to the hybrid core.

[0055] In this embodiment, the data is double-grouped based on the grid geometry of the fuel assembly and the axial heating method to ensure the homogeneity of the merged data.

[0056] In this embodiment, the fuel assembly cells can be categorized into typical cells and guide tube cells based on their structural differences. A typical cell refers to a cell containing only fuel rods (e.g., in a 5×5 rod bundle structure, all 25 rods are fuel rods, with no other functional components); a guide tube cell refers to a cell containing a guide tube (used to house control rods or measuring devices) (e.g., in a 5×5 rod bundle structure, the central rod is a guide tube, and the remaining 24 are fuel rods). The fuel assembly can also be categorized into uniform heating and non-uniform heating based on their axial power distribution. Uniform heating refers to the same heating power at all axial positions of the fuel assembly; non-uniform heating refers to the axial heating power of the fuel assembly distributed according to a specific pattern.

[0057] Assume the hybrid reactor core consists of fuel assemblies A, B, and C, and that they do not meet the first preset condition. Then, all critical heat flux density test data for A, B, and C need to be cross-grouped according to the two dimensions mentioned above. For example, the critical heat flux density test data points for fuel assemblies A, B, and C are shown in Table 1.

[0058] Table 1

[0059] The critical heat flux density test data points of fuel assembly A, fuel assembly B, and fuel assembly C are grouped together to develop the critical heat flux density relationship CHF. 混合堆芯 And obtain the critical heat flux density design limit CHF 限值 The critical heat flux density test data points after grouping are shown in Table 2.

[0060] Table 2

[0061] After obtaining the above multiple combined data, construct the target relation CHF. 混合堆芯 .

[0062] First, the variation trend of critical heat flux density (CHF) with parameters such as mass flux density, vapor content, and pressure in all merged data is analyzed. Common variation patterns across all grouped data (e.g., CHF is positively correlated with mass flux density and negatively correlated with vapor content) are summarized as the basis for determining the initial expression. Based on these variation patterns, an appropriate function form, such as a power function or polynomial, is selected.

[0063] The four sets of combined data (comprising hundreds of test points) obtained above were used as input for fitting. Mathematical optimization algorithms, such as the least squares method, were used to fit the coefficients in the initial expression, minimizing the overall error (e.g., sum of squared residuals) between the predicted and measured values ​​for all test points. After fitting, the target critical heat flux density relationship for the hybrid reactor core was obtained. Subsequently, the target critical heat flux density relationship was validated (e.g., using the double 95% criterion) to ensure its accuracy and conservatism.

[0064] This embodiment ensures that the merged data has the same physical background by using dual grouping based on grid geometry and axial heating method, avoiding deviations caused by data mixing of different structures or heating conditions. By analyzing the variation law of CHF with key parameters to determine the initial expression, it ensures that even in complex mixed cores with large differences in fuel assembly characteristics, a reliable and universal critical heat flux density relationship can be developed, laying a solid foundation for the accurate calculation of thermal safety margin.

[0065] In some embodiments, after step S220 and before step S300, the steps may include, but are not limited to, the following: Acquire multiple sets of thermal parameters of the hybrid reactor core at multiple acquisition times during operation; Based on each set of thermal parameters, the critical heat flux density corresponding to each set of thermal parameters is calculated using the target critical heat flux density relationship. The maximum value among the obtained critical heat flux densities is taken as the upper limit of the critical heat flux density, wherein the upper limit is used to limit the critical heat flux density calculated according to the target critical heat flux density formula.

[0066] In this embodiment, after constructing the target critical heat flux density relationship and before calculating the thermal safety margin, it is necessary to determine the upper limit value (CHF) of the critical heat flux density through sampling of operating parameters. 限值 This ensures the conservatism of the safety margin analysis.

[0067] Specifically, during reactor operation, multiple sets of thermal parameters are collected in real time through a sensor system. These thermal parameters need to cover key variables affecting CHF, including but not limited to: system pressure, coolant mass flow density, coolant inlet temperature / enthalpy, steam content, and fuel assembly heating length. The timing and operating conditions of the data collection need to cover the entire operating range of the reactor core, including steady-state conditions, transient conditions (such as power increases and coolant flow rate changes), and boundary conditions.

[0068] Statistical sampling methods (such as the Monte Carlo method) are used to sample all possible combinations of thermal parameters of the reactor. These parameter combinations reflect the state of the reactor at different times, such as normal operation and expected transient operation.

[0069] Each set of thermal parameters is used as input and substituted into the previously constructed target critical heat flux density (CHF) formula in batches to calculate the predicted CHF value for each specific operating condition. All calculated CHF predictions are iterated and compared to find the maximum value. This maximum value is defined as the upper limit of the design critical heat flux density for the hybrid reactor core. This ensures that even if the target CHF formula predicts an excessively high CHF value under extreme operating conditions, the actual CHF used will not exceed the maximum CHF level observed in actual operation of the hybrid reactor core, thus avoiding an overestimation of the safety margin.

