Burnup calculation data acquisition method, burnup calculation method, device and system

Abstract

The invention discloses a burnup calculation data acquisition method, a burnup calculation method, a burnup calculation device and a burnup calculation system. The burnup calculation data acquisition method comprises the following steps: acquiring burnup point information of a spent fuel assembly in a burnup life period; and according to the burn-up point information, establishing a microscopic reaction rate integral average model of the target nuclide in the spent fuel assembly. According to the burnup point information and the microscopic reaction rate integral average model, burnup calculation data of the target nuclide are obtained through calculation, and the burnup calculation data comprise the average neutron flux density value and the average reaction cross section value of each burnup step. According to the method for acquiring the burnup calculation data, more accurate section data and flux density data can be obtained, so that the accuracy of burnup calculation is effectively improved.

Description

Technical Field

[0001] The present invention specifically relates to a method for acquiring fuel consumption calculation data, a fuel consumption calculation method and device, and a system. Background Art

[0002] In the radiation protection design process of spent fuel reprocessing plants, the source term calculation of spent fuel assemblies is particularly important. It is the basis of the source terms of each process in the spent fuel reprocessing plant and the basis of radiation protection design. Burnup calculation is an important type of source term calculation. Through burnup calculation, the nuclide content in the spent fuel can be obtained, which helps to evaluate the decay process of radioactive nuclides.

[0003] The ORIGEN series of programs are widely used internationally as ignition consumption and radionuclide decay programs, which can be used for the evaluation and calculation of spent fuel and high-level radioactive waste. The ORIGEN series of programs contains multiple different ignition consumption programs, which have many common features, but also have great differences. They are the same in the solution method of the burnup equation, but the databases used (mainly including cross-section data and flux density data) are very different. At present, the burnup calculation usually directly uses the database supporting the ORIGEN series of programs. Due to the differences in supporting databases, and the problem that the database selected in the calculation does not completely match the problem being calculated, the results of different ignition consumption calculation programs when calculating the same burnup problem are quite different.

[0004] How to obtain more accurate cross-sectional data and flux density data to improve the accuracy of fuel consumption calculation has become a technical problem that needs to be solved urgently. Summary of the invention

[0005] The technical problem to be solved by the present invention is to provide a method for obtaining fuel consumption calculation data, a fuel consumption calculation method and device and a system in view of the above-mentioned deficiencies in the prior art. The method for obtaining fuel consumption calculation data can obtain more accurate cross-sectional data and flux density data, thereby effectively improving the accuracy of fuel consumption calculation.

[0006] According to an embodiment of the first aspect of the present invention, there is provided a method for obtaining fuel consumption calculation data, comprising the following steps:

[0007] S11. Obtaining the burnup point information of the spent fuel assembly within its burnup life;

[0008] S12. Establishing a microscopic reaction rate integral average model of target nuclides in the spent fuel assembly according to the burnup point information;

[0009] S13. Calculate the burnup calculation data of the target nuclide according to the burnup point information and the microscopic reaction rate integral average model, wherein the burnup calculation data includes an average neutron flux density value and an average reaction cross section value.

[0010] Preferably, the burnup point information includes the number of burnup points, the number of burnup points is multiple, and the burnup point information also includes the burnup depth corresponding to each burnup point, the neutron flux density value corresponding to each burnup point, and the reaction cross section value corresponding to each burnup point.

[0011] Preferably, the step S11, obtaining the burnup point information of the spent fuel assembly within the burnup life, includes:

[0012] Obtaining the burnup life of the spent fuel assembly and a burnup depth parameter corresponding to the burnup life;

[0013] According to the burnup life and the corresponding burnup depth parameter within the burnup life, the burnup points are divided to obtain a burnup point division result, wherein the burnup point division result includes: the number of burnup points and the burnup depth corresponding to each burnup point;

[0014] Establishing a neutronics solution model for the spent fuel assembly;

[0015] According to the division result of the burnup points and the neutronics solution model, the neutron flux density value corresponding to each burnup point and the reaction cross section value corresponding to each burnup point are calculated.

[0016] Preferably, the neutronics solution model includes: a geometric model of the spent fuel assembly, material properties and boundary conditions and neutron source information.

[0017] Preferably, the step S12, establishing a microscopic reaction rate integral average model of target nuclides in the spent fuel assembly according to the burnup point information, comprises:

[0018] According to the burnup point information, a burnup integral point is selected, and a burnup step is formed between two adjacent burnup integral points;

[0019] Establish the microscopic reaction rate integral average model of the target nuclide for each burnup step;

[0020] The microscopic reaction rate integral average model of any burnup step is:

[0021]

[0022]

[0023]

[0024] In the formula, is the average neutron flux density of the corresponding burnup step between two adjacent burnup integration points, φ i is the neutron flux density at each burnup point in the burnup step, is the average reaction rate of the burnup step, R i is the microscopic reaction rate of each burnup point in the burnup step, is the average reaction cross section of the burnup step, BU j is the burnup depth of each burnup point in the burnup step, startBU is the lower bound of the integration interval, stopBU is the upper bound of the integration interval, and num_integral is the number of integration points.

[0025] Preferably, the step S13, according to the burnup point information and the microscopic reaction rate integral average model, calculates the burnup calculation data of the target nuclide, the burnup calculation data including the average neutron flux density value and the average reaction cross section value, including:

[0026] According to the burnup point information, integral point information corresponding to each burnup step is obtained, wherein the integral point information includes: neutron flux density of each burnup point in the burnup step, microscopic reaction rate of each burnup point in the burnup step, and burnup depth of each burnup point in the burnup step;

[0027] The integral point information corresponding to each burnup step is respectively substituted into the microscopic reaction rate integral average model to calculate the average neutron flux density value and the average reaction cross section value of the target nuclide in each burnup step.

[0028] According to an embodiment of the second aspect of the present invention, there is provided a method for calculating fuel consumption, comprising the following steps:

[0029] S21. Establish the burnup equation of the target nuclide;

[0030] S22, obtaining the initial nuclide content of the target nuclide;

[0031] S23, according to the above-mentioned method for obtaining burnup calculation data, obtaining the average neutron flux density and average reaction cross section of each burnup step;

[0032] S24. According to the initial nuclide content, the average neutron flux density value of each burnup step and the average reaction cross section value, the burnup equation is iteratively calculated to obtain the nuclide content of the target nuclide at the end of the burnup life.

[0033] Preferably, the burnup equation of the target nuclide is:

[0034]

[0035] Where: N i (t) is the content of target nuclide i at time t; σ a,i is the microscopic neutron absorption cross section of the i-th nuclide group (unit: barn); φ is the single group neutron flux density (unit: cm -2 s-1 );σ k,i is the single-group microscopic neutron cross section for the kth nuclide to produce the ith nuclide (unit: barn); b j,i is the decay branch ratio of the jth nuclide to produce the ith nuclide; is the decay constant of the ith nuclide.

[0036] Preferably, the step S24, iteratively calculating the burnup equation according to the initial nuclide content, the average neutron flux density of each burnup step, and the average reaction cross section to obtain the nuclide content of the target nuclide at the end of the burnup life, comprises:

[0037] Substituting the initial nuclide content of the target nuclide, the average neutron flux density of the first burnup step, and the average reaction cross section into the burnup equation, and calculating the iterative value of the nuclide content of the next burnup step;

[0038] In the iterative calculation of each subsequent burnup step, the iterative value of the nuclide content of the previous burnup step, as well as the average neutron flux density and the average reaction cross section corresponding to the current burnup step are substituted into the burnup equation to calculate the iterative value of the nuclide content of the next burnup step;

[0039] It is determined whether the last burnup step is reached. If so, the nuclide content iteration value is output, which is the nuclide content of the target nuclide at the end of the burnup life.

