Calculation method, calculation device, and program

Abstract

To provide a method for calculating a nuclear constant in consideration of an influence of a heterogeneous medium in a core axial direction.SOLUTION: The calculation method is a calculation method executed by a computer, and includes calculating a first neutron flux distribution in an axial direction of a reactor core by three-dimensional non-homogeneous single assembly calculation, and calculating a second neutron flux distribution in the axial direction of the reactor core by three-dimensional homogeneous single assembly calculation. A step of calculating a second neutron flux distribution in the axial direction of the core, a step of calculating an axial distribution of a correction factor such that a reaction rate based on the first neutron flux distribution can be reproduced by a three dimensional homogeneous single cluster calculation to which a correction factor of a nuclear constant based on a ratio between the first neutron flux and the second neutron flux is applied; The method includes a step of updating the second neutron flux distribution, a step of calculating the axial distribution of the correction factor until the correction factor converges, and a step of repeating the step of updating the second neutron flux distribution.SELECTED DRAWING: Figure 1

Description

[Technical Field]

[0001] The present disclosure relates to a calculation method, a calculation device, and a program. [Background technology]

[0002] Fast reactor core analysis employs a two-stage calculation method: an assembly nuclear constant calculation and a core calculation. In the assembly nuclear constant calculation, a non-homogeneous neutron transport calculation is performed for a two-dimensional single-assembly system, assuming an infinite axial sequence of fuel (axial infinity), and focusing on a single fuel assembly in the radial direction, taking into account the geometrical shapes of the fuel pellets, cladding, coolant, etc. Then, cross-section data (data representing the ease and probability of neutron-material reactions) are averaged using the obtained neutron flux distribution within the assembly as a weight to create the assembly homogeneous nuclear constant. In the second core calculation, the assembly homogeneous nuclear constants created in the first stage are used as input to perform homogeneous neutron diffusion or transport calculations for the entire three-dimensional core system homogenized within the assembly, and core nuclear properties such as the effective multiplication factor, reactivity coefficient, power distribution, and sodium void reactivity are calculated.

[0003] In this two-stage calculation, the effect of neutron flow due to heterogeneous media in the axial direction of the core is not taken into account in the calculation of the nuclear constants for axial infinity in the first stage. In particular, in fast reactor systems, the mean free path of neutrons (the average distance traveled from the time a neutron is generated until it is annihilated by a nuclear reaction) is several tens of centimeters long, and the influence of heterogeneous media is significant. Therefore, the treatment of axial infinity in the first stage becomes a source of error in the axial reaction rate distribution and the associated overall core nuclear characteristics.

[0004] Patent Document 1 discloses a method for performing a coarse-mesh core calculation in which assemblies are homogenized using an assembly nuclear constant table, and a heterogeneous core calculation assuming the same core state as the coarse-mesh core calculation, and then comparing the results of these two core calculations to determine correction factors for the assembly homogeneous cross section and discontinuity factors at the assembly boundary used in the coarse-mesh core calculation. Non-Patent Document 1 also discloses a reaction-rate ratio preservation method as a method for preserving the reaction rate obtained from a three-dimensional heterogeneous transport calculation in a homogeneous calculation, Non-Patent Document 2 discloses the Simultaneous-SPH method, which is an extension of the SPH method so that it can handle homogeneous transport calculations. [Prior art documents] [Patent documents]

[0005] [Patent Document 1] Japanese Patent Application Laid-Open No. 2003-66179 [Non-patent literature]

[0006] [Non-Patent Document 1] S.Kosaka, et al., New Control Rod Homogenization Method for Fast Reactors, J.Nucl.Sci.Technol., 31, 7, pp.647-653, 1994. [Non-patent document 2] G.Chiba, et al., A note on application of superhomogenization factors to integro-differential neutron transport equations, J.Nucl.Sci.Technol., 49, 2, pp.272-280, 2012. Summary of the Invention [Problem to be solved by the invention]

[0007] A method for calculating nuclear constants that takes into account the influence of heterogeneous media in the axial direction of the core is needed.