[0070] This embodiment determines the upper limit of CHF using actual operating data, constrains the calculation results of the target CHF relationship, avoids overly high CHF predictions due to model bias or extreme parameters, and ensures that the thermal safety margin evaluation is conservative. Through global optimization, it actively identifies and limits the worst heat transfer capacity boundary, so that the final target thermal safety margin (such as minimum DNBR) can meet strict nuclear safety criteria (such as the 95 / 95 criterion), greatly enhancing the reliability of hybrid core safety analysis.

[0071] In some embodiments, step S300 may include, but is not limited to, steps S310 to S340: Step S310: For each fuel assembly, based on the target critical heat flux density relationship, calculate the predicted critical heat flux density of the fuel assembly under various test conditions. The predicted critical heat flux density includes the predicted critical heat flux density at the minimum deviation from the nucleus-boiling ratio point and the predicted critical heat flux density at the minimum burnout point; and... Step S320: For each test condition, perform the following processing: Based on the predicted critical heat flux density of the fuel assembly under various test conditions, the measured critical heat flux density at the minimum deviation from the nucleus-boiling ratio point and the measured critical heat flux density at the minimum burnout point in the test results corresponding to the fuel assembly, calculate the ratio of the predicted value to the measured value at the minimum deviation from the nucleus-boiling ratio point and the ratio of the predicted value to the measured value at the minimum burnout point under the test condition; Step S330: If each test condition meets the second preset condition, then the thermal safety margin of the reactor core is calculated according to the target critical heat flux density relationship. The second preset condition is that the ratio of the predicted value to the measured value of the minimum deviation from the nucleus boiling ratio point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level, and the ratio of the predicted value to the measured value of the minimum burn-out point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level. Step S340: If any test condition does not meet the second preset condition, the target critical heat flux density relationship is adjusted according to the preset strategy to obtain a new critical heat flux density relationship. The new critical heat flux density relationship is then used as the target critical heat flux density relationship. The process then jumps to the step of calculating the critical heat flux density corresponding to each set of thermal parameters based on the target critical heat flux density relationship.

[0072] The preset strategy includes at least one of the following: removing test results that deviate from the normal distribution from the test results of various fuel assemblies and changing the initial expression of the critical heat flux density relationship corresponding to the hybrid reactor core.

[0073] In this embodiment, the target critical heat flux density relationship CHF is determined. 混合堆芯 and the critical heat flux density design limit CHF 限值 Afterwards, statistical verification is required to ensure the conservatism and applicability of the relational formula, and finally the target thermal safety margin of the hybrid reactor core is calculated.

[0074] Specifically, for each fuel assembly (e.g., fuel assemblies A, B, and C) in the hybrid reactor core, the predicted critical heat flux (CHF) values ​​for that assembly under various test conditions are calculated based on the established target critical heat flux relationship. The predicted CHF values ​​include two types: the predicted CHF value at the minimum deviation from the nucleus-boiling ratio (DNBR) point, and the predicted CHF value at the minimum burnout (BO) point. The minimum DNBR point CHF prediction value refers to the predicted CHF value corresponding to the position with the minimum DNBR value in the entire rod bundle channel under that test condition. This point represents the most dangerous point based on traditional thermal design criteria. The minimum BO point CHF prediction value refers to the predicted CHF value corresponding to the position where the earliest burnout was actually observed in the experiment under that test condition. This point is based on real physical phenomena and is generally more conservative.

[0075] For each test condition, based on the predicted critical heat flux density of the fuel assembly and its corresponding measured value in the original critical heat flux density test, two key ratios are calculated: the ratio of the measured value (M) to the predicted value (P) of the minimum DNBR point under this test condition (M / P); and the ratio of the measured value (M) to the predicted value (P) of the minimum BO point under this test condition (M / P). If M / P > 1, it indicates that the formula prediction is conservative (the actual measured safety boundary is higher than the predicted value); if M / P = 1, it indicates that the formula prediction is perfect; if M / P < 1, it indicates that the formula prediction is not conservative (the actual measured safety boundary is lower than the predicted value, which is dangerous).