[0040] According to an embodiment of a third aspect of the present invention, there is provided a burnup calculation data acquisition device, comprising: a processing module, a first establishment module and a first calculation module; the processing module is used to obtain the burnup point information of the spent fuel assembly within the burnup life; the first establishment module is electrically connected to the processing module, and is used to establish a microscopic reaction rate integral average model of the target nuclide in the spent fuel assembly according to the burnup point information; the first calculation module is electrically connected to the first establishment module and the processing module, respectively, and is used to calculate the burnup calculation data of the target nuclide according to the burnup point information and the microscopic reaction rate integral average model, wherein the burnup calculation data includes an average neutron flux density value and an average reaction cross section value.

[0041] Preferably, the processing module includes an acquisition unit, an execution unit, a modeling unit and a solution unit; the acquisition unit is used to acquire the burnup life of the spent fuel assembly and the corresponding burnup depth parameter within the burnup life; the execution unit is electrically connected to the acquisition unit, and is used to divide the burnup points according to the burnup life and the corresponding burnup depth parameter within the burnup life, and obtain the division result of the burnup points, and the division result of the burnup points includes: the number of burnup points and the burnup depth corresponding to each burnup point; the modeling unit is used to establish a neutronics solution model of the spent fuel assembly; the solution unit is electrically connected to the modeling unit and the execution unit, respectively, and is used to calculate the neutron flux density value corresponding to each burnup point and the reaction cross-section value corresponding to each burnup point according to the division result of the burnup points and the neutronics solution model.

[0042] According to an embodiment of a fourth aspect of the present invention, there is provided a fuel consumption calculation system, comprising: a second establishing module, an acquiring module, a second calculating module and the above-mentioned fuel consumption calculation data acquiring device,

[0043] The second establishing module is used to establish a burnup equation of the target nuclide;

[0044] The acquisition module is used to acquire the initial nuclide content of the target nuclide;

[0045] The burnup calculation data acquisition device is used to obtain the average neutron flux density and average reaction cross section of each burnup step;

[0046] The second calculation module is electrically connected to the second establishment module, the acquisition module and the burnup calculation data acquisition device, respectively, and is used to iteratively calculate the burnup equation according to the initial nuclide content, the average neutron flux density value of each burnup step and the average reaction cross section value to obtain the nuclide content of the target nuclide at the end of the burnup life.

[0047] The method for acquiring fuel consumption calculation data in the present invention aims to solve the problem that the cross-sectional data used in the ORIGEN series programs in the prior art do not fully match the specific calculation problems, which leads to deviations in the calculation results of the same fuel consumption problem (i.e., the accuracy of the calculation results of the fuel consumption problem is low). By dividing the fuel consumption steps and adopting the integral averaging method, a more accurate fuel consumption calculation database can be calculated, thereby effectively improving the accuracy of the fuel consumption calculation.

[0048] First, the entire burnup life of the spent fuel assembly to be burnup calculated is divided into burnup points to obtain the burnup depth of each burnup point; and the existing third-party neutronics program (for example: MCNP Monte Carlo program) is used to calculate the neutron flux density and reaction cross section at each burnup point to obtain the neutron flux density and reaction cross section value of each burnup point. Specifically, by establishing a neutronics solution model for the spent fuel assembly, the neutronics solution model includes the geometric model, material properties, boundary conditions and neutron source information of the spent fuel assembly, and the neutronics model is imported into the third-party software (for example: MCNP Monte Carlo program) for solution, the cross-sectional data and flux data at each burnup point that are more in line with the actual situation of the spent fuel assembly can be obtained. Therefore, compared with the supporting database (mainly referring to the cross-sectional database) of the ORIGEN series program directly used in the prior art, this method can obtain more accurate cross-sectional data and flux data at each burnup point.

[0049] Then, according to the preset number of burnup steps (for example, 7), the corresponding burnup integral points are selected from the above burnup points, so as to divide the burnup life into multiple burnup steps. In the ignition consumption program (such as the ORIGEN series program), the more detailed the division of the burnup life (that is, the more burnup steps are obtained), the more accurate the calculation result is, however, this will lead to a large increase in calculation time. Preferably, the method will divide the burnup life into 7 burnup steps (that is, eight burnup integral points, and one burnup step is between two adjacent burnup integral points) according to the burnup depth, for example: 0, 5, 10, 15, 20, 30, 40, 50 (unit is GWd / tU) these eight burnup integral points.

[0050] According to the division results of the burnup step (i.e., eight burnup integral points and the burnup depth corresponding to each burnup integral point), a microscopic reaction rate integral average model is established for each burnup step.

[0051] Furthermore, the calculated neutron flux density and reaction cross section values ​​at each burnup point are substituted into the micro-reaction rate integral average model (the micro-reaction rate is the product of the neutron flux density and the reaction cross section), and integral average calculation is performed for each burnup step. The average neutron flux density and average reaction cross section values ​​of each burnup step obtained by the integral average calculation are input into the ORIGEN series program as an input file. The ORIGEN series program performs an iterative calculation of the initial nuclide content of the target nuclide in the spent fuel assembly based on the input file, and finally obtains a more accurate nuclide content of the target nuclide at the end of the burnup life.

[0052] It should be noted that when solving the above-mentioned micro-reaction rate integral average model, each burnup step needs to be solved separately. Specifically, the corresponding information (i.e., burnup depth, neutron flux density, reaction cross section value) of multiple burnup points corresponding to each burnup step (specifically, two adjacent burnup integral points and multiple burnup points between two burnup integral points) is substituted into the integral average model to calculate the average neutron flux density value and average reaction cross section value of each burnup step.

[0053] Furthermore, the integral average method takes into account the continuous change of the microscopic reaction rate within the burnup step, and obtains the average effect of the entire step through integral calculation, which can more accurately reflect the real change of the neutron flux density value and the reaction cross section value during the burnup process. Therefore, the average neutron flux density value and the average reaction cross section value calculated by the integral average method can better fit the burnup problem to be calculated.

[0054] Therefore, the method for acquiring fuel consumption calculation data can obtain more accurate flux density data and cross-sectional data, thereby effectively improving the accuracy of fuel consumption calculation. BRIEF DESCRIPTION OF THE DRAWINGS

[0055] Figure 1 It is a schematic diagram of the generation and disappearance of nuclide i;

[0056] Figure 2 It is an iterative calculation process of the ignition consumption calculation program in some embodiments of the present invention;

[0057] Figure 3 1 is a comparison of the mass results of actinide nuclides in some embodiments of the present invention. DETAILED DESCRIPTION

[0058] The following will be combined with the accompanying drawings in the present invention to clearly and completely describe the technical solutions in the invention. Obviously, the described embodiments are part of the embodiments of the present invention, rather than all the embodiments. Based on the embodiments in the present invention, all other embodiments obtained by ordinary technicians in this field without creative work are within the scope of the present invention.

[0059] In the description of the present invention, the terms “first” and “second” are used for descriptive purposes only and should not be understood as indicating or implying relative importance.

