Reactor core performance calculation device

Abstract

To provide a core performance calculator capable of accurately obtaining the power distribution in the core.SOLUTION: The core performance calculator includes: a nuclear constant memory that stores nuclear constants pre-estimated by fuel assembly analysis; and a 3D core nuclear thermal hydraulic characteristic analysis unit that obtains core characteristics including the output of fuel assemblies. The nuclear constant memory stores, as nuclear constants, the response relationship between the neutrons entering the fuel assembly cell and the nuclear properties of the fuel assembly, and the response relationship between neutrons emitted from fuel rods and the nuclear characteristics of fuel assemblies. The 3D core nuclear thermal hydraulic characteristic analysis unit obtains an effective neutron multiplication factor by using the response relationship memorized in the nuclear constant memory, and the core performance calculator for obtaining the fuel assembly output using the effective neutron multiplication factor is constituted.SELECTED DRAWING: Figure 2

Description

[Technical field]

[0001] The present invention relates to a core performance calculation device for monitoring the operating state of a nuclear reactor. [Background technology]

[0002] The purpose of core performance calculations is to calculate the three-dimensional power distribution within the reactor core using data measured in the plant, evaluate the power of each fuel assembly or each fuel rod, and confirm that the reactor is being operated under conditions where these are within preset thermal limit values ​​(see, for example, Patent Document 1, etc.).

[0003] The core performance calculations can be performed periodically, for example, at a specified interval of one to six hours, or whenever required by the user, to confirm the operating status of the reactor at that time.

[0004] A conventional method for calculating reactor core performance will be described below. First, reactor core status data such as the core flow rate, core thermal power, and control rod position, plant data such as neutron detector measurements, nuclear constants such as the infinite neutron multiplication factor and macroscopic cross section, which have been calculated in advance for each fuel assembly, and other necessary data are read in. The nuclear constants to be input here are tabulated with parameters such as burnup and moderator density. Next, the flow rate distribution to each fuel assembly is calculated, and the void distribution is calculated. Furthermore, the moderator density distribution is calculated from the void distribution, and the moderator density, burnup, etc. are used to calculate the nuclear constants for each fuel assembly and for each node divided vertically within the core, using the nuclear constants that have already been read. Next, using the calculated nuclear constants, the three-dimensional neutron flux distribution in the core is calculated using a diffusion equation, and the power distribution is calculated from the neutron flux distribution. In this way, the power distribution within the core can be calculated.

[0005] In the process of calculating the power distribution described above, the power distribution in the height direction (axial direction) is corrected to match the measurement value of the neutron detector. However, no correction is made in the horizontal (radial) direction. Therefore, it is important to accurately evaluate the power distribution before correction in the horizontal direction (radial direction). [Prior art documents] [Patent documents]

[0006] [Patent Document 1] Japanese Patent Application Publication No. 11-295473 Summary of the Invention [Problem to be solved by the invention]

[0007] Recently, reactor cores containing a mixture of fuel assemblies with significantly different enrichments, or a mixture of uranium and mixed oxide fuel assemblies (MOX fuel assemblies), have been put into practical use. In such cores, the nuclear characteristics of adjacent fuel assemblies are significantly different.

[0008] On the other hand, nuclear constants such as macroscopic cross sections used in conventional core performance calculations are evaluated under the assumption that adjacent fuel assemblies are all identical to themselves, and are spatially homogenized within the fuel assembly cell, which includes half the area of ​​the gap between the fuel assemblies. Furthermore, since the three-dimensional neutron flux distribution is calculated using a diffusion equation using nuclear constants, errors occur due to the diffusion approximation. Therefore, for a core in which the nuclear characteristics of adjacent fuel assemblies differ significantly, the assumptions for evaluating the nuclear constants differ significantly from reality, and the accuracy decreases due to homogenization of the nuclear constants and diffusion approximation.

[0009] In order to solve the above-mentioned problems, the present invention provides a core performance calculation device capable of determining the power distribution in a core with high accuracy.

[0010] The above and other objects of the present invention and novel features of the present invention will become apparent from the description of this specification and the accompanying drawings. [Means for solving the problem]