[0076] If the calculation results for all test conditions meet the second preset condition, the critical heat flux density formula is deemed to have passed verification. For all test conditions, the following must be simultaneously satisfied: the ratio of the minimum DNBR point has a preset probability (e.g., 95%) that it is less than the critical heat flux density design limit at a preset confidence level (e.g., 95%); the ratio of the minimum BO point has a preset probability (e.g., 95%) that it is less than the critical heat flux density design limit at a preset confidence level (e.g., 95%). Subsequently, a formal thermal safety margin calculation can be performed on the hybrid reactor core based on this formula. The second preset condition refers to the statistical evaluation of the ratio data calculated for all test conditions, confirming that both the minimum DNBR point (M / P) value and the minimum BO point (M / P) value have a 95% probability (at a 95% confidence level) that they are less than the critical heat flux density design limit (CHF). 限值 This principle is often referred to as the "double 95% criterion" in the field of nuclear safety. The 95% confidence level means that the statistical results are reliable, with 95% certainty that the conclusion "M / P ≤ CHF limit" is correct (excluding misjudgments caused by sampling errors); the 95% probability means that the coverage of the results means that more than 95% of the test conditions verified meet the "M / P ≤ CHF limit" (ensuring that the relationship is conservative in the vast majority of test conditions).

[0077] If the second preset condition is met, it proves that the target critical heat flux density relationship is applicable and conservative for the hybrid reactor core. Subsequently, this relationship can be embedded into the sub-channel analysis program or system program to simulate the thermal-hydraulic state of the hybrid reactor core under all operating conditions, ultimately obtaining the minimum DNBR value for the entire reactor core, which is the target thermal safety margin.

[0078] If, during the statistical verification process, any test condition is found to fail to meet the second preset condition (i.e., the "double 95% criterion"), the target critical heat flux density formula will not be used immediately. Instead, it will be adjusted according to a preset strategy to obtain a new critical heat flux density formula. Subsequently, the new critical heat flux density formula will be used as the target critical heat flux density formula again, and the process will jump to the step of "calculating the critical heat flux density corresponding to each set of thermal parameters using the target critical heat flux density formula." This means that the adjusted formula needs to go through the entire process of determining the design upper limit, calculating the predicted value, and conducting statistical verification again, thus forming an iterative optimization closed loop until a valid and qualified target formula is generated.

[0079] Pre-defined strategies may include at least one of the following: removing experimental results that deviate from a normal distribution, adjusting the grouping method and regrouping, and changing the initial expression. Removing experimental results that deviate from a normal distribution involves data cleaning by eliminating outliers in the experimental data. These outliers may originate from random errors during the experiment, transient failures of measuring equipment, or specific non-representative operating conditions. Their presence can distort the fit of the relational expression, leading to systematic biases in predictions. Regrouping the experimental data to optimize the data organization structure can be attempted by refining the grouping dimensions or regrouping based on the range of certain key parameters (such as mass flow rate or vapor content). Adjusting the form of the relational expression and changing the function structure can improve the model's prediction accuracy and conservatism.

[0080] This embodiment ensures the predictive reliability of the target CHF relationship across the entire operating range by covering multiple test conditions and comparing the two key hazardous locations: the minimum DNBR point and the minimum BO point. The second preset condition, through the dual constraints of "preset probability + design limit", ensures that the target CHF relationship is conservative in most operating conditions, avoiding overestimation of the safety margin. By introducing an iterative adjustment mechanism based on data cleaning and model optimization, it ensures that a critical heat flux density relationship that meets both high accuracy requirements and nuclear safety conservatism requirements can be derived.

[0081] In some embodiments, step S300 may also include, but is not limited to, steps S350 to S360: Step S350: Obtain the penalty value of the pressure drop coefficient of various fuel components; Step S360: Based on the target critical heat flux density relationship and the pressure drop coefficient penalty value, perform thermal safety margin analysis to obtain the target thermal safety margin.

[0082] Specifically, step S350 may include, but is not limited to, the following steps: Obtain the pressure drop coefficient uncertainty value for each of the various fuel assemblies, and the pressure drop coefficient uncertainty value is used to quantify the prediction error range of the pressure drop coefficient for each of the fuel assemblies; The first value is obtained by calculating the sum of squares of the uncertainty values ​​of the pressure drop coefficient for each fuel assembly; Calculate the square root of the first value to obtain the penalty value of the pressure drop coefficient.

[0083] In a hybrid reactor core composed of multiple fuel assemblies, different types of fuel assemblies have different pressure drop coefficients due to differences in their internal structures, such as mixing grids. This results in varying resistance as coolant flows through different assemblies, thus affecting the coolant flow distribution throughout the core. Because the pressure drop coefficients of different fuel assemblies are different, a pressure drop coefficient penalty is introduced in the calculation to address this issue.

[0084] In this embodiment, the pressure drop coefficient penalty value is a comprehensive quantification of the uncertainty of the pressure drop coefficient of all fuel assemblies, used to compensate for the risk of coolant flow distribution caused by pressure drop differences. Sources of pressure drop coefficient uncertainty include: based on fuel assembly design data, test data, and manufacturing tolerances, obtained through statistical analysis, including but not limited to: deviations caused by pressure sensor accuracy and flow measurement errors when measuring the assembly pressure drop coefficient in a high-temperature, high-pressure loop; pressure drop coefficient fluctuations caused by mixing grid processing accuracy and assembly errors; and the impact of coolant corrosion and scaling on the assembly flow channels during core operation, leading to estimated pressure drop coefficient deviations.