[0060] In the description of the present invention, each unit and module involved may correspond to only one physical structure, or may be composed of multiple physical structures, or multiple units and modules may be integrated into one physical structure; the units and modules involved may be implemented by software or by hardware, for example, the units and modules may be located in a processor.

[0061] In the description of the present invention, the functions and steps marked in the flowcharts and block diagrams of the present invention may occur in an order different from that marked in the drawings if there is no conflict.

[0062] First, it should be noted that in this application, a burnup point refers to a time node obtained by dividing the life of a spent fuel assembly according to the burnup depth, and each burnup point has a corresponding burnup depth. A burnup integral point refers to a burnup point selected when performing an integral average calculation. In other words, a burnup integral point is a part of a burnup point. In some embodiments, there are several burnup points between two adjacent burnup integral points. In this application, the time period between two adjacent burnup integral points is a burnup step.

[0063] The microscopic reaction rate is the product of the flux density value and the cross-sectional value. The integral average model in the present application is applied to this equation relationship, and therefore, the integral average model in the present application is a microscopic reaction rate integral average model.

[0064] Example 1

[0065] See also Figure 1 The present invention discloses a method for obtaining fuel consumption calculation data, comprising:

[0066] Follow these steps:

[0067] S11. Obtaining the burnup point information of the spent fuel assembly within its burnup life;

[0068] S12. Establishing a microscopic reaction rate integral average model of the target nuclides in the spent fuel assembly based on the burnup point information;

[0069] S13. According to the burnup point information and the microscopic reaction rate integral average model, the burnup calculation data of the target nuclide is calculated, and the burnup calculation data includes an average neutron flux density value and an average reaction cross section value.

[0070] It should be noted that in the radiation protection design process of spent fuel reprocessing plants, the source term calculation of spent fuel assemblies is particularly important. It is the basis of the source terms of each process in the spent fuel reprocessing plant and the basis of radiation protection design. Burnup calculation is an important type of source term calculation. Through burnup calculation, the nuclide content in the spent fuel can be obtained, which helps to evaluate the decay process of radioactive nuclides.

[0071] The ORIGEN series of programs are widely used internationally for ignition consumption and radionuclide decay programs, which can be used for the evaluation and calculation of spent fuel and high-level waste. The ORIGEN series of programs were developed by the Oak Ridge National Laboratory in the United States and have many of the same features, but they are also very different. They are the same in the method of solving the burnup equation, but the databases used (mainly including cross-section data and flux density data) are very different. At present, burnup calculations usually directly use the database that comes with the ORIGEN series of programs. Due to the differences in the supporting databases, and the problem that the database selected in the calculation does not completely match the problem being calculated, different ignition consumption calculation programs have very different results when calculating the same burnup problem.

[0072] Specifically, see Figure 1 , Figure 1 Schematic diagram of the generation and disappearance of nuclide i, showing the burnup chain of target nuclide i.

[0073] The equation for the change of the nuclear density of each nuclide in the burnup chain with time is as follows: For each nuclide in the burnup chain, the burnup equation related to it can be written, as shown in equation (1):

[0074]

[0075] Where: a,i — microscopic neutron absorption cross section of the i-th nuclide group (unit: barn); φ— single-group neutron flux density (unit: cm -2 s -1 );σ k,i — single-group microscopic neutron cross section of the kth nuclide to produce the ith nuclide (unit: barn); b j,i is the decay branch ratio of the jth nuclide to produce the ith nuclide; λ i decay is the decay constant of the ith nuclide.

[0076] Furthermore, N i (t) is the concentration (or content) of nuclide i at time t, and the right formula describes the process of generation and disappearance of nuclide i, such as Figure 1 As shown. There are two ways for the generation and disappearance of nuclide i, namely, nuclide decay and absorption. The first term on the right side of equation (1) describes the process of nuclide i transforming into another nuclide through absorption (nuclide i transforms into another nuclide by capturing neutrons) and spontaneous decay. The second term on the right side of equation (1) describes the process of nuclide j transforming into nuclide i through spontaneous decay, where N j (t) is the concentration of nuclide j; the third term on the right side of equation (1) describes the process of nuclide k transforming into nuclide i by absorbing neutrons, where N k(t) is the concentration of nuclide k.

[0077] It should be noted that, in this embodiment, i, j, and k are merely identification codes used to represent a certain target nuclide, so as to facilitate the description of the burnup chain.

[0078] The above equation (1) can be written in a more compact matrix form as follows:

[0079]

[0080] Where: —Matrix of nuclide density changing with time; A—burnup matrix; N(t)—matrix of nuclide density.

[0081] The solution of equation (2) can be expressed as follows with the help of matrix exponentials:

[0082] N(t)=e tA ·N(0)

[0083] Where: N(0)—nuclear density at the initial moment / b -1 cm -1 .

[0084] In summary, it can be understood that equations (1) and (2) are the burnup equations, that is, the equations to be solved by the ignition consumption calculation program (ORIGEN series program). The result of the solution is the nuclide density of the target nuclide in the spent fuel.

[0085] In this embodiment, the method for acquiring fuel consumption calculation data is designed to address the problem that the cross-sectional data used in the ORIGEN series programs in the prior art do not fully match the specific calculation problem, which in turn leads to deviations in the calculation results for the same fuel consumption problem (i.e., the accuracy of the calculation results of the fuel consumption problem is low). By dividing the fuel consumption steps and using the integral averaging method, a more accurate fuel consumption calculation database can be calculated, thereby effectively improving the accuracy of the fuel consumption calculation.

[0086] First, the entire burnup life of the spent fuel assembly to be burnt up is divided into burnup points to obtain the burnup depth of each burnup point. In the present embodiment, the burnup life of the spent fuel assembly can be divided into 50 burnup points (the burnup points here can be evenly divided). Then the existing third-party neutronics program (for example: MCNP Monte Carlo program) is used to calculate the neutron flux density and the reaction cross section at each burnup point to obtain the neutron flux density and the reaction cross section value of each burnup point. Specifically, by setting up a neutronics solution model for the spent fuel assembly, the neutronics solution model includes the geometric model, material properties and boundary conditions and neutron source information of the spent fuel assembly, and the neutronics model is imported into the third-party software (for example: MCNP Monte Carlo program) for solution, and the cross-sectional data and flux data under each burnup point that are more in line with the actual situation of the spent fuel assembly can be obtained. Therefore, compared with the prior art that directly uses the supporting database of the ORIGEN series programs (mainly the cross-section database), this method can obtain more accurate cross-section data and flux data at each burnup point.

[0087] In other words, in this embodiment, step S11, obtaining the burnup point information of the spent fuel assembly within the burnup life, includes:

[0088] Obtain the burnup life of the spent fuel assembly and the corresponding burnup depth parameters within the burnup life.

[0089] The burnup points are divided according to the burnup life and the corresponding burnup depth parameters within the burnup life to obtain the burnup point division results, which include: the number of burnup points, the corresponding burnup depths of each burnup point (i.e., the burnup life is divided into multiple burnup steps according to the burnup depth).

[0090] Establish a neutronics solution model for spent fuel assemblies.

[0091] According to the division results of the burnup points and the neutronics solution model, the neutron flux density value corresponding to each burnup point and the reaction cross section value corresponding to each burnup point are calculated.