[0011] The core performance calculation device of the present invention includes a nuclear constant memory device that stores nuclear constants that have been evaluated in advance by fuel assembly analysis, and a three-dimensional core nuclear thermal-hydraulic characteristic analysis device that calculates core characteristics including the power of the fuel assembly. In the core performance calculation device of the present invention, the nuclear constant memory device stores, as nuclear constants, the response relationship between neutrons flowing into the fuel assembly cells and the nuclear characteristics of the fuel assembly, and the response relationship between neutrons generated from the fuel rods and the nuclear characteristics of the fuel assembly, and in the three-dimensional core nuclear thermal-hydraulic characteristic analysis device, the response relationship stored in the nuclear constant memory device is used to determine the effective neutron multiplication factor, and the fuel assembly power is determined using the effective neutron multiplication factor. The first core performance calculation device of the present invention further stores, in the nuclear constant memory device, nuclear constants such as the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out of the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, the response relationship between neutrons generated from a fuel rod and neutrons flowing out of a fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutron flux, the response relationship between neutrons flowing into a fuel assembly cell and a macroscopic cross-sectional area, The response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out of the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by those neutrons, the response relationship between neutrons generated from a fuel rod and neutron flux, and the response relationship between neutrons generated from a fuel rod and macroscopic cross section are stored in the three-dimensional core nuclear thermal-hydraulic characteristic analysis device, and the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out of the fuel assembly cell and neutrons generated from a fuel rod, which are stored in a nuclear constant storage device, is stored in the three-dimensional core nuclear thermal-hydraulic characteristic analysis device. a step of calculating the outflow neutron flow and updating the fuel rod neutron production rate using the response relationship of the above, the inflow neutron flow, the fuel rod neutron production rate, and the effective neutron multiplication factor, a step of setting the inflow neutron flow of each node to the outflow neutron flow of an adjacent node, and a step of repeating the steps from the step of calculating the outflow neutron flow and updating the fuel rod neutron production rate to the step of setting the inflow neutron flow of each node until the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron production rate converge. After the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron production rate converge, a step of calculating a neutron flux using the response relationship between neutrons flowing into a fuel assembly cell and a neutron flux, the response relationship between neutrons generated from a fuel rod and a neutron flux, the calculated flowing neutron flow, the updated fuel rod neutron production rate, and the effective neutron multiplication factor, all of which are stored in a nuclear constant memory device, in a three-dimensional core nuclear thermal hydraulic characteristic analysis device; a step of calculating a macroscopic cross-sectional area using the response relationship between neutrons flowing into a fuel assembly cell and a macroscopic cross-sectional area, all of which are stored in a nuclear constant memory device, the response relationship between neutrons generated from a fuel rod and a macroscopic cross-sectional area, and the calculated neutron flux;The system is configured to execute a step of calculating the effective neutron multiplication factor using the calculated macroscopic cross section, and a step of repeating the steps of calculating the outflow neutron flow and updating the fuel rod neutron production rate, through to the step of calculating the effective neutron multiplication factor, until the effective neutron multiplication factor converges. After the effective neutron multiplication factor converges, the power distribution in the core is calculated using the neutron flux and the fission cross section, and the fuel assembly power is determined. The second reactor core performance calculation apparatus of the present invention further stores, in a nuclear constant memory device, as nuclear constants, the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out of the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, and the response relationship between neutrons generated from a fuel rod and neutrons flowing out of a fuel assembly cell. In the calculation, the outflowing neutron flow is calculated using the response relationship between neutrons flowing into the fuel assembly cell and neutrons flowing out from the fuel assembly cell, the response relationship between neutrons flowing into the fuel assembly cell and neutrons generated from the fuel rod, the response relationship between neutrons generated from the fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, the response relationship between neutrons generated from the fuel rod and neutrons flowing out from the fuel assembly cell, the inflowing neutron flow, the fuel rod neutron production rate, and the effective neutron multiplication factor, all of which are stored in the nuclear constant memory device. and updating the fuel rod neutron production rate; setting the inflow neutron flow of each node to the outflow neutron flow of an adjacent node; and repeating the steps from the calculation of the outflow neutron flow and updating the fuel rod neutron production rate to the setting of the inflow neutron flow of each node until the inflow neutron flow, outflow neutron flow, and fuel rod neutron production rate converge. After the inflow neutron flow, outflow neutron flow, and fuel rod neutron production rate converge, The system is configured to execute the following steps: updating the effective neutron multiplication factor by multiplying the ratio of the outflow neutron flow to the inflow neutron flow in the entire core; and repeating the steps of calculating the outflow neutron flow and updating the fuel rod neutron generation rate, through to updating the effective neutron multiplication factor, until the effective neutron multiplication factor converges.After the effective neutron multiplication factor converges, the fuel rod neutron generation rate is calculated, and the calculated fuel rod neutron generation rate is integrated over the fuel assembly and converted into a power distribution, thereby determining the fuel assembly power. The third core performance calculation device of the present invention further stores, in the nuclear constant memory device, the following nuclear constants: a response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out from the fuel assembly cell; a response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod; a response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons; a response relationship between neutrons generated from a fuel rod and neutrons flowing out from a fuel assembly cell; a response relationship between neutrons flowing into a fuel assembly cell and neutron flux; and a response relationship between neutrons flowing into a fuel assembly cell and a macroscopic cross section. The relationship between the response of neutrons flowing into a fuel assembly cell and the response of neutrons flowing out of the fuel assembly cell, the response of neutrons flowing into a fuel assembly cell and the response of neutrons generated from a fuel rod, the response of neutrons generated from a fuel rod and the response of neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, and the response of neutrons generated from a fuel rod and the response of neutrons flowing out of a fuel assembly cell are stored in a nuclear constant memory device in a three-dimensional core nuclear thermal-hydraulic characteristic analysis device. a step of calculating a response matrix using the inflow neutron flow, the fuel rod neutron production rate, and the effective neutron multiplication factor; a step of calculating an outflow neutron flow by multiplying the response matrix by the inflow neutron flow, and updating the fuel rod neutron production rate using the response relationship between the neutrons flowing into the fuel assembly cell and the neutrons generated from the fuel rod, the response relationship between the neutrons generated from the fuel rod and the neutrons generated from other fuel rods due to the nuclear fission reaction induced by the neutrons, the inflow neutron flow, the fuel rod neutron production rate, and the effective neutron multiplication factor; and a step of setting the inflow neutron flow of each node to the outflow neutron flow of an adjacent node. and repeating the steps from the step of calculating the response matrix to the step of setting the inflow neutron flow of each node until the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron production rate converge. After the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron production rate converge, in a three-dimensional core nuclear thermal-hydraulic characteristic analyzer, using the response relationship between neutrons flowing into the fuel assembly cell and neutron flux, the response relationship between neutrons generated from the fuel rod and neutron flux, the calculated inflow neutron flow, the updated fuel rod neutron production rate, and the effective neutron multiplication factor,The system includes the steps of: calculating the neutron flux; calculating the macroscopic cross-sectional area using the response relationship between neutrons flowing into the fuel assembly cells and the macroscopic cross-sectional area, and the response relationship between neutrons generated from the fuel rods and the macroscopic cross-sectional area, which are stored in a nuclear constant storage device; calculating the effective neutron multiplication factor using the calculated neutron flux and the calculated macroscopic cross-sectional area; and repeating the steps from calculating the outflowing neutron flow and updating the fuel rod neutron production rate to calculating the effective neutron multiplication factor until the effective neutron multiplication factor converges. After the effective neutron multiplication factor converges, the neutron flux and the fission cross-sectional area are used to calculate the power distribution in the core, and the fuel assembly power is determined. Effect of the Invention

[0012] According to the above-mentioned core performance calculation device of the present invention, the power distribution in the core can be obtained with high accuracy. In particular, even in a core in which the nuclear characteristics of adjacent fuel assemblies differ significantly, the power distribution within the core can be determined with high accuracy.