[0085] For example, if the uncertainty of the pressure drop coefficient of fuel assembly A is A, the uncertainty of the pressure drop coefficient of fuel assembly B is B, and the uncertainty of the pressure drop coefficient of fuel assembly C is C, then the penalty for the pressure drop coefficient of the fuel assemblies is... .

[0086] In the final thermal safety margin analysis, the validated target critical heat flux density (CHF) relationship is used as the model for calculating local CHF values ​​in the program. In the program settings, the calculated pressure drop factor penalty value is used as an uncertainty input or bias factor for the overall core channel pressure drop calculation. The program considers the additional disturbance introduced by this penalty value when calculating flow distribution. The program is run to calculate a series of operating conditions covering normal operation and expected transients. The program comprehensively considers the CHF capability predicted by the target relationship and the penalized, more severe flow distribution, ultimately searching for and outputting the minimum DNBR value for the entire hybrid core under all operating conditions. This value is the final determined target thermal safety margin.

[0087] This embodiment quantifies the superimposed risk of pressure drop uncertainty of each fuel assembly by using the pressure drop coefficient penalty value, and makes a conservative correction to the CHF prediction value, so that the final target thermal safety margin has higher reliability, ensuring that the conclusion of the safety analysis fully covers the potential risks and meets the strict conservative requirements of nuclear safety regulations.

[0088] This application's embodiments systematically analyze the mixing grid geometry characteristics and critical heat flux density test data of different types of fuel assemblies in a hybrid reactor core. For cases with similar and significantly different assembly characteristics, differentiated strategies are employed: using existing relationships combined with pressure drop coefficient penalties, or reconstructing general relationships by merging test data. A rigorous statistical verification and iterative optimization mechanism based on the "double 95% criterion" is introduced. Finally, thermal safety margin calculations are completed under the condition of considering pressure drop uncertainty penalties. This effectively solves the problem that traditional methods cannot accurately handle hybrid reactor cores with large differences in critical heat flux densities, significantly improving the conservatism, accuracy, and engineering applicability of thermal safety assessments. This application's embodiments are applicable to the thermal safety analysis of hybrid reactor cores composed of fuel assemblies with essentially the same critical heat flux density, as well as those composed of fuel assemblies with significant differences in critical heat flux density; improving thermal safety margins and ensuring the conservatism of thermal safety assessments for hybrid reactor core fuel assemblies.

[0089] Please see Figure 3 This application also provides a device 400 for determining the thermal safety margin of a reactor core, which can implement the above-mentioned method for determining the thermal safety margin of a reactor core. The device includes: The acquisition module 10 is used to acquire initial information of multiple fuel assemblies in the hybrid reactor core. The initial information includes the mixing grid geometry of the fuel assembly and the test results obtained by conducting a critical heat flux density test on the fuel assembly. The mixing grid geometry of the fuel assembly includes at least one of the following: whether the fuel assembly includes mixing blades, the thickness of the fuel assembly, the height of the fuel assembly, and the pressure drop coefficient of the fuel assembly. The determination module 20 is used to determine the target critical heat flux density relationship corresponding to the hybrid reactor core based on the initial information of various fuel assemblies. The calculation module 30 is used to calculate the thermal safety margin of the hybrid core according to the target critical heat flux density relationship, and obtain the target thermal safety margin.

[0090] In some implementations, the determining module 20 may include: The selection submodule is used to select one critical heat flux density relationship from the critical heat flux density relationships corresponding to the various fuel assemblies in the hybrid reactor core as the target critical heat flux density relationship if the various fuel assemblies in the hybrid reactor core meet the first preset conditions. The first preset conditions include: each fuel assembly includes a swirl bar; each fuel assembly does not include a swirl bar; the difference in thickness between all fuel assemblies is less than a first preset threshold; the difference in height between all fuel assemblies is less than a second preset threshold; the difference in pressure drop coefficient between all fuel assemblies is less than a third preset threshold; and the difference in test results corresponding to all fuel assemblies is less than a fourth preset threshold. A submodule is constructed to merge the test results corresponding to the various fuel assemblies in the hybrid reactor core if the various fuel assemblies do not meet the first preset condition, to obtain multiple sets of merged data, and to construct a target critical heat flux density relationship based on the multiple sets of merged data.

[0091] In some implementations, building a submodule may include: A grouping unit is used to group multiple fuel assemblies according to the grid geometry of the fuel assembly and the axial heating method of the fuel assembly, to obtain multiple fuel assembly groups. The grid geometry includes typical grids or guide tube grids, and the axial heating method includes uniform heating or non-uniform heating. The merging unit is used to group multiple fuel assemblies and obtain multiple sets of merged data based on the test results corresponding to the fuel assemblies in the same group.