[0092] It should be noted that the burnup points can be divided by the staff. After the burnup points are divided, the results of the burnup point division (i.e., the number of burnup points and the corresponding burnup depth of each burnup point pair) are manually input into the third-party neutronics program. Then, the staff establishes a neutronics solution model, i.e., establishes a geometric model of the spent fuel assembly, sets the boundary conditions, material properties and neutron source information of the geometric model (spent fuel assembly), etc. Next, the neutronics solution model is imported into the third-party neutronics program for solution to obtain the neutron flux density corresponding to each burnup point and the reaction cross section corresponding to each burnup point, i.e., the burnup point information is obtained.

[0093] Furthermore, the burnup point information described in step S11 includes the number of burnup points, which is multiple (50-100), as well as the burnup depth corresponding to each burnup point, the neutron flux density value corresponding to each burnup point, and the reaction cross section value corresponding to each burnup point.

[0094] Then, according to the preset number of burnup steps (for example, 7-29), corresponding burnup integral points are selected from the above burnup points, so as to divide the burnup life into multiple burnup steps. Furthermore, the conventional burnup point division process is usually divided into 29 steps. The 29 burnup steps here are 30 burnup points selected from the burnup depth range of 0GWd / tU to 55GWd / tU, namely: 0.01, 0.10, 0.50, 1.00, 2.00, 3.00, 4.00, 5.00, 6.00, 7.00, 8.00, 9.00, 10.00, 12.50, 15.00, 17.50, 20.00, 22.50, 25.00, 27.50, 30.00, 32.50, 35.00, 37.50, 40.00, 42.50, 45.00, 47.50, 50.00 (unit: GWd / tU). In this embodiment, it is found through simulation calculation that when the burnup step is selected as 7 steps, the ignition consumption calculation result of the transport program (i.e., the third-party neutronics program, such as the MCNP Monte Carlo program) can well reproduce the result when the integral point is 29 steps, and the maximum relative deviation of the nuclide mass obtained by 7 steps and 29 steps is 1.52%. In other words, the burnup life of the spent fuel assembly is divided into seven steps, and the cross-sectional data and flux density data are obtained after calculation by the third-party neutronics program. After the cross-sectional data and flux density data are substituted into the ORIGEN series program, the relative deviation of the mass of the target nuclide obtained by iterative calculation is within 1.52%, which can meet the accuracy requirements of the burnup calculation. Therefore, the 7-step micro-reaction rate integral average value calculated by the transport analysis program is provided to the ignition consumption calculation program for use. The 8 burnup points corresponding to the 7 burnup steps are: 0, 5, 10, 15, 20, 30, 40, 50 (unit: GWd / tU).

[0095] In this embodiment, the number of burnup steps may range from 7 to 29, for example, 7 burnup steps, 15 burnup steps, or 29 burnup steps. Of course, 7 burnup steps are sufficient to obtain accurate nuclide masses, so in this embodiment, the number of burnup steps is preferably 7.

[0096] It is easy to understand that in the ignition consumption program (such as the ORIGEN series program), the more detailed the division of the burnup life (that is, the more burnup steps are obtained), the more accurate the calculation result is. However, dividing too many burnup steps will lead to a large increase in calculation time. Preferably, the method will divide the burnup life into 7-11 burnup steps according to the burnup depth. Exemplarily, the method will divide the burnup life into 7 burnup steps (that is, eight burnup points, one burnup step between two adjacent burnup points) according to the burnup depth, for example: 0, 5, 10, 15, 20, 30, 40, 50 (unit is GWd / tU) these eight burnup points.

[0097] It should be noted that the change trend of the nuclide content at the lower burnup point is usually more dramatic, so the burnup integral points selected at the lower burnup point are more dense to better fit the change trend of the nuclide. In other words, the preset strategy for selecting the burnup integral points is to select the integral points based on the uniformity of the burnup depth, that is, to give priority to selecting points in areas where the burnup depth changes significantly. Of course, in other embodiments, other methods of dividing the burnup steps can also be used, such as dividing into seven burnup steps at equal intervals.

[0098] In this embodiment, according to the division results of the burnup step (ie, eight burnup integration points and the burnup depth corresponding to each burnup integration point), a microscopic reaction rate integral average model is established for each burnup step.

[0099] Specifically, in this embodiment, step S12 specifically includes:

[0100] According to the information of the burn-up points, a burn-up integral point is selected from the burn-up points, and the distance between two adjacent burn-up integral points is a burn-up step;

[0101] The microscopic reaction rate integral average model of the target nuclide is established for each burnup step.

[0102] The microscopic reaction rate integral average model of any burnup step is:

[0103]

[0104]

[0105]

[0106] In the formula, is the average neutron flux density of the corresponding burnup step between two adjacent burnup integration points, φ i is the neutron flux density at each burnup point in the burnup step, is the average reaction rate of the burnup step, R i is the microscopic reaction rate of each burnup point in the burnup step, is the average reaction cross section of the burnup step, BU j is the burnup depth of each burnup point in the burnup step, startBU is the lower bound of the integration interval, stopBU is the upper bound of the integration interval, and num_integral is the number of integration points.

[0107] Then, according to the calculated neutron flux density value and reaction cross section value at each burnup point, they are substituted into the microscopic reaction rate integral average model (the microscopic reaction rate is the product of the neutron flux density value and the reaction cross section value) to perform integral average calculation. Specifically, step S13, according to the burnup point information and the microscopic reaction rate integral average model, the burnup calculation data of the target nuclide is calculated, including:

[0108] According to the burnup point information, the integral point information corresponding to each burnup step is obtained, wherein the integral point information includes: the neutron flux density of each burnup point in the burnup step, the microscopic reaction rate of each burnup point in the burnup step, and the burnup depth of each burnup point in the burnup step;

[0109] The integral point information corresponding to each burnup step is substituted into the microscopic reaction rate integral average model to calculate the average neutron flux density and average reaction cross section of the target nuclide in each burnup step.

[0110] The average neutron flux density value and average reaction cross section value of each burnup step obtained by the integral average calculation are input into the ORIGEN series of programs as input files. Based on the input files, the ORIGEN series of programs perform iterative calculations of the initial nuclide content of the target nuclides in the spent fuel assembly for each burnup step, and ultimately obtain a more accurate nuclide content of the target nuclides at the end of the burnup life.

[0111] It should be noted that when solving the above-mentioned micro-reaction rate integral average model, each burnup step needs to be solved separately. Specifically, the corresponding information (i.e., burnup depth, neutron flux density, reaction cross section value) of multiple burnup points corresponding to each burnup step (specifically, two adjacent burnup integral points and multiple burnup points between two burnup integral points) is substituted into the integral average model to calculate the average neutron flux density value and average reaction cross section value of each burnup step.

[0112] This method obtains the flux data and cross-section data of each burnup point, and then performs an integral average calculation for each burnup step to obtain the microscopic reaction rate corresponding to the target integration point. Initially, the flux and cross-section data of a large number of burnup points were given by a third-party program. If too many burnup points are used, the complexity of the script program will increase, and the improvement of the calculation accuracy will be of little significance. Therefore, this model can be used to obtain the microscopic reaction rate at the target integration point (for example, 29 and 7 are used for comparison analysis here).

[0113] It should be noted that the integral average method takes into account the continuous change of the microscopic reaction rate within the burnup step, and the average effect of the entire step is obtained through integral calculation, which can more accurately reflect the real change of the neutron flux density value and the reaction cross section value during the burnup process. Therefore, the average neutron flux density value and the average reaction cross section value calculated by the integral average method can be more suitable for the burnup problem to be calculated.