[0008] The present disclosure provides a calculation method, a calculation device, and a program that can solve the above problems. [Means for solving the problem]

[0009] The calculation method disclosed herein is a calculation method executed by a computer, and includes the steps of: calculating a first neutron flux distribution in the axial direction of a reactor core by a three-dimensional heterogeneous single-assembly calculation; calculating a second neutron flux distribution in the axial direction of the reactor core by a three-dimensional homogeneous single-assembly calculation; calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation applying a nuclear constant correction factor based on a ratio between the first neutron flux and the second neutron flux so that the reaction rate based on the first neutron flux distribution can be reproduced; updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and repeating the steps of calculating the axial distribution of the correction factor and updating the second neutron flux distribution until the correction factor converges.

[0010] The computing device disclosed herein includes: a means for calculating a first neutron flux distribution in the axial direction of a core by a three-dimensional heterogeneous single-assembly calculation; a means for calculating a second neutron flux distribution in the axial direction of the core by a three-dimensional homogeneous single-assembly calculation; a means for calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation to which a nuclear constant correction factor based on a ratio between the first neutron flux and the second neutron flux is applied, so that the reaction rate based on the first neutron flux distribution can be reproduced; a means for updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and a means for repeating the step of calculating the axial distribution of the correction factor and the step of updating the second neutron flux distribution until the correction factor converges.

[0011] The program of the present disclosure also causes a computer to execute the steps of: calculating a first neutron flux distribution in the axial direction of a core by a three-dimensional heterogeneous single-assembly calculation; calculating a second neutron flux distribution in the axial direction of the core by a three-dimensional homogeneous single-assembly calculation; calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation to which a correction factor of a nuclear constant based on a ratio between the first neutron flux and the second neutron flux is applied so as to reproduce the reaction rate based on the first neutron flux distribution; updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and repeating the steps of calculating the axial distribution of the correction factor and updating the second neutron flux distribution until the correction factor converges. [Effects of the Invention]

[0012] According to the calculation method, calculation device, and program disclosed herein, by calculating a correction factor for the nuclear constant at each position in the axial direction of the core, it is possible to calculate a nuclear constant that takes into account the effect of neutron flow due to heterogeneous media in the axial direction of the core. [Brief explanation of the drawings]

[0013] [Figure 1] 1 is a block diagram illustrating an example of a reactor core analysis device according to an embodiment. [Figure 2] FIG. 2 is a diagram showing an example of a radial core cross section according to an embodiment; [Figure 3] FIG. 2 is a diagram showing an example of an axial core cross section according to an embodiment; [Figure 4] FIG. 10 is a diagram illustrating an example of a correction factor according to the embodiment. [Figure 5] 3 is a flowchart illustrating an example of a core analysis process according to the embodiment. [Figure 6] 1 is a diagram illustrating an example of a hardware configuration of a reactor core analyzer according to an embodiment. DETAILED DESCRIPTION OF THE INVENTION

[0014] <Embodiment> The reactor core analysis device of the present disclosure will be described below with reference to FIGS. (composition) FIG. 1 is a block diagram showing an example of a reactor core analysis device according to an embodiment. When performing core calculations for a fast reactor, the core analysis device 10 calculates correction factors for nuclear constants for each axial position of the core, and inputs the nuclear constants obtained by multiplying the nuclear constants by the correction factors for each axial position of the core, thereby performing accurate core analysis. The core analysis device 10 includes an input receiving unit 11, a nuclear constant calculation unit 12, a core calculation unit 13, and a memory unit 14.

[0015] The input receiving unit 11 receives information and instructions input using an input device such as a keyboard, a mouse, a touch panel, or a button. For example, the input receiving unit 11 receives input of parameters necessary for reactor core calculation. The input receiving unit 11 records the received information in the storage unit 14 or outputs it to the nuclear constant calculation unit 12.