[0092] In some implementations, the construction submodule may further include: The determining unit is used to determine the initial expression of the critical heat flux density relationship corresponding to the hybrid core based on the variation law of the test results of multiple sets of the combined data. The variation law includes the variation trend of critical heat flux density with mass flux density, steam content, pressure or heating length. A fitting unit is used to fit the coefficients in the initial expression to obtain the target critical heat flux density relationship corresponding to the hybrid core.

[0093] In some implementations, the construction submodule may further include: The first acquisition unit is used to acquire multiple sets of thermal parameters collected at multiple acquisition times during the operation of the hybrid reactor core. The first calculation unit is used to calculate the critical heat flux density corresponding to each set of thermal parameters based on the target critical heat flux density relationship. The second calculation unit is used to take the maximum value among the obtained critical heat flux densities as the upper limit of the critical heat flux density, wherein the upper limit is used to limit the critical heat flux density calculated according to the target critical heat flux density formula.

[0094] In some implementations, the computing module 30 may include: The first calculation submodule is used to calculate, for each of the fuel assemblies, the predicted critical heat flux density of the fuel assembly under various test conditions according to the target critical heat flux density relationship. The predicted critical heat flux density includes the predicted critical heat flux density at the minimum deviation from the nucleus boiling ratio point and the predicted critical heat flux density at the minimum burnout point. The second calculation submodule is used to perform the following processing for each test condition: based on the predicted critical heat flux density of the fuel assembly under various test conditions, the measured critical heat flux density at the minimum deviation from the nucleus-boiling ratio point and the measured critical heat flux density at the minimum burnout point in the test results corresponding to the fuel assembly, the ratio of the predicted value to the measured value at the minimum deviation from the nucleus-boiling ratio point and the ratio of the predicted value to the measured value at the minimum burnout point under the test condition are calculated. The third calculation submodule is used to calculate the thermal safety margin of the reactor core according to the target critical heat flux density relationship if each test condition meets the second preset condition. The second preset condition is that the ratio of the predicted value to the measured value of the minimum deviation from the nucleus boiling ratio point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level, and the ratio of the predicted value to the measured value of the minimum burn-out point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level.

[0095] In some implementations, the computing module 30 may further include: The adjustment submodule is used to adjust the target critical heat flux density relationship according to a preset strategy if any test condition does not meet the second preset condition, to obtain a new critical heat flux density relationship, and to use the new critical heat flux density relationship as the target critical heat flux density relationship. The module then jumps to the step of calculating the critical heat flux density corresponding to each set of thermal parameters based on the target critical heat flux density relationship. The preset strategy includes at least one of the following: removing test results that deviate from a normal distribution from the test results of various fuel assemblies and changing the initial expression of the critical heat flux density relationship corresponding to the hybrid reactor core.

[0096] In some implementations, the computing module 30 may further include: The acquisition submodule is used to acquire the penalty value of the pressure drop coefficient of various fuel components; The analysis submodule is used to perform thermal safety margin analysis based on the target critical heat flux density relationship and the pressure drop coefficient penalty value, so as to obtain the target thermal safety margin.

[0097] In some implementations, the acquisition submodule may include: The second acquisition unit is used to acquire the pressure drop coefficient uncertainty value corresponding to each of the various fuel assemblies, and the pressure drop coefficient uncertainty value is used to quantify the prediction error range of the pressure drop coefficient of each of the fuel assemblies; The third calculation unit is used to calculate the sum of squares of the uncertainty values ​​of the pressure drop coefficient for each fuel assembly to obtain the first value; The fourth calculation unit is used to calculate the square root of the first value to obtain the penalty value of the pressure drop coefficient.

[0098] The specific implementation of the device for determining the thermal safety margin of the reactor core is basically the same as the specific implementation of the method for determining the thermal safety margin of the reactor core described above, and will not be repeated here.

[0099] The method and apparatus for determining the thermal safety margin of a reactor core provided in this application acquire initial information of various fuel assemblies in a hybrid reactor core. This initial information includes the geometric characteristics of the mixing grid of the fuel assemblies (such as whether they include mixing vanes, thickness, height, and pressure drop coefficient, at least one of these), and test results obtained from critical heat flux density tests on the fuel assemblies. Based on the initial information of the various fuel assemblies, a target critical heat flux density relationship is determined for the hybrid reactor core. The thermal safety margin of the hybrid reactor core is calculated based on the target critical heat flux density relationship to obtain the target thermal safety margin. This application significantly improves the accuracy and applicability of the thermal safety margin calculation for hybrid reactor cores, ensures the conservatism of the thermal safety evaluation of hybrid reactor core fuel assemblies, and can be effectively applied to engineering practice of thermal safety margin analysis of hybrid reactor core fuel assemblies, providing reliable technical support for the safe operation of hybrid nuclear reactor cores.