[0114] Therefore, the method for acquiring fuel consumption calculation data can obtain more accurate flux density data and cross-sectional data, thereby effectively improving the accuracy of fuel consumption calculation.

[0115] Example 2

[0116] The present invention also discloses a fuel consumption calculation method, comprising the following steps:

[0117] S21. Establish the burnup equation of the target nuclide.

[0118] S22. Obtain the initial nuclide content of the target nuclide.

[0119] S23. According to the burnup calculation data acquisition method in Example 1, the average neutron flux density and average reaction cross section of each burnup step are obtained.

[0120] S24. According to the initial nuclide content, the average neutron flux density value of each burnup step and the average reaction cross section value, the burnup equation is iteratively calculated to obtain the nuclide content of the target nuclide at the end of the burnup life.

[0121] It should be noted that if Figure 1 As shown, the burnup chain of target nuclide i. The equation for the change of the nuclear density of each nuclide in the burnup chain over time is as follows: For each nuclide in the burnup chain, the burnup equation related to it can be written. The burnup equation of the target nuclide is:

[0122]

[0123] Where: N i (t) is the content of target nuclide i at time t; σ a,i is the microscopic neutron absorption cross section of the i-th nuclide group (unit: barn); φ is the single group neutron flux density (unit: cm -2 s -1 );σ k,i is the single-group microscopic neutron cross section for the kth nuclide to produce the ith nuclide (unit: barn); b j,i is the decay branch ratio of the jth nuclide to produce the ith nuclide; is the decay constant of the ith nuclide.

[0124] The burnup calculation method in this embodiment is used to solve the burnup equation to obtain the nuclide content of the target nuclide.

[0125] Since the half-life of nuclide decay spans from nanoseconds to infinity, the problem is very rigid, so directly solving the (burnup) equation will lead to serious numerical instability. For solving the burnup equation, the ORIGEN series of programs use the Taylor expansion truncation method, which has high computational efficiency and can ensure that the accuracy of the main nuclides that are critical to transport calculations is within a reasonable and controllable range.

[0126] The Taylor expansion and truncation method, that is, the burnup calculation method based on Taylor expansion and truncation, refers to approximating the burnup equation through the principle of Taylor series expansion and simplifying the calculation process through truncation technology. This method separates long- and short-lived nuclides and calculates them separately to reduce the rigidity of the burnup matrix, thereby increasing the calculation accuracy and stability.

[0127] The ORIGEN series of programs include ORIGEN2, ORIGEN-S, ORIGEN-ARP and other versions. The ORIGEN series of programs have an independent burnup database and a series of problem-related cross-section data, fission yield database and photon database. The burnup database of the ORIGEN program stores the decay constants and reaction channel branching ratios required for burnup calculations; the cross-section and fission yield databases of the ORIGEN program are problem-related. When using the ORIGEN program, the cross-section database needs to be selected based on information such as reactor type, fuel enrichment, and maximum burnup depth.

[0128] In this embodiment, the VERA-3a problem in the VERA benchmark problem is a benchmark problem for simulating complex physical processes of a specific type of reactor (such as a pressurized water reactor, a boiling water reactor, etc.), which is intended to test the ability of the calculation program in processing key physical phenomena such as neutron transport, thermal hydraulic coupling, and fuel burnup. In this embodiment, taking the ignition consumption calculation of a three-dimensional pressurized water reactor group as an example, the cross-section database provided by the ORIGEN series program is used to calculate the ignition consumption of the three-dimensional pressurized water reactor component. The results of the important actinide nuclide content obtained by different ignition consumption programs have a large deviation. Therefore, the burnup calculation method in this embodiment introduces a reliable third-party program (i.e., a neutronics program, which can perform accurate and reliable cross-section processing and transport calculation) to provide the ORIGEN series program with high-precision cross-sections (i.e., cross-section data and flux density data obtained by the burnup calculation data acquisition method in Example 1), so that the maximum deviation of the actinide nuclide mass and radioactivity results calculated by different ignition consumption calculation programs is less than 2%. The above-mentioned neutronics program may adopt an existing program as long as it can perform accurate and reliable cross-section processing and transport calculations, such as: MCNP (Monte Carlo) program and the like.

[0129] Specifically, the burnup calculation method in this embodiment introduces a neutron calculation program to accurately model the geometry of the entire core, and can automatically perform accurate cross-section processing and transport calculations, thereby providing the ignition consumption calculation program with the neutron flux at each burnup point and the microscopic reaction cross section of each nuclide (i.e., step 23).

[0130] It can be seen that the starting point of the burnup calculation method in this embodiment is that the calculation results of the ORIGEN series of programs for the same burnup problem have deviations. After in-depth analysis, it is found that the reason is that the ORIGEN program is an ignition consumption program. Because the cross-sectional data used are different, and the cross-sectional data used are not completely matched with the specific calculation problem, it fails to provide the microscopic reaction rate related to the problem. Due to the large difference in microscopic reaction rate, the calculation results of different ignition consumption calculation programs have large deviations. During the entire life cycle, the ORIGEN program is approximate in the solution process of the burnup equation, that is, the microscopic reaction rate is considered to be a constant in two adjacent burnup steps, and how to correctly obtain the microscopic reaction rate is the key to the burnup calculation. In the coupled system of the physical program coupled with the ignition consumption program, the burnup steps are divided very finely, or a certain estimation and correction strategy is adopted, so that a more accurate microscopic reaction rate within two burnup steps can be obtained. Therefore, in order to improve the calculation accuracy of actinide nuclides, it is necessary to provide the ignition consumption program with the correct microscopic reaction rate. The implementation plan of this method is:

[0131] 1) Accurately calculate the burnup of the single-element problem through a third-party calculation program to obtain the microscopic reaction rate of the target nuclide between each burnup point;

[0132] 2) Within the burnup life calculated by the third-party calculation program, several burnup points are selected as integration points, and the microscopic reaction rates within two adjacent integration points are integrated and averaged according to time to obtain the integrated average value of the microscopic reaction rates of the target nuclide within the integration points;

[0133] 3) The obtained integrated average value of the microscopic reaction rate is provided to the ignition consumption calculation program to obtain the calculation result.

[0134] Based on this scheme, a source term analysis script was developed. It uses the single group flux at each burnup point calculated by the neutronics program and the reaction cross section of the nuclide to obtain the integrated average neutron flux density and the integrated average microscopic cross section according to the "microscopic reaction rate integrated average method", and then uses the average cross section to generate the cross section replacement program input file for each burnup point (cross section replacement), and then uses the average flux to generate the ignition consumption calculation program input file for each burnup point (burnup calculation), and finally runs the program to obtain the ignition consumption calculation program output file for each integral point, that is, to obtain the source term calculation results of the analyzed component. The flowchart of the script is as follows Figure 1 The cross-section replacement program is used to process the binary nuclear database required by the ignition consumption calculation program, so as to calculate the nuclide concentration and related radiation sources and decay heat during irradiation and decay.

[0135] See also Figure 2 , the text coupling process of the ignition consumption calculation program is as follows Figure 2 shown.