[0016] The nuclear constant calculation unit 12 (1) calculates nuclear constants such as various cross sections. These nuclear constants are base nuclear constants before being multiplied by correction factors. The method for calculating these nuclear constants is not limited to a specific method. For example, as described in the "Background Art" section, the nuclear constants may be calculated in a two-dimensional radial system with infinite axial direction. (2) The nuclear constant calculation unit 12 also calculates the axial distribution of correction factors for the nuclear constants. Specifically, the nuclear constant calculation unit 12 calculates the axial neutron flux distribution by a three-dimensional heterogeneous single-assembly calculation (a calculation in which a three-dimensional heterogeneous transport calculation is performed focusing on one fuel assembly), calculates the axial neutron flux distribution by a three-dimensional homogeneous single-assembly calculation (a calculation in which a three-dimensional homogeneous transport calculation is performed focusing on one fuel assembly), and based on these calculation results (neutron flux ratios), calculates the axial distribution of correction factors in the core so that the reaction rate obtained using the neutron flux calculated by inputting the nuclear constant obtained by multiplying the correction factors in the three-dimensional homogeneous single-assembly calculation is equal to the reaction rate obtained by the three-dimensional heterogeneous single-assembly calculation (so that the reaction rate can be preserved). At this time, the nuclear constant calculation unit 12 calculates the axial correction factor by a method combining the reaction rate ratio conservation method (Non-Patent Document 1) and the Simultaneous-SPH method (Non-Patent Document 2). The nuclear constant calculation unit 12 multiplies various nuclear constants such as cross sections by correction factors at each position in the core axis direction to calculate nuclear constants that reflect the effect of neutron inflow through heterogeneous media in the core axis direction. Note that the "reaction rate obtained using the neutron flux calculated by inputting the nuclear constant obtained by multiplying by the correction factor" can be calculated using the nuclear constant Σ, neutron flux Φ, and nuclear constant correction factor f using the following formula: Reaction rate = (f × Σ) × Φ, where Φ is the output of the three-dimensional homogeneous single-assembly calculation when f × Σ is input.

[0017] The core calculation unit 13 performs a three-dimensional core calculation using the nuclear constants at each position in the core axial direction calculated by the nuclear constant calculation unit 12 as input, and calculates core nuclear characteristics such as the effective multiplication factor, reactivity coefficient, power distribution, and sodium void reactivity.

[0018] The memory unit 14 stores various setting information, processing data during calculation, etc. The memory unit 14 also stores a heterogeneous transport calculation code 141, a homogeneous transport calculation code 142, and a core calculation code 143. The heterogeneous transport calculation code 141 is a computer program that calculates neutron flux distribution, etc., using a three-dimensional heterogeneous single-assembly calculation. The homogeneous transport calculation code 142 is a computer program that calculates neutron flux distribution, etc., using a three-dimensional homogeneous single-assembly calculation. The core calculation code 143 is a computer program that performs neutron transport calculations (or diffusion calculations) in a three-dimensional core system and calculates various nuclear characteristics, such as the power distribution in the core and the effective multiplication factor.

[0019] Figure 2 shows a radial cross section of a core 1. The core 1 is composed of multiple assemblies 2, each of which is composed of a large number of fuel 3. In a typical core analysis, the cross-sectional area of ​​each medium is calculated for the radial cross section of the core 1, taking into account the influence of different media such as fuel pellets, cladding, and coolant. The homogenized cross-sectional area for each medium in the assembly 2 is then applied to the corresponding media assembly in the core 1 for core calculation. However, this method does not take into account the difference in neutron flux distribution in the axial direction of the core. Figure 3 shows an axial cross section of a fast reactor. Figure 3(a) shows a cross section of an axially homogeneous core, and Figure 3(b) shows a cross section of an axially heterogeneous core. An axially homogeneous core is composed of the following layers, from bottom to top: lower shielding, lower blanket, fuel region, upper blanket, and upper shielding. In the case of an axially heterogeneous core, it is composed of the following layers from bottom to top: lower shielding, lower blanket, fuel region, inner blanket, fuel region, sodium plenum, and upper shielding. Because the neutron flux distribution in each layer is different, the accuracy of the core neutron characteristics calculated by the general core analysis described above may not be sufficient.