[0013] Problems, configurations and effects other than those described above will become apparent from the following description of the embodiments. [Brief description of the drawings]

[0014] [Figure 1] FIG. 1 is a schematic configuration diagram of a core performance calculation device according to a first embodiment of the present invention. [Diagram 2] 1 is a schematic flowchart of a flow of power distribution calculation in a three-dimensional core nuclear thermal-hydraulic characteristic analysis device of a core performance calculation device according to a first embodiment of the present invention. [Diagram 3]11 is a schematic flowchart of a flow of power distribution calculation in a three-dimensional core nuclear thermal-hydraulic characteristic analysis device of a core performance calculation device according to a second embodiment of the present invention. [Figure 4] 1 is a flowchart showing an outline of a flow of power distribution calculation using a direct response matrix method. DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

[0015] Hereinafter, the embodiments and examples of the present invention will be described with reference to text and drawings. However, the structures, materials, and other specific configurations shown in the present invention are not limited to the embodiments and examples described here, and can be appropriately combined or improved without changing the gist of the invention. Elements that are not directly related to the present invention are not shown.

[0016] The core performance calculation device of the present invention includes a nuclear constant memory device that stores nuclear constants that have been evaluated in advance by fuel assembly analysis, and a three-dimensional core nuclear thermal-hydraulic characteristic analysis device that calculates core characteristics including the power of the fuel assembly. In the core performance calculation device of the present invention, the nuclear constant memory device stores, as nuclear constants, the response relationship between neutrons flowing into the fuel assembly cells and the nuclear characteristics of the fuel assembly, and the response relationship between neutrons generated from the fuel rods and the nuclear characteristics of the fuel assembly, and in the three-dimensional core nuclear thermal-hydraulic characteristic analysis device, the response relationship stored in the nuclear constant memory device is used to determine the effective neutron multiplication factor, and the fuel assembly power is determined using the effective neutron multiplication factor.

[0017] According to the core performance calculation device of the present invention, the response relationship between neutrons flowing into the fuel assembly cell and the nuclear characteristics of the fuel assembly, and the response relationship between neutrons generated from the fuel rods and the nuclear characteristics of the fuel assembly are stored as nuclear constants, and the effective neutron multiplication factor is calculated using these stored response relationships, and the effective neutron multiplication factor is used to calculate the fuel assembly power. As a result, the response relationship between neutrons flowing into the fuel assembly cell and the nuclear characteristics of the fuel assembly, and the response relationship between neutrons generated from the fuel rods and the nuclear characteristics of the fuel assembly are not homogenized within the fuel assembly cell, so that the assumptions for nuclear constant evaluation can be adapted and the method can be applied without problems even in a core in which the nuclear characteristics of adjacent fuel assemblies differ significantly. Furthermore, since the diffusion approximation is not used in the calculation of the power distribution, errors due to the diffusion approximation do not occur, and the power distribution in the core can be obtained with high accuracy. In particular, even in a core in which the nuclear characteristics of adjacent fuel assemblies differ significantly, the power distribution within the core can be determined with high accuracy.

[0018] In the configuration of the above-mentioned core performance calculation device, the three-dimensional core nuclear thermal-hydraulic characteristic analysis device can be further configured to execute the following steps: calculating the outflow neutron flow and updating the fuel rod neutron generation rate using the response relationship between neutrons flowing into the fuel assembly cell and the fuel assembly nuclear characteristics, the response relationship between neutrons generated from the fuel rod and the fuel assembly nuclear characteristics, the inflow neutron flow, the fuel rod neutron generation rate, and the effective neutron multiplication factor; setting the inflow neutron flow of each node to the outflow neutron flow of an adjacent node; and repeating the calculation of the outflow neutron flow and the update of the fuel rod neutron generation rate until the inflow neutron flow, outflow neutron flow, and fuel rod neutron generation rate converge. That is, in the case of this configuration, the neutron flow and the fuel rod neutron generation rate can be calculated without calculating the response matrix that needs to be calculated by the direct response matrix method described later. Therefore, compared with the case where the response matrix is ​​calculated by the direct response matrix method, it is possible to significantly reduce the amount of calculation and the calculation time.

[0019] In the configuration of the above-mentioned core performance calculation device, the response relationship with the fuel assembly nuclear characteristics can be further configured as a response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out of the fuel assembly cell, a response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, a response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, and a response relationship between neutrons generated from a fuel rod and neutrons flowing out of a fuel assembly cell. That is, in the case of this configuration, the neutron flow and the fuel rod neutron generation rate can be obtained by using the four types of response relationships without obtaining a response matrix that needs to be obtained by the direct response matrix method described later. Therefore, compared with the case where the response matrix is ​​obtained by employing the direct response matrix method, it is possible to significantly reduce the amount of calculation and the calculation time.

[0020] In addition, in the configuration of the above-mentioned core performance calculation device, a three-dimensional core nuclear thermal-hydraulic characteristic analysis device can be configured to determine the effective neutron multiplication factor using the response relationship between neutrons flowing into the fuel assembly cell and neutron flux, the response relationship between neutrons flowing into the fuel assembly cell and the macroscopic cross-sectional area, the response relationship between neutrons generated from the fuel rods and neutron flux, and the response relationship between neutrons generated from the fuel rods and the macroscopic cross-sectional area. That is, in the case of this configuration, the effective neutron multiplication factor can be directly evaluated using the response relationship with the neutron flux and the macroscopic cross section, which makes it possible to improve the convergence of the effective neutron multiplication factor and significantly reduce the calculation time compared to the method of adjusting it based on the increase or decrease in the neutron flow.