[0136] In addition, the calculation found that when the burnup step is selected as 7 steps, the ignition consumption calculation results of the transport program can well reproduce the results when the integration point is 29 steps. The maximum relative deviation of the nuclide mass obtained by 7 steps and 29 steps is 1.52%. Therefore, the 7-step micro-reaction rate integral average calculated by the transport analysis program is provided to the ignition consumption calculation program. The 29 burnup steps here are 30 burnup points selected from the burnup depth range of 0GWd / tU to 55GWd / tU, namely: 0.01, 0.10, 0.50, 1.00, 2.00, 3.00, 4.00, 5.00, 6.00, 7.00, 8.00, 9.00, 10.00, 12.50, 15.00, 17.50, 20. 00, 22.50, 25.00, 27.50, 30.00, 32.50, 35.00, 37.50, 40.00, 42.50, 45.00, 47.50, 50.00 (unit: GWd / tU), the 8 burnup points corresponding to the 7 burnup steps are: 0, 5, 10, 15, 20, 30, 40, 50 (unit: GWd / tU).

[0137] In other words, the third-party calculation program divides the precise burnup calculation into 29 steps in total. When the number of integration points is also 29, the ignition consumption module of the third-party program performs 29 burnup calculations. Compared with the 29-step precise transport-burnup coupling calculation performed by the third-party program, the maximum relative deviation of the nuclide mass is 1.52%. The source of the error should be the integration method surface, not the division accuracy of the burnup step; that is, when the number of integration points is selected as 7 steps, the ignition consumption calculation of the third-party program performs 7-step burnup calculation, which can well restore the result of 29 integration points. Therefore, the 7-step micro-reaction rate integral average of the third-party program is selected and provided to ORIGEN-S.

[0138] The use of the source analysis script is divided into the following four steps:

[0139] 1) Generation of 7-step integrated average neutron flux density and cross section:

[0140] The third-party calculation program performs burnup calculation on the component, and can obtain the average single-group neutron flux and single-group reaction cross section of the entire component at each burnup point, which are used as the basic input of the source term analysis script program. In other words, step S23, according to the burnup calculation data acquisition method in Example 1, obtain the average neutron flux density and average reaction cross section of each burnup step.

[0141] The average flux and average cross section are calculated using the microscopic reaction rate integrated average method:

[0142]

[0143]

[0144]

[0145] In the formula, is the average neutron flux density between two burnup integration points (i.e., burnup steps), φ i is the neutron flux density at each burnup point between two burnup integration points, is the average reaction rate between two burnup integration points (i.e., burnup steps), R i is the microscopic reaction rate at each burnup point between two burnup integration points, is the average microscopic cross section between two burnup integration points (i.e., burnup steps), BU j It is the burnup depth of each corresponding (burnup point) between two burnup integration points, startBU is the lower bound of the integration interval, stopBU is the upper bound of the integration interval, and num_integral is the number of integration points.

[0146] It should be noted that there is a fuel step between every two adjacent fuel integration points. In other words, the fuel points corresponding to the two ends of each fuel step are fuel integration points, and there are multiple fuel points between two adjacent fuel integration points. Each fuel point between two fuel integration points can be understood as multiple fuel points between the two fuel integration points.

[0147] 2) Generate the input file of the section replacement program at each integration point:

[0148] The section replacement program input file for each integration point is generated based on the average section obtained in the first step. The function of this input file is to replace the section in the section library used by the point section calculation program with the problem-related section given by the third-party program.

[0149] 3) Generate the input file of the ignition consumption calculation program at each integration point:

[0150] The input file of the ignition consumption calculation program for each integration point is generated according to the average neutron flux density obtained in the first step.

[0151] 4) Run the section replacement program and the ignition consumption calculation program to obtain the source term calculation results:

[0152] This step is Figure 2 The iterative process of the 7 burnup steps is controlled by a batch script, the purpose of which is to use the database and nuclear density obtained by running the cross section replacement program and the ignition consumption calculation program at each step as the input file for the next step. By continuously replacing the cross section and flux of the ignition consumption calculation program through the script, the mass results of the actinide nuclides after replacing the cross section and flux of the 7 steps can be obtained.

[0153] In other words, S24, based on the initial nuclide content, the average neutron flux density value of each burnup step and the average reaction cross section value, the burnup equation is iteratively calculated to obtain the nuclide content of the target nuclide at the end of the burnup life.

[0154] Specifically, step S24 includes:

[0155] Substituting the initial nuclide content of the target nuclide, the average neutron flux density of the first burnup step, and the average reaction cross section into the burnup equation, and calculating the iterative value of the nuclide content of the next burnup step;

[0156] In the iterative calculation of each subsequent burnup step, the iterative value of the nuclide content of the previous burnup step, as well as the average neutron flux density and the average reaction cross section corresponding to the current burnup step are substituted into the burnup equation to calculate the iterative value of the nuclide content of the next burnup step;

[0157] It is determined whether the last burnup step is reached. If so, the nuclide content iteration value is output, which is the nuclide content of the target nuclide at the end of the burnup life.

[0158] like Figure 2As shown, the time-integrated average cross section (i.e., the average reaction cross section value of each burnup step mentioned above) is used to generate a cross section replacement file with the number of integral points; and the time-integrated average flux (i.e., the average neutron flux density value of each burnup step mentioned above) and the integrated ignition consumption (i.e., the burnup depth corresponding to the burnup integration point) are used to generate an ignition consumption calculation file with the number of integral points. The batch script generates and runs iterative calculations. First, the first iteration is performed, that is, the average neutron flux density value and the average reaction cross section value are substituted into the burnup equation as the input values ​​of the single-group neutron flux density and the single-group microscopic neutron cross section of the first burnup step, respectively, to calculate the input values ​​of the single-group neutron flux density and the single-group microscopic neutron cross section of the next burnup step. The input values ​​of the single-group neutron flux density and the single-group microscopic neutron cross section are iterated to the next burnup step, and the cross section replacement program is run to obtain the cross section library of the next burnup step. The ignition consumption calculation program is run to obtain the atomic nucleus density file, and density_pass is run to replace the new atomic nucleus density into the input file of the ignition consumption calculation program. Determine whether the last burnup step has been reached. If not, continue the iteration; if so, end the iteration and output the final nuclear density file, that is, obtain the nuclide content of the target nuclide.

[0159] For example, for nuclides such as Am-241 and Am-243, which can produce ground state and excited state respectively through radiation capture reaction, their radiation capture cross section replacement method is different from that of other nuclides. The reaction channel cross sections for producing ground state and excited state nuclides should be given separately according to the reaction channel branching ratio. In addition, if there are still some nuclides with large deviations in the calculated results, it is necessary to compare the decay constants of these nuclides and replace them so that the decay constants used by the two ignition consumption calculation programs are consistent. After the above processing, the maximum deviation of the actinide nuclide mass and radioactivity results calculated by the two ignition consumption calculation programs is less than 2%. Figure 3 As shown, the table lists the comparison of the calculation results of two different ignition consumption programs before and after using the present invention (the table gives the relative deviation of the calculation results of two different ignition consumption programs for decay of 0 days, 1 year, 5 years, 10 years, and 30 years). Through the application and comparative analysis of the method in this embodiment, it is found and proved that the method is helpful to improve the accuracy of the calculation of the ignition consumption program.

[0160] It should be noted that Figure 3 The 0 day in the table shown in is the nuclide content of the target nuclide (i.e., actinide nuclide) at the end of the burnup life calculated in this embodiment. The ORIGEN-S program can continue to iterate based on the nuclide content of the target nuclide on day 0, and finally obtain the decay results of actinide nuclides for 1 year, 5 years, 10 years, and 30 years.