[0020] Therefore, in this embodiment, nuclear constants are calculated taking into account not only the radial direction of the core but also the influence of heterogeneous media in the axial direction. The calculation of the axial nuclear constant is described below. First, a heterogeneous transport calculation is performed using a three-dimensional single-assembly system that simulates the axial material composition of the core. An axial distribution of the nuclear constant correction factor is created in advance so that the obtained axial reaction rate distribution can be reproduced in a homogeneous calculation. That is, a homogeneous transport calculation is performed using a three-dimensional single-assembly system, and the corrected nuclear constant obtained by multiplying the corrected nuclear constant by the nuclear constant correction factor is input. An axial distribution of the correction factor is created so that the axial reaction rate distribution obtained using the calculated neutron flux is the same as the axial reaction rate distribution obtained by performing a heterogeneous transport calculation using the three-dimensional single-assembly system. Next, the nuclear constant is corrected using the created correction factor, and the corrected nuclear constant is applied to a three-dimensional whole-core calculation, thereby improving the prediction accuracy of the core nuclear characteristics. In this way, by creating an axial distribution of the correction factor in advance and performing a core calculation after correcting the nuclear constant with the correction factor, nuclear characteristics can be calculated with high accuracy. Although various techniques for reducing the homogenization error in the radial direction of the core have been proposed, according to this embodiment, the homogenization error in the axial direction of the core can be reduced.

[0021] Figure 4 shows an example of the axial correction factor distribution. The vertical axis of Figure 4 shows the correction factor, and the horizontal axis shows the distance and height from the bottom of the core. 0 to h0 are the lower shielding, h0 to h1 are the lower blanket, h1 to h2 are the fuel region, h2 to h3 are the upper blanket, and h3 to h4 are the upper shielding. L1 is a graph showing the correction factor for each axial position.

[0022] The correction factor is calculated as follows: A point in the axial direction of the core is set as the reference point. For example, in the case of Figure 4, the center of the height direction of the fuel region is set as reference point P. Next, in accordance with the reaction rate ratio conservation method, the correction factor is calculated assuming that the ratio of the reaction rate at reference point P to the reaction rate at each position in the axial direction is preserved when a heterogeneous transport calculation is performed and when a homogeneous transport calculation is performed.

[0023]

number

[0024] R on the left side of the above equation (1) het ig represents the reaction rate obtained by the heterogeneous transport calculation. On the right side, φ(r, E) represents the neutron flux, Σ(r, E) represents the cross section, i represents the homogenized region, g represents the neutron energy group, r represents the position, and E represents the neutron energy. The reaction rate R~ obtained by the homogeneous transport calculation hom ig can be expressed by the following equation (2). hom ig is the neutron flux at i and g obtained by homogeneous transport calculation, Σ ig are the cross sections at i and g obtained by homogeneous transport calculation, and f ig represents the correction factor in i and g.

[0025]

number

[0026] In the reaction rate ratio conservation method, the following equation is assumed to hold, and the correction factor f ig Calculate the subscript 0 of the denominator. ig = 1 (i.e., reference point P).

[0027]

number

[0028] Substituting equations (1) and (2) into equation (3), we obtain the correction factor f ig By rearranging the above, we obtain the following equation (4).

[0029]

number

[0030] The correction factor f is calculated by equation (4). igOnce the correction factor is obtained, the nuclear constants are corrected by the Simultaneous-SPH method. An example of the correction method is shown below.

[0031]

number

[0032] Equation (5) is one of the equations used in homogeneous transport calculations. Ω is a vector representing the neutron flight direction, Ψ is the angular neutron flux, and f is a correction factor f for each energy group and position. ig (The value obtained from equation (4)). Σ tr is the transport cross section, Σ s is the scattering cross section, vΣ f is the production cross section. For example, the transport cross section Σ tr is corrected by multiplying it by 1 / f, and the scattering cross section Σ s For the cross section vΣ, the correction is made by multiplying by f. f is corrected by multiplying by f. Also, the self-group scattering cross section Σ, which indicates the probability of transition from the same energy group to the same energy group before and after the transition, sg→g is corrected using the following equation (6). sg→g is the corrected self-scattering cross section.