[0021] In the configuration of the above-mentioned core performance calculation device, the response relationships with the fuel assembly nuclear characteristics can be configured as follows: a response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out from the fuel assembly cell; a response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod; a response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons; a response relationship between neutrons generated from a fuel rod and neutrons flowing out from a fuel assembly cell; a response relationship between neutrons flowing into a fuel assembly cell and neutron flux; a response relationship between neutrons flowing into a fuel assembly cell and a macroscopic cross-sectional area; a response relationship between neutrons generated from a fuel rod and neutron flux; and a response relationship between neutrons generated from a fuel rod and a macroscopic cross-sectional area. That is, in the case of this configuration, the neutron flow and the neutron generation rate of the fuel rod can be obtained by using the response relationship between neutrons and neutrons, without obtaining the response matrix that needs to be obtained by the direct response matrix method described later. Therefore, compared to the case where the response matrix is ​​obtained by adopting the direct response matrix method, it is possible to significantly reduce the amount of calculation and significantly reduce the calculation time. In addition, since the effective neutron multiplication factor can be directly evaluated using the response relationship with the neutron flux and the macroscopic cross section, it is possible to improve the convergence of the effective neutron multiplication factor and significantly reduce the calculation time compared to the method of adjusting it from the increase and decrease of the neutron flow.

[0022] As a method for accurately evaluating the power distribution in a reactor core, for example, there is a method called the direct response matrix method. In this method, the power distribution in the core is calculated using four types of response relations, which are previously obtained in an analysis of a fuel assembly cell, instead of the macroscopic cross section as the nuclear constant. The four types of response relationships mentioned above are: the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out of the fuel assembly cell (T); the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod (S); the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons (A); and the response relationship between neutrons generated from a fuel rod and neutrons flowing out of a fuel assembly cell (L). These four types of response relationships T, S, A, and L are expressed as matrices and are not homogenized within the fuel assembly cell.

[0023] This direct response matrix method can also be applied to the core performance calculation device of the present invention to perform power distribution calculations. When the direct response matrix method is applied to the core performance calculation device of the present invention, the four types of response relationships T, S, A, and L of the direct response matrix method described above are adopted for the response relationship between neutrons flowing into the fuel assembly cells and the nuclear characteristics of the fuel assembly, and the response relationship between neutrons generated from the fuel rods and the nuclear characteristics of the fuel assembly, which are stored in the nuclear constant storage device.

[0024] FIG. 4 shows a schematic flow chart of the power distribution calculation flow using the direct response matrix method. First, in step S21, a response matrix R of each node is calculated from the four types of response relationships T, S, A, and L according to the following formula (11). R=T+SL / k+SAL / k 2 +SA 2 L / k 3 +SA 3 L / k 4 +···(11) Here, k is the effective neutron multiplication factor.

[0025] Next, in step S22, for each node, the response matrix R is multiplied by the inflow neutron flow Jin to obtain the outflow neutron flow Jout. That is, the outflow neutron flow Jout is obtained as shown in the following equation (12). Jout=R*Jin (12)

[0026] Furthermore, in step S23, the inflow neutron flow of each node is set to the outflow neutron flow of the adjacent node, and the inflow neutron flow Jin is calculated. That is, set Jin = Jout of the adjacent node.

[0027] This neutron flow calculation is repeated until convergence. That is, in step S24, it is determined whether the inflow neutron flow Jin and the outflow neutron flow Jout are converging, and if they are converging, the process proceeds to step S25, and if they are not converging, the process returns to step S22.

[0028] After the neutron flow converges, in step S25, the effective neutron multiplication factor k is updated by multiplying the ratio of the outflowing neutron flow to the inflowing neutron flow in the entire core. That is, the effective neutron multiplication factor k is updated as shown in the following equation (13). k i+1 =k i *Jout in all cores / Jin in all cores (13) where k i is the effective neutron multiplication factor at the i-th iteration, k i+1 is the effective neutron multiplication factor in the (i+1)th iteration.

[0029] Next, the response matrix R is calculated using the updated effective neutron multiplication factor k, and the calculation of the neutron flow is repeated until the effective neutron multiplication factor k converges. That is, in step S26, it is determined whether the effective neutron multiplication factor k has converged. If the effective neutron multiplication factor k has not converged, the process returns to step S21 and calculates the response matrix R. On the other hand, if the effective neutron multiplication factor k has converged, the process proceeds to step S27.

[0030] After the effective neutron multiplication factor k has converged, the power distribution in the core is calculated using the inflow neutron flow and response relationship obtained. That is, in step S27, the neutron generation rate of the fuel rod is calculated. Then, the calculated neutron generation rate of the fuel rod is integrated over the fuel assembly and converted into a power distribution.

[0031] According to the power distribution calculation by the direct response matrix method described above, the nuclear constants are not homogenized and the diffusion approximation is not used in the power distribution calculation, so that the power distribution can be calculated with high accuracy. Even in a core where the nuclear characteristics of adjacent fuel assemblies are significantly different, the power distribution in the core can be calculated with high accuracy.

[0032] However, the power distribution calculation using the direct response matrix method shown in the flowchart of Figure 4 has the following two problems from the viewpoint of applying the direct response matrix method to core performance calculations.

[0033] The first problem is the enormous amount of calculation required to calculate the response matrix R. The addition in the formula for calculating the response matrix R shown in step S21 of Fig. 4 continues forever, but since the contribution of higher-order terms is smaller, it is possible to terminate it at a realistic number of terms. However, since many multiplications and additions of the matrix are required, it requires a large amount of calculation and a large amount of calculation time.