[0161] It can be seen that the result calculated according to the fuel consumption calculation method in this embodiment can effectively reduce the relative deviation, thereby significantly improving the accuracy of the fuel consumption calculation result.

[0162] Example 3

[0163] The invention also discloses a fuel consumption calculation data acquisition device, comprising: a processing module, a first establishing module and a first calculating module.

[0164] The processing module is used to obtain the burnup point information of the spent fuel assembly during its burnup life. The first establishment module is electrically connected to the processing module, and is used to establish a microscopic reaction rate integral average model of the target nuclide in the spent fuel assembly according to the burnup point information. The first calculation module is electrically connected to the first establishment module and the processing module respectively, and is used to calculate the burnup calculation data of the target nuclide according to the burnup point information and the microscopic reaction rate integral average model, and the burnup calculation data includes the average neutron flux density value and the average reaction cross section value.

[0165] It should be noted that the processing module, the first establishing module and the first calculating unit may all be program modules constructed on a computer.

[0166] Specifically, the first building module can be an import unit constructed by programming on a computer, and the staff imports the established micro-reaction rate integral average model into the first building module.

[0167] In this embodiment, the above-mentioned burnup point information includes the number of burnup points, the number of burnup points is multiple, and the burnup point information also includes the burnup depth corresponding to each burnup point, the neutron flux density value corresponding to each burnup point, and the reaction cross-section value corresponding to each burnup point.

[0168] The processing module includes an acquisition unit, an execution unit, a modeling unit and a solution unit. The acquisition unit is used to acquire the burnup life of the spent fuel assembly and the corresponding burnup depth parameter within the burnup life; the execution unit is electrically connected to the acquisition unit and is used to divide the burnup points according to the burnup life and the corresponding burnup depth parameter within the burnup life, and obtain the division result of the burnup points, and the division result of the burnup points includes: the number of burnup points and the burnup depth corresponding to each burnup point; the modeling unit is used to establish a neutronics solution model for the spent fuel assembly; the solution unit is electrically connected to the modeling unit and the execution unit respectively, and is used to calculate the neutron flux density value corresponding to each burnup point and the reaction cross-section value corresponding to each burnup point according to the division result of the burnup points and the neutronics solution model.

[0169] Among them, the acquisition unit can be an input module programmed and constructed on a computer, and the staff manually inputs the burnup life of the spent fuel assembly and the corresponding burnup depth parameters within the burnup life. The execution unit can be a division module programmed and constructed on a computer, and the staff divides the burnup steps (i.e., the burnup points) according to the burnup life and burnup depth parameters. In this embodiment, it is divided into 7 burnup steps, i.e., 8 burnup integral points. Of course, a more detailed division can be performed. In order to balance the calculation accuracy and the time for solving the calculation, the division range of the burnup steps is (7-29). The modeling unit can be an import unit programmed and constructed on a computer, and the staff pre-stores the established neutronics solution model into the modeling unit. The solution unit can use an existing third-party neutronics program, such as: MCNP (Monte Carlo) program.

[0170] Further, the first calculation module includes a first processing unit and a second processing unit, wherein the first processing unit is electrically connected to the processing module, and is used to obtain the integral point information of two burnup integral points corresponding to each burnup step according to the burnup point information, and the integral point information includes: the burnup depth corresponding to the burnup integral point, the neutron flux density value corresponding to the burnup integral point, and the reaction cross section value corresponding to the burnup integral point. The second processing unit is electrically connected to the first processing unit and the first establishment module, and is used to substitute the integral point information of the two burnup integral points corresponding to each burnup step into the microscopic reaction rate integral average model, and calculate the average neutron flux density value and the average reaction cross section value of the target nuclide in each burnup step.

[0171] In summary, the fuel consumption calculation data acquisition device in this embodiment can calculate a more accurate fuel consumption calculation database by dividing the fuel consumption steps and adopting the integral averaging method, thereby effectively improving the accuracy of the fuel consumption calculation.

[0172] Example 4

[0173] A fuel consumption calculation system is provided, comprising: a second establishing module, an acquiring module, a second calculating module and the fuel consumption calculation data acquiring device in embodiment 3.

[0174] It should be noted that the second establishment module is used to establish the burnup equation of the target nuclide. The acquisition module is used to obtain the initial nuclide content of the target nuclide. The above-mentioned burnup calculation data acquisition device is used to obtain the average neutron flux density and the average reaction cross section of each burnup step. The second calculation module is electrically connected to the second establishment module, the acquisition module and the burnup calculation data acquisition device, respectively, and is used to iteratively calculate the burnup equation according to the initial nuclide content, the average neutron flux density value of each burnup step and the average reaction cross section value, so as to obtain the nuclide content of the target nuclide at the end of the burnup life.

[0175] It should be noted that the second establishing module, the obtaining module and the second calculating module can all be program modules constructed by programming on a computer.

[0176] The second establishment module can be an input module programmed and constructed on a computer, and the staff will pre-store the established burnup equation of the target nuclide into the computer. The acquisition module can be an input module on the computer, and the staff will manually input the initial nuclide content of the acquired target nuclide into the acquisition module. The second calculation module can use the ORIGEN series program to perform iterative calculations based on the initial nuclide content, the average neutron flux density value of each burnup step (as an input file of flux density data) and the average reaction cross section value (as an input file of cross section data) to obtain the nuclide content of the target nuclide at the end of the burnup life.

[0177] In other words, the present burnup calculation system adopts the burnup calculation data acquisition device in Example 3, wherein the third-party calculation program performs accurate burnup calculation on the single-element problem to obtain the microscopic reaction rate of the target nuclide between each burnup point. During the burnup life calculated by the third-party calculation program, several burnup points are selected as integral points, and the microscopic reaction rates within two adjacent integral points are integrated and averaged according to time (that is, each burnup step is integrated and averaged), and the integral average value of the microscopic reaction rate of the target nuclide within the integral point is obtained. Finally, the obtained integral average value of the microscopic reaction rate is provided to the ignition consumption calculation program for calculation to obtain the calculation result.

[0178] Specifically, the second calculation module includes an initialization unit, an iterative calculation unit and a judgment unit. The initialization unit is used to substitute the initial nuclide content of the target nuclide, the average neutron flux density of the first burnup step and the average reaction cross section into the burnup equation to calculate the iterative value of the nuclide content of the next burnup step.

[0179] The iterative calculation unit is electrically connected to the initialization unit, and is used to substitute the iterative value of the nuclide content of the previous burnup step, and the average neutron flux density and the average reaction cross section corresponding to the current burnup step into the burnup equation in the iterative calculation of each subsequent burnup step, so as to calculate the iterative value of the nuclide content of the next burnup step.

[0180] The judgment unit is electrically connected to the iterative calculation unit and is used to judge whether the last burnup step has been reached. If so, an output signal is sent to the iterative calculation unit. The iterative calculation unit outputs a nuclide content iteration value according to the output signal, which is the nuclide content of the target nuclide at the end of the burnup life.

[0181] like Figure 3 As shown in the table in , the fuel consumption calculation system in this embodiment can reduce the relative deviation and significantly improve the accuracy of the fuel consumption calculation (ie, source term calculation) results.

[0182] It is to be understood that the above embodiments are merely exemplary embodiments used to illustrate the principles of the present invention, but the present invention is not limited thereto. For those of ordinary skill in the art, various modifications and improvements can be made without departing from the spirit and essence of the present invention, and these modifications and improvements are also considered to be within the scope of protection of the present invention.