[0100] The embodiments described in this application are for the purpose of more clearly illustrating the technical solutions of the embodiments of this application, and do not constitute a limitation on the technical solutions provided by the embodiments of this application. As those skilled in the art will know, with the evolution of technology and the emergence of new application scenarios, the technical solutions provided by the embodiments of this application are also applicable to similar technical problems.

[0101] Those skilled in the art will understand that the technical solutions shown in the figures do not constitute a limitation on the embodiments of this application, and may include more or fewer steps than shown, or combine certain steps, or different steps.

[0102] The device embodiments described above are merely illustrative. The units described as separate components may or may not be physically separate; that is, they may be located in one place or distributed across multiple network units. Some or all of the modules can be selected to achieve the purpose of this embodiment according to actual needs.

[0103] Those skilled in the art will understand that all or some of the steps in the methods disclosed above, as well as the functional modules / units in the systems and devices, can be implemented as software, firmware, hardware, or suitable combinations thereof.

[0104] The terms “first,” “second,” “third,” “fourth,” etc. (if present) in the specification and accompanying drawings of this application are used to distinguish similar objects and are not necessarily used to describe a specific order or sequence. It should be understood that such data can be interchanged where appropriate so that the embodiments of this application described herein can be implemented in orders other than those illustrated or described herein. Furthermore, the terms “comprising” and “having,” and any variations thereof, are intended to cover non-exclusive inclusion; for example, a process, method, system, product, or apparatus that comprises a series of steps or units is not necessarily limited to those steps or units explicitly listed, but may include other steps or units not explicitly listed or inherent to such processes, methods, products, or apparatus.

[0105] It should be understood that in this application, "at least one (item)" means one or more, and "more than" means two or more. "And / or" is used to describe the relationship between related objects, indicating that three relationships can exist. For example, "A and / or B" can represent three cases: only A exists, only B exists, and both A and B exist simultaneously, where A and B can be singular or plural. The character " / " generally indicates that the preceding and following related objects are in an "or" relationship. "At least one (item) of the following" or similar expressions refer to any combination of these items, including any combination of single or plural items. For example, at least one (item) of a, b, or c can represent: a, b, c, "a and b", "a and c", "b and c", or "a and b and c", where a, b, and c can be single or multiple.

[0106] In the several embodiments provided in this application, it should be understood that the disclosed apparatus and methods can be implemented in other ways. For example, the apparatus embodiments described above are merely illustrative; for instance, the division of the units described above is only a logical functional division, and in actual implementation, there may be other division methods. For example, multiple units or components may be combined or integrated into another system, or some features may be ignored or not executed. Furthermore, the coupling or direct coupling or communication connection shown or discussed may be through some interfaces; the indirect coupling or communication connection between apparatuses or units may be electrical, mechanical, or other forms.

[0107] The units described above as separate components may or may not be physically separate. The components shown as units may or may not be physical units; that is, they may be located in one place or distributed across multiple network units. Some or all of the units can be selected to achieve the purpose of this embodiment according to actual needs.

[0108] Furthermore, the functional units in the various embodiments of this application can be integrated into one processing unit, or each unit can exist physically separately, or two or more units can be integrated into one unit. The integrated unit can be implemented in hardware or as a software functional unit.

[0109] If the integrated unit is implemented as a software functional unit and sold or used as an independent product, it can be stored in a computer-readable storage medium. Based on this understanding, the technical solution of this application, in essence, or the part that contributes to the prior art, or all or part of the technical solution, can be embodied in the form of a software product. This computer software product is stored in a storage medium and includes multiple instructions to cause a computer device (which may be a personal computer, server, or network device, etc.) to execute all or part of the steps of the methods of the various embodiments of this application. The aforementioned storage medium includes various media capable of storing programs, such as USB flash drives, portable hard drives, read-only memory (ROM), random access memory (RAM), magnetic disks, or optical disks.

[0110] The preferred embodiments of the present application have been described above with reference to the accompanying drawings, but this does not limit the scope of the claims of the present application. Any modifications, equivalent substitutions, and improvements made by those skilled in the art without departing from the scope and substance of the embodiments of the present application shall be within the scope of the claims of the present application.