[0034] The second problem is the poor convergence of the iterative calculation to obtain the effective neutron multiplication factor k. The effective neutron multiplication factor k is adjusted by the rate at which deviations from its converged value increase or decrease the neutron flow through the response matrix R. However, the effective neutron multiplication factor k adjusts only a part of the response matrix R, and can only adjust a part of the deviation of the effective neutron multiplication factor k from the convergence value. Furthermore, as the effective neutron multiplication factor k approaches the convergence value, the adjustment amount becomes smaller and the convergence becomes slower.

[0035] Therefore, in the core performance calculation apparatus of the present invention, it is possible to further adopt a configuration for solving the first problem described above, or a configuration for solving the second problem described above.

[0036] As a configuration for solving the first problem described above, there is a configuration in which the neutron production rate of the fuel rod is treated as an unknown quantity, and the inflow and outflow neutron flows and the neutron production rate of the fuel rod are obtained using four types of response relationships. This eliminates the need to calculate the response matrix R, significantly reducing the amount of calculation and the calculation time.

[0037] One configuration for solving the second problem mentioned above is to obtain a new response relationship in advance in an analysis of a fuel assembly cell, and to directly evaluate the effective neutron multiplication factor using the response relationship. The new response relationships are the response relationship between neutrons flowing into the fuel assembly cell and neutron flux (Fs), the response relationship between neutrons flowing into the fuel assembly cell and macroscopic cross section (Cs), the response relationship between neutrons generated from the fuel rod and neutron flux (Ff), and the response relationship between neutrons generated from the fuel rod and macroscopic cross section (Cf). Note that since there are multiple types of macroscopic cross section, such as neutron absorption cross section and neutron production cross section, the response relationship with the macroscopic cross section is calculated for each type of macroscopic cross section. After the inflow neutron flow and fuel rod neutron production rate are obtained, the neutron flux and macroscopic cross section of each node are calculated using these response relationships Fs, Cs, Ff, and Cf. Using these neutron fluxes and macroscopic cross sections, the effective neutron multiplication factor is directly calculated as the ratio of the neutron production rate and neutron absorption rate in the entire core. This improves convergence and significantly reduces calculation time compared to the direct response matrix method, which adjusts from the increase or decrease in neutron flow.

[0038] Next, specific examples of the present invention will be described below. In addition, the embodiment described below has the same effect of being able to calculate the power distribution with high accuracy as when the direct response matrix method is applied to the core performance calculation device of the present invention, and is configured to be able to solve one or both of the two problems that arise when the direct response matrix method is applied to the core performance calculation device of the present invention.

[0039] Example 1 A core performance calculation device according to a first embodiment of the present invention will be described with reference to FIGS. FIG. 1 shows a schematic configuration diagram of a core performance calculation device according to a first embodiment of the present invention.

[0040] The core performance calculation device 1 of this embodiment comprises a plant data input device 4, a nuclear constant storage device 5, a three-dimensional core nuclear thermal-hydraulic characteristic analysis device 6, a requirement input device 7, and a display device 8. The plant data input device 4 receives input of core state data from the nuclear reactor 2 and the core 3 and measurement values ​​from the neutron detector. The nuclear constant storage device 5 stores nuclear constants calculated in advance for each fuel assembly. The three-dimensional core nuclear thermal-hydraulic characteristic analysis device 6 reads data from the plant data input device 4 and the nuclear constant storage device 5, and performs analysis of power distribution and the like. The request input device 7 receives requests from operators of the reactor. The display device 8 displays the analysis results from the three-dimensional core nuclear and thermal-hydraulic characteristic analysis device 6 .

[0041] The device configuration shown in FIG. 1 is the same as that of a conventional core performance calculation device. The difference from the conventional core performance calculation device is the contents of the nuclear constants stored in the nuclear constant storage device 5 and the calculation method in the three-dimensional core nuclear thermal-hydraulic characteristic analysis device 6.

[0042] In the nuclear constant memory device 5 of this embodiment, eight types of response relationships are stored in place of the macroscopic cross-sectional areas homogenized by fuel assembly cells, which were stored in the nuclear constant memory device of the conventional core performance calculation device. The eight types of response relationships are: response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out of the fuel assembly cell (T); response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod (S); response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons (A); response relationship between neutrons generated from a fuel rod and neutrons flowing out of a fuel assembly cell (L); response relationship between neutrons flowing into a fuel assembly cell and neutron flux (Fs); response relationship between neutrons flowing into a fuel assembly cell and the macroscopic cross-sectional area (Cs); response relationship between neutrons generated from a fuel rod and neutron flux (Ff); and response relationship between neutrons generated from a fuel rod and the macroscopic cross-sectional area (Cf).

[0043] FIG. 2 shows a schematic flow chart of the flow of power distribution calculation in the three-dimensional core nuclear thermal-hydraulic characteristic analysis device 6 of the core performance calculation device according to the first embodiment of the present invention.

[0044] First, in step S1, the outflowing neutron flow Jout is calculated using the four types of response relationships T, S, A, L, the inflowing neutron flow Jin, the neutron production rate P of the fuel rod, and the effective neutron multiplication factor k, and the neutron production rate P of the fuel rod is updated. That is, the outflow neutron flow Jout is calculated by the following equation (1), and the neutron production rate P of the fuel rod is updated by the equation (2). Jout = T * Jin + L * P / k (1) P = S * Jin + A * P / k (2)

[0045] Next, in step S2, the incoming neutron flow of each node is set to the outgoing neutron flow of the adjacent node, and the incoming neutron flow Jin is calculated. That is, set Jin = Jout of the adjacent node.

[0046] This calculation of the neutron flow and the neutron production rate distribution of the fuel rods is repeated until convergence. That is, in step S3, it is determined whether the inflow neutron flow Jin, outflow neutron flow Jout, and neutron production rate P of the fuel rod are converged. If they are converged, the process proceeds to step S4, and if they are not converged, the process returns to step S1.