Claims

1. A method for obtaining fuel consumption calculation data, characterized in that: The steps include: S11. Obtaining the burnup point information of the spent fuel assembly within its burnup life; S12. Establishing a microscopic reaction rate integral average model of target nuclides in the spent fuel assembly according to the burnup point information; S13. Calculate the burnup calculation data of the target nuclide according to the burnup point information and the microscopic reaction rate integral average model, wherein the burnup calculation data includes an average neutron flux density value and an average reaction cross section value.

2. The method according to claim 1, characterized in that The burnup point information includes the number of burnup points, which may be multiple. The burnup point information also includes the burnup depth corresponding to each burnup point, the neutron flux density value corresponding to each burnup point, and the reaction cross section value corresponding to each burnup point.

3. The method according to claim 1, characterized in that The step S11, obtaining the burnup point information of the spent fuel assembly within the burnup life, comprises: Obtaining the burnup life of the spent fuel assembly and a burnup depth parameter corresponding to the burnup life; According to the burnup life and the corresponding burnup depth parameter within the burnup life, the burnup points are divided to obtain a burnup point division result, wherein the burnup point division result includes: the number of burnup points and the burnup depth corresponding to each burnup point; Establishing a neutronics solution model for the spent fuel assembly; According to the division result of the burnup points and the neutronics solution model, the neutron flux density value corresponding to each burnup point and the reaction cross section value corresponding to each burnup point are calculated.

4. The method according to claim 3, characterized in that: The neutronics solution model includes: a geometric model of the spent fuel assembly, material properties, boundary conditions and neutron source information.

5. The method according to any one of claims 1 to 4, characterized in that: The S12, establishing a microscopic reaction rate integral average model of the target nuclides in the spent fuel assembly according to the burnup point information, comprises: According to the burnup point information, a burnup integral point is selected from the burnup points, and a burnup step is formed between two adjacent burnup integral points; Establish the microscopic reaction rate integral average model of the target nuclide for each burnup step; The microscopic reaction rate integral average model of any burnup step is: In the formula, is the average neutron flux density of the corresponding burnup step between two adjacent burnup integration points, φ i is the neutron flux density at each burnup point in the burnup step, is the average reaction rate of the burnup step, R i is the microscopic reaction rate of each burnup point in the burnup step, is the average reaction cross section of the burnup step, BU j is the burnup depth of each burnup point in the burnup step, startBU is the lower bound of the integration interval, stopBU is the upper bound of the integration interval, and num_integral is the number of integration points.

6. The method according to claim 5, characterized in that The step S13, according to the burnup point information and the microscopic reaction rate integral average model, calculates the burnup calculation data of the target nuclide, wherein the burnup calculation data includes an average neutron flux density value and an average reaction cross section value, including: According to the burnup point information, integral point information corresponding to each burnup step is obtained, wherein the integral point information includes: neutron flux density of each burnup point in the burnup step, microscopic reaction rate of each burnup point in the burnup step, and burnup depth of each burnup point in the burnup step; The integral point information corresponding to each burnup step is respectively substituted into the microscopic reaction rate integral average model to calculate the average neutron flux density value and the average reaction cross section value of the target nuclide in each burnup step.

7. A method for calculating fuel consumption, characterized in that: The steps include: S21. Establish the burnup equation of the target nuclide; S22, obtaining the initial nuclide content of the target nuclide; S23. According to the method for obtaining burnup calculation data according to any one of claims 1 to 6, obtaining an average neutron flux density and an average reaction cross section for each burnup step; S24. According to the initial nuclide content, the average neutron flux density value of each burnup step and the average reaction cross section value, the burnup equation is iteratively calculated to obtain the nuclide content of the target nuclide at the end of the burnup life.

8. The method according to claim 7, characterized in that The burnup equation of the target nuclide is: Where: N i (t) is the content of target nuclide i at time t; σ a,i is the microscopic neutron absorption cross section of the i-th nuclide single group; φ is the single-group neutron flux density; σ k,i is the single group microscopic neutron cross section of the kth nuclide to produce the ith nuclide; b j,i is the decay branch ratio of the jth nuclide to produce the ith nuclide; is the decay constant of the ith nuclide.

9. The method according to claim 8, characterized in that The step S24, iteratively calculating the burnup equation according to the initial nuclide content, the average neutron flux density of each burnup step, and the average reaction cross section to obtain the nuclide content of the target nuclide at the end of the burnup life, comprises: substituting the initial nuclide content of the target nuclide, the average neutron flux density of the first burnup step, and the average reaction cross section into the burnup equation to calculate the iterative value of the nuclide content of the next burnup step; In the iterative calculation of each subsequent burnup step, the iterative value of the nuclide content of the previous burnup step, as well as the average neutron flux density and the average reaction cross section corresponding to the current burnup step are substituted into the burnup equation to calculate the iterative value of the nuclide content of the next burnup step; It is determined whether the last burnup step is reached. If so, the nuclide content iteration value is output, which is the nuclide content of the target nuclide at the end of the burnup life.

10. A fuel consumption calculation data acquisition device, characterized in that: include: a processing module, a first establishing module and a first calculating module; The processing module is used to obtain the burnup point information of the spent fuel assembly during its burnup life; The first establishing module is electrically connected to the processing module and is used to establish a microscopic reaction rate integral average model of the target nuclide in the spent fuel assembly according to the burnup point information; The first calculation module is electrically connected to the first establishment module and the processing module, respectively, and is used to calculate the burnup calculation data of the target nuclide according to the burnup point information and the microscopic reaction rate integral average model, and the burnup calculation data includes an average neutron flux density value and an average reaction cross section value.

11. The device according to claim 10, characterized in that The burnup point information includes the number of burnup points, the number of burnup points is multiple, and the burnup point information also includes the burnup depth corresponding to each burnup point, the neutron flux density value corresponding to each burnup point, and the reaction cross section value corresponding to each burnup point; The processing module includes an acquisition unit, an execution unit, a modeling unit and a solution unit; The acquisition unit is used to acquire the burnup life of the spent fuel assembly and the burnup depth parameter corresponding to the burnup life; The execution unit is electrically connected to the acquisition unit, and is used to divide the burnup points according to the burnup life and the corresponding burnup depth parameter within the burnup life, and obtain the burnup point division result, wherein the burnup point division result includes: the number of burnup points and the burnup depth corresponding to each burnup point; The modeling unit is used to establish a neutronics solution model for the spent fuel assembly; The solving unit is electrically connected to the modeling unit and the execution unit, respectively, and is used to calculate the neutron flux density value corresponding to each burnup point and the reaction cross-section value corresponding to each burnup point according to the division result of the burnup point and the neutronics solution model.

12. A fuel consumption calculation system, characterized in that: include: The second establishing module, the acquiring module, the second calculating module and the fuel consumption calculation data acquiring device according to claim 10 or 11, The second establishing module is used to establish a burnup equation of the target nuclide; The acquisition module is used to acquire the initial nuclide content of the target nuclide; The burnup calculation data acquisition device is used to obtain the average neutron flux density and average reaction cross section of each burnup step; The second calculation module is electrically connected to the second establishment module, the acquisition module and the burnup calculation data acquisition device, respectively, and is used to iteratively calculate the burnup equation according to the initial nuclide content, the average neutron flux density value of each burnup step and the average reaction cross section value to obtain the nuclide content of the target nuclide at the end of the burnup life.

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