Claims

1. A method of determining a thermal safety margin of a reactor core, characterized by, The method includes: The initial information of various fuel assemblies in the reactor core is obtained. The initial information includes the mixing grid geometry of the fuel assembly and the test results obtained by conducting a critical heat flux density test on the fuel assembly. The mixing grid geometry of the fuel assembly includes at least one of the following: whether the fuel assembly includes a mixing blade, the thickness of the fuel assembly, the height of the fuel assembly, and the pressure drop coefficient of the fuel assembly. Based on the initial information of various fuel assemblies, the target critical heat flux density relationship corresponding to the reactor core is determined; The thermal safety margin of the reactor core is calculated based on the target critical heat flux density relationship to obtain the target thermal safety margin. The step of determining the target critical heat flux density relationship corresponding to the reactor core based on initial information of various fuel assemblies includes: If multiple fuel assemblies in the reactor core meet the first preset condition, then one critical heat flux density relationship is selected from the critical heat flux density relationships corresponding to the multiple fuel assemblies as the target critical heat flux density relationship. If multiple fuel assemblies in the reactor core do not meet the first preset condition, the test results corresponding to the multiple fuel assemblies are merged to obtain multiple sets of merged data, and the target critical heat flux density relationship is constructed based on the multiple sets of merged data. The first preset condition includes: Each of the fuel assemblies includes a turbulence wing; Each of the aforementioned fuel assemblies does not include a turbulence wing; The difference in thickness between all the fuel assemblies is less than a first preset threshold; The difference in height between all the fuel assemblies is less than a second preset threshold; The difference in pressure drop coefficients among all the fuel assemblies is less than a third preset threshold; The difference between the test results corresponding to all the fuel components is less than a fourth preset threshold; The construction of the target critical heat flux density relationship based on multiple sets of merged data includes: Based on the variation patterns of the test results of multiple sets of combined data, an initial expression for the critical heat flux density relationship corresponding to the reactor core is determined. The variation patterns include the trend of critical heat flux density with mass flux density, steam content, pressure, or heating length. By fitting the coefficients in the initial expression, the target critical heat flux density relationship corresponding to the reactor core is obtained; The step of calculating the thermal safety margin of the reactor core based on the target critical heat flux density relationship to obtain the target thermal safety margin includes: For each of the fuel assemblies, based on the target critical heat flux density relationship, predicted critical heat flux density values ​​for the fuel assembly are calculated under various test conditions. These predicted critical heat flux density values ​​include predicted values ​​for the critical heat flux density at the minimum deviation from the nucleus-boiling ratio point and predicted values ​​for the critical heat flux density at the minimum burnout point; and... For each test condition, the following processing is performed: Based on the predicted critical heat flux density of the fuel assembly under various test conditions, the measured critical heat flux density at the minimum deviation from the nucleus-boiling ratio point and the measured critical heat flux density at the minimum burnout point in the test results corresponding to the fuel assembly, the ratio of the predicted value to the measured value at the minimum deviation from the nucleus-boiling ratio point and the ratio of the predicted value to the measured value at the minimum burnout point are obtained under the test condition; If each test condition meets the second preset condition, then the thermal safety margin of the reactor core is calculated according to the target critical heat flux density relationship. The second preset condition is that the ratio of the predicted value to the measured value of the minimum deviation from the nucleus boiling ratio point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level, and the ratio of the predicted value to the measured value of the minimum burn-out point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level.

2. The method according to claim 1, characterized in that, The test results for various fuel assemblies are merged to obtain multiple sets of merged data, including: Based on the grid geometry of the fuel assembly and the axial heating method of the fuel assembly, multiple fuel assembly groups are obtained. The grid geometry includes typical grids or guide tube grids, and the axial heating method includes uniform heating or non-uniform heating. For multiple fuel component groups, multiple sets of merged data are obtained based on the test results corresponding to the fuel components in the same group.

3. The method according to claim 1, characterized in that, After constructing the target critical heat flux density relationship based on multiple sets of merged data, and before calculating the thermal safety margin of the reactor core based on the target critical heat flux density relationship to obtain the target thermal safety margin, the method further includes: Acquire multiple sets of thermal parameters of the reactor core at multiple acquisition times during operation; Based on each set of thermal parameters, the critical heat flux density corresponding to each set of thermal parameters is calculated using the target critical heat flux density relationship. The maximum value among the obtained critical heat flux densities is taken as the upper limit of the critical heat flux density, wherein the upper limit is used to limit the critical heat flux density calculated according to the target critical heat flux density formula.

4. The method according to claim 3, characterized in that, The method further includes: If any test condition does not meet the second preset condition, the target critical heat flux density relationship is adjusted according to the preset strategy to obtain a new critical heat flux density relationship. The new critical heat flux density relationship is then used as the target critical heat flux density relationship. The process then jumps to the step of calculating the critical heat flux density corresponding to each set of thermal parameters based on the target critical heat flux density relationship. The preset strategy includes at least one of the following: removing test results that deviate from the normal distribution from the test results of various fuel assemblies and changing the initial expression of the critical heat flux density relationship corresponding to the reactor core.