[0047] In step S4, the neutron flux φs resulting from the inflow neutrons and the neutron flux φf resulting from neutrons generated from the fuel rods are calculated using the calculated inflow neutron flow Jin, the neutron production rate P of the updated fuel rods, and two types of response relationships Fs and Ff. That is, the neutron flux φs resulting from the inflow neutrons is calculated using the following equation (3), and the neutron flux φf resulting from the neutrons generated from the fuel rods is calculated using the following equation (4). φs=Fs*Jin (3) φf=Ff*P / k (4)

[0048] Next, in step S5, the two types of response relations Cs and Cf are averaged using the neutron fluxes φs and φf calculated in step S4 as weights to obtain the macroscopic cross section Σ. That is, the macroscopic cross section Σ is obtained according to the following equation (5). Σ=(Cs*φs+Cf*φf) / (φs+φf) (5)

[0049] Next, in step S6, the neutron fluxes φs and φf calculated in step S4 and the macroscopic cross section Σ calculated in step S5 are used to directly calculate the effective neutron multiplication factor k as the ratio of the neutron production rate to the neutron absorption rate in the entire core. That is, the effective neutron multiplication factor k is calculated as shown in the following equation (6). k=νΣf(φs+φf) / Σa(φs+φf) (6) where νΣf is the neutron production cross section and Σa is the neutron absorption cross section.

[0050] These steps are repeated until the effective neutron multiplication factor k converges. That is, in step S7, it is determined whether the effective neutron multiplication factor k has converged. If the effective neutron multiplication factor k has not converged, the process returns to step S1. On the other hand, if the effective neutron multiplication factor k has converged, the process proceeds to step S8.

[0051] After the effective neutron multiplication factor k has converged, in step S8, the power distribution in the core is calculated using the determined neutron flux and fission cross section.

[0052] According to this embodiment, the effective neutron multiplication factor k and the power distribution of the fuel assembly are obtained using the response relationships T, S, A, L, Fs, Cs, Ff, and Cf without using the homogenization cross section as the nuclear constant. As a result, the response relationships T, S, A, L, Fs, Cs, Ff, and Cf are not homogenized within the fuel assembly cell, so that the assumptions for nuclear constant evaluation can be adapted and the method can be applied without problems even in cores where the nuclear characteristics of adjacent fuel assemblies are significantly different. Furthermore, since the diffusion approximation is not used in the calculation of the power distribution, errors due to the diffusion approximation do not occur, and the power distribution within the core can be calculated with higher accuracy than with conventional core performance calculation devices. In particular, even in a core in which the nuclear characteristics of adjacent fuel assemblies differ significantly, the power distribution within the core can be determined with high accuracy.

[0053] In addition, according to this embodiment, the response matrix obtained directly by the response matrix method is not obtained, and the fuel rod neutron production rate P is treated as an unknown quantity, and the inflow and outflow neutron flows Jin, Jout and the fuel rod neutron production rate P are obtained using four types of response relationships T, S, A, and L. This eliminates the need to calculate the response matrix, significantly reducing the amount of calculation and the calculation time. In other words, the first problem in the case of applying the direct response matrix method described above can be solved.

[0054] Moreover, according to this embodiment, the effective neutron multiplication factor k is directly evaluated using the response relationships Fs, Cs, Ff, and Cf with the neutron flux and the macroscopic cross section. This improves convergence and significantly reduces calculation time compared to the direct response matrix method, which adjusts the neutron flow by increasing or decreasing it. In other words, it solves the second problem that occurs when the direct response matrix method is applied.

[0055] Example 2 A core performance calculation device according to a second embodiment of the present invention will be described below. The configuration of the core performance calculation apparatus of this embodiment is the same as that of the core performance calculation apparatus 1 of the first embodiment.

[0056] In this embodiment, the nuclear constants stored in the nuclear constant memory device 5 are the same four types of response relationships as in the direct response matrix method: the response relationship (T) between neutrons flowing into a fuel assembly cell and neutrons flowing out of the fuel assembly cell, the response relationship (S) between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship (A) between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by those neutrons, and the response relationship (L) between neutrons generated from a fuel rod and neutrons flowing out of a fuel assembly cell.

[0057] This embodiment differs from the first embodiment in that the nuclear constant memory device 5 does not store the response relationships Fs, Cs, Ff, and Cf with the neutron flux or the macroscopic cross section. In this embodiment, since the response relationships Fs, Cs, Ff, and Cf with the neutron flux or the macroscopic cross section are not stored, it is not possible to improve the convergence of the iterative calculation for obtaining the effective neutron multiplication factor k. However, in this embodiment, similarly to the first embodiment, the neutron flow and the neutron production rate of the fuel rods can be obtained without obtaining a response matrix.

[0058] FIG. 3 shows a schematic flow chart of the power distribution calculation flow in the three-dimensional core nuclear thermal-hydraulic characteristic analysis device 6 of the core performance calculation device according to the second embodiment of the present invention.

[0059] First, in step S11, the outflowing neutron flow Jout is calculated using four types of response relationships T, S, A, L, the inflowing neutron flow Jin, the neutron production rate P of the fuel rod, and the effective neutron multiplication factor k, and the neutron production rate P of the fuel rod is updated. That is, similarly to step S1 in FIG. 2, the outflow neutron flow Jout is calculated by the following formula (1), and the neutron production rate P of the fuel rod is updated by the formula (2). Jout = T * Jin + L * P / k (1) P = S * Jin + A * P / k (2)

[0060] Next, in step S12, the incoming neutron flow of each node is set to the outgoing neutron flow of the adjacent node, and the incoming neutron flow Jin is calculated. That is, set Jin = Jout of the adjacent node.