5. The method according to claim 1, characterized in that, The step of calculating the thermal safety margin of the reactor core based on the target critical heat flux density relationship to obtain the target thermal safety margin includes: Obtain the penalty value for the pressure drop coefficient of various fuel assemblies; Based on the target critical heat flux density relationship and the pressure drop coefficient penalty value, a thermal safety margin analysis is performed to obtain the target thermal safety margin.

6. The method according to claim 5, characterized in that, The penalty value for obtaining the pressure drop coefficient of various fuel assemblies includes: Obtain the pressure drop coefficient uncertainty value for each of the various fuel assemblies, and the pressure drop coefficient uncertainty value is used to quantify the prediction error range of the pressure drop coefficient for each of the fuel assemblies; The first value is obtained by calculating the sum of squares of the uncertainty values ​​of the pressure drop coefficient for each fuel assembly; Calculate the square root of the first value to obtain the penalty value of the pressure drop coefficient.

7. A device for determining the thermal safety margin of a reactor core, characterized in that, The device includes: The acquisition module is used to acquire initial information of various fuel assemblies in the reactor core. The initial information includes the mixing grid geometry of the fuel assembly and the test results obtained by conducting a critical heat flux density test on the fuel assembly. The mixing grid geometry of the fuel assembly includes at least one of the following: whether the fuel assembly includes mixing blades, the thickness of the fuel assembly, the height of the fuel assembly, and the pressure drop coefficient of the fuel assembly. The determination module is used to determine the target critical heat flux density relationship corresponding to the reactor core based on the initial information of various fuel assemblies; if various fuel assemblies in the reactor core meet a first preset condition, then one critical heat flux density relationship is selected from the critical heat flux density relationships corresponding to the various fuel assemblies as the target critical heat flux density relationship; if various fuel assemblies in the reactor core do not meet the first preset condition, then the test results corresponding to the various fuel assemblies are merged to obtain multiple sets of merged data, and the target critical heat flux density relationship is constructed based on the multiple sets of merged data; wherein, the first preset condition includes: each fuel assembly includes a turbulence fin; each fuel assembly None of them include stirring wings; the difference in thickness between all the fuel assemblies is less than a first preset threshold; the difference in height between all the fuel assemblies is less than a second preset threshold; the difference in pressure drop coefficient between all the fuel assemblies is less than a third preset threshold; the difference in test results corresponding to all the fuel assemblies is less than a fourth preset threshold; based on the variation law of the test results of multiple sets of merged data, an initial expression for the critical heat flux density relationship corresponding to the reactor core is determined, the variation law including the variation trend of critical heat flux density with mass flux density, steam content, pressure or heating length; the coefficients in the initial expression are fitted to obtain the target critical heat flux density relationship corresponding to the reactor core; The calculation module is used to calculate the thermal safety margin of the reactor core according to the target critical heat flux density formula, and obtain the target thermal safety margin; for each fuel assembly, according to the target critical heat flux density formula, the predicted critical heat flux density of the fuel assembly under various test conditions is calculated, the predicted critical heat flux density includes the predicted value of the critical heat flux density at the minimum deviation from the nucleus-boiling ratio point and the predicted value of the critical heat flux density at the minimum burnout point; for each test condition, the following processing is performed: based on the predicted critical heat flux density of the fuel assembly under various test conditions, the measured value of the critical heat flux density at the minimum deviation from the nucleus-boiling ratio point in the test results corresponding to the fuel assembly, and the predicted value of the critical heat flux density at the minimum deviation from the nucleus-boiling ratio point, the predicted value of the critical heat flux density at the minimum burnout point, the predicted value of the critical heat flux density at the minimum deviation from the nucleus-boiling ratio point, and the predicted value of the critical heat flux density at the minimum burnout point, ... burnout point, the predicted value of the critical heat flux density at the minimum burnout point, the predicted value of the critical heat flux density at the minimum burnout point, the predicted value of the critical heat flux density at the minimum burnout point, the predicted value of the critical heat flux density at the minimum burnout point, the predicted value of the critical heat flux density at the minimum burnout point, the predicted value of the critical heat flux density at the minimum burnout point, the predicted value of the critical heat flux density at the minimum burnout point, the predicted value of the critical heat The measured value of the critical heat flux density at the small burn-out point is calculated to obtain the ratio of the predicted value to the measured value of the minimum deviation from the nucleus boiling ratio point under the test conditions, and the ratio of the predicted value to the measured value of the minimum burn-out point. If each test condition meets the second preset condition, the thermal safety margin of the reactor core is calculated according to the target critical heat flux density relationship. The second preset condition is that the ratio of the predicted value to the measured value of the minimum deviation from the nucleus boiling ratio point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level, and the ratio of the predicted value to the measured value of the minimum burn-out point has a preset probability of being less than the upper limit of the critical heat flux density at a preset confidence level.

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