[0061] Furthermore, the calculation of the neutron flow and the neutron production rate distribution of the fuel rods is repeated until convergence is reached. That is, in step S13, it is determined whether the inflow neutron flow Jin, outflow neutron flow Jout, and neutron production rate P of the fuel rod are converged. If they are converged, the process proceeds to step S14, and if they are not converged, the process returns to step S11.

[0062] In step S14, the effective neutron multiplication factor k is updated by multiplying the ratio of the outflowing neutron flow to the inflowing neutron flow in the entire core. That is, similar to step S25 in FIG. 4, the effective neutron multiplication factor k is updated according to the following formula (13). k i+1 =k i *Jout in all cores / Jin in all cores (13) where k i is the effective neutron multiplication factor at the i-th iteration, k i+1 is the effective neutron multiplication factor in the (i+1)th iteration.

[0063] Next, the updated effective neutron multiplication factor k is used to calculate the outflow neutron flow Jout and update the neutron production rate P of the fuel rod. The calculation of the outflow neutron flow Jout and the update of the neutron production rate P of the fuel rod are repeated until the effective neutron multiplication factor k converges. That is, in step S15, it is determined whether the effective neutron multiplication factor k has converged. If the effective neutron multiplication factor k has not converged, the process returns to step S11, the outflow neutron flow Jout is calculated, and the neutron production rate P of the fuel rod is updated. On the other hand, if the effective neutron multiplication factor k has converged, the process proceeds to step S16.

[0064] After the effective neutron multiplication factor k has converged, the power distribution in the core is calculated using the inflow neutron flow and response relationship obtained. That is, in step S16, the neutron generation rate of the fuel rods is calculated in the same manner as in step S27 of Fig. 4. Then, the calculated neutron generation rate of the fuel rods is integrated over the fuel assembly and converted into a power distribution.

[0065] According to this embodiment, the effective neutron multiplication factor k and the power distribution of the fuel assembly are obtained using the response relationships T, S, A, and L without using the homogenization cross section as the nuclear constant. As a result, the response relationships T, S, A, and L are not homogenized within the fuel assembly cell, so that the assumptions for nuclear constant evaluation can be adapted and the method can be applied without problems even in cores where the nuclear characteristics of adjacent fuel assemblies are significantly different. Furthermore, since the diffusion approximation is not used in the calculation of the power distribution, errors due to the diffusion approximation do not occur, and the power distribution within the core can be calculated with higher accuracy than with conventional core performance calculation devices. In particular, even in a core in which the nuclear characteristics of adjacent fuel assemblies differ significantly, the power distribution within the core can be determined with high accuracy.

[0066] Furthermore, according to this embodiment, as in the case of embodiment 1, the response matrix is ​​not calculated directly using the response matrix method, and the fuel rod neutron production rate P is treated as an unknown quantity, and the inflow and outflow neutron flows Jin, Jout and the fuel rod neutron production rate P are calculated using four types of response relationships T, S, A, and L. This eliminates the need to calculate the response matrix, significantly reducing the amount of calculation and the calculation time. In other words, the first problem in the case of applying the direct response matrix method described above can be solved.

[0067] (Modification) The above-mentioned second embodiment is a configuration that solves only the first of the two problems when the direct response matrix method is applied, but it is also possible to configure it to solve only the second problem when the direct response matrix method is applied. For example, the outflow of S22 in FIG. Neutron flow With calculations fuel rod After the neutron production rate P is calculated and updated, and Jin, Jout, and P converge, the second problem can be solved by carrying out steps similar to S4 to S8 in FIG.

[0068] The present invention is not limited to the above-described embodiments and examples, and includes various modified examples. For example, the above-described embodiments and examples have been described in detail to easily explain the present invention, and the present invention is not necessarily limited to those having all of the configurations described. [Explanation of symbols]

[0069] 1...core performance calculation device, 2...nuclear reactor, 3...core, 4...plant data input device, 5...nuclear constant storage device, 6...three-dimensional core nuclear thermal-hydraulic characteristic analysis device, 7...request input device, 8...display device

Claims

1. A core performance calculation device having a nuclear constant storage device that stores nuclear constants previously evaluated by fuel assembly analysis, and a three-dimensional core nuclear thermal hydraulic characteristic analysis device that calculates core characteristics including the power of the fuel assembly, The nuclear constant memory device stores, as the nuclear constants, the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out from the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, the response relationship between neutrons generated from a fuel rod and neutrons flowing out from a fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutron flux, the response relationship between neutrons flowing into a fuel assembly cell and a macroscopic cross-sectional area, the response relationship between neutrons generated from a fuel rod and neutron flux, and the response relationship between neutrons generated from a fuel rod and a macroscopic cross-sectional area, In the three-dimensional core nuclear thermal-hydraulic characteristic analysis device, the following steps are executed: calculating the outflow neutron flow and updating the fuel rod neutron generation rate using the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out from the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods by nuclear fission reactions induced by the neutrons, and the response relationship between neutrons generated from a fuel rod and neutrons flowing out from a fuel assembly cell, the inflow neutron flow, the fuel rod neutron generation rate, and the effective neutron multiplication factor, all of which are stored in the nuclear constant storage device; setting the inflow neutron flow of each node to the outflow neutron flow of an adjacent node; and repeating the steps from the calculation of the outflow neutron flow and updating the fuel rod neutron generation rate to the setting of the inflow neutron flow of each node until the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron generation rate converge; After the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron production rate have converged, the three-dimensional core nuclear thermal-hydraulic characteristic analysis device calculates a neutron flux using the response relationship between neutrons flowing into a fuel assembly cell and neutron flux, the response relationship between neutrons generated from a fuel rod and neutron flux, which are stored in the nuclear constant storage device, the calculated inflow neutron flow, the updated fuel rod neutron production rate, and the effective neutron multiplication factor; A step of calculating a macroscopic cross-sectional area using the response relationship between the inflowing neutrons and the macroscopic cross-sectional area, the response relationship between the neutrons generated from the fuel rod and the macroscopic cross-sectional area, and the calculated neutron flux; a step of calculating an effective neutron multiplication factor using the calculated neutron flux and the calculated macroscopic cross-sectional area; and a step of repeating the steps from the calculation of the outflowing neutron flow and the update of the fuel rod neutron production rate to the step of calculating the effective neutron multiplication factor until the effective neutron multiplication factor converges are performed; After the effective neutron multiplication factor has converged, the power distribution in the core is calculated using the neutron flux and the fission cross section, and the fuel assembly power is obtained. A reactor core performance calculation device comprising:

2. A core performance calculation device having a nuclear constant storage device that stores nuclear constants previously evaluated by fuel assembly analysis, and a three-dimensional core nuclear thermal hydraulic characteristic analysis device that calculates core characteristics including the power of the fuel assembly, The nuclear constant memory device stores, as the nuclear constants, the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out from the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from another fuel rod due to a nuclear fission reaction induced by the neutrons, and the response relationship between neutrons generated from a fuel rod and neutrons flowing out from a fuel assembly cell, In the three-dimensional core nuclear thermal-hydraulic characteristic analysis device, the following steps are executed: calculating the outflow neutron flow and updating the fuel rod neutron generation rate using the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out from the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods by nuclear fission reactions induced by the neutrons, and the response relationship between neutrons generated from a fuel rod and neutrons flowing out from a fuel assembly cell, the inflow neutron flow, the fuel rod neutron generation rate, and the effective neutron multiplication factor, all of which are stored in the nuclear constant storage device; setting the inflow neutron flow of each node to the outflow neutron flow of an adjacent node; and repeating the steps from the calculation of the outflow neutron flow and updating the fuel rod neutron generation rate to the setting of the inflow neutron flow of each node until the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron generation rate converge; After the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron production rate have converged, the three-dimensional core nuclear thermal-hydraulic characteristic analysis device performs a step of updating the effective neutron multiplication factor by multiplying the ratio of the outflow neutron flow and the inflow neutron flow in the entire core, and a step of repeating the steps from the step of calculating the outflow neutron flow and updating the fuel rod neutron production rate to the step of updating the effective neutron multiplication factor until the effective neutron multiplication factor has converged, After the effective neutron multiplication factor converges, the neutron generation rate of the fuel rod is calculated, and the calculated neutron generation rate of the fuel rod is integrated for the fuel assembly and converted into a power distribution, thereby obtaining the fuel assembly power. A reactor core performance calculation device comprising:

3. A core performance calculation device having a nuclear constant storage device that stores nuclear constants previously evaluated by fuel assembly analysis, and a three-dimensional core nuclear thermal hydraulic characteristic analysis device that calculates core characteristics including the power of the fuel assembly, The nuclear constant memory device stores, as the nuclear constants, the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out from the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, the response relationship between neutrons generated from a fuel rod and neutrons flowing out from a fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutron flux, the response relationship between neutrons flowing into a fuel assembly cell and a macroscopic cross-sectional area, the response relationship between neutrons generated from a fuel rod and neutron flux, and the response relationship between neutrons generated from a fuel rod and a macroscopic cross-sectional area, a step of calculating a response matrix in the three-dimensional core nuclear thermal-hydraulic characteristic analysis device using the response relationship between neutrons flowing into a fuel assembly cell and neutrons flowing out from the fuel assembly cell, the response relationship between neutrons flowing into a fuel assembly cell and neutrons generated from a fuel rod, the response relationship between neutrons generated from a fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, the response relationship between neutrons generated from a fuel rod and neutrons flowing out from a fuel assembly cell, the flow of inflowing neutrons, the neutron generation rate of a fuel rod, and the effective neutron multiplication factor, all of which are stored in the nuclear constant storage device; The steps of multiplying the response matrix by the inflow neutron flow to calculate the outflow neutron flow, and updating the fuel rod neutron generation rate using the response relationship between neutrons flowing into the fuel assembly cell and neutrons generated from the fuel rod, the response relationship between neutrons generated from the fuel rod and neutrons generated from other fuel rods due to nuclear fission reactions induced by the neutrons, the inflow neutron flow, the fuel rod neutron generation rate, and the effective neutron multiplication factor, setting the inflow neutron flow of each node to the outflow neutron flow of an adjacent node, and repeating the steps from the step of calculating the response matrix to the step of setting the inflow neutron flow of each node until the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron generation rate converge, are performed, After the inflow neutron flow, the outflow neutron flow, and the fuel rod neutron production rate have converged, the three-dimensional core nuclear thermal-hydraulic characteristic analysis device calculates a neutron flux using the response relationship between neutrons flowing into a fuel assembly cell and neutron flux, the response relationship between neutrons generated from a fuel rod and neutron flux, which are stored in the nuclear constant storage device, the calculated inflow neutron flow, the updated fuel rod neutron production rate, and the effective neutron multiplication factor; A step of calculating a macroscopic cross-sectional area using the response relationship between the inflowing neutrons and the macroscopic cross-sectional area, the response relationship between the neutrons generated from the fuel rod and the macroscopic cross-sectional area, and the calculated neutron flux; a step of calculating an effective neutron multiplication factor using the calculated neutron flux and the calculated macroscopic cross-sectional area; and a step of repeating the steps from the calculation of the outflowing neutron flow and the update of the fuel rod neutron production rate to the step of calculating the effective neutron multiplication factor until the effective neutron multiplication factor converges are performed; After the effective neutron multiplication factor has converged, the power distribution in the core is calculated using the neutron flux and the fission cross section, and the fuel assembly power is obtained. A reactor core performance calculation device comprising:

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