[0033]

number

[0034] The nuclear constant calculation unit 12 calculates the correction factor f ig Calculation of the calculated correction factor f ig Correction of the nuclear constants using, 3D homogeneous transport calculation using the corrected nuclear constants, and correction factor f ig Repeat the calculation until convergence. Correction factor f ig When convergence occurs, the correction factor f ig The nuclear constants are corrected by the Simultaneous-SPH method using the above formula, and the nuclear constants after correction are input to the reactor core calculation unit 13 to perform reactor core calculations.

[0035] 4 at each position of the lower shielding is input to the core calculation code 143 as the nuclear constant at the corresponding axial position of the lower shielding. Similarly, the value obtained by multiplying the correction factor at each position of the lower blanket by the nuclear constant of the lower blanket is input to the core calculation code 143 as the nuclear constant at the corresponding axial position of the lower blanket, the value obtained by multiplying the correction factor at each position of the fuel region by the nuclear constant of the fuel region is input to the core calculation code 143 as the nuclear constant at the corresponding axial position of the fuel region, the value obtained by multiplying the correction factor at each position of the upper blanket by the nuclear constant of the upper blanket is input to the core calculation code 143 as the nuclear constant at the corresponding axial position of the upper blanket, and the value obtained by multiplying the correction factor at each position of the upper shielding by the nuclear constant of the upper shielding is input to the core calculation code 143 as the nuclear constant at the corresponding axial position of the upper shielding.

[0036] (operation) Next, the operation of the reactor core analyzer 10 will be described with reference to FIG. FIG. 5 is a flowchart showing an example of a core analysis process according to the embodiment. First, a user inputs information necessary for core analysis, such as the core configuration illustrated in FIG. 3, into the core analysis device 10 and instructs the device to execute the core analysis. The input receiving unit 11 receives the input information and an instruction to execute the core analysis. The core analysis device 10 then executes the following process. First, the nuclear constant calculation unit 12 performs a three-dimensional heterogeneous single-assembly calculation (step S1). The nuclear constant calculation unit 12 uses the heterogeneous transport calculation code 141 to perform a three-dimensional neutron transport calculation to calculate the neutron flux distribution in the axial direction of the core (step S2). Next, the nuclear constant calculation unit 12 calculates the homogeneous nuclear constant. The homogeneous nuclear constant is the base nuclear constant described above. Various methods for calculating the homogeneous nuclear constant may be used. For example, it may be calculated using an axially infinite two-dimensional radial heterogeneous calculation. Furthermore, it is not necessary to calculate the homogeneous nuclear constant after step S2, and it may be calculated in advance before executing the process of the flowchart in FIG. 5. When the fast reactor has an axially homogeneous core, the nuclear constant calculation unit 12 calculates the nuclear constants for each of the upper shielding, upper blanket, fuel region, lower blanket, and lower shielding. This allows the homogeneous nuclear constants to be calculated individually for each layer of the upper shielding, upper blanket, fuel region, lower blanket, and lower shielding.

[0037] Next, the nuclear constant calculation unit 12 performs a three-dimensional homogeneous single-assembly calculation using the calculated homogeneous nuclear constant as an input (step S3). The nuclear constant calculation unit 12 performs a three-dimensional neutron transport calculation using the homogeneous transport calculation code 142 to calculate the distribution of neutron flux in the core axis direction. Next, the nuclear constant calculation unit 12 calculates the distribution of homogeneous neutron flux in the core axis direction to calculate a correction factor (step S4). The nuclear constant calculation unit 12 aggregates the neutron flux calculated in step S1 and the neutron flux calculated in step S3, and calculates the homogeneous neutron flux at each position at a predetermined interval in the core axis direction. Next, the nuclear constant calculation unit 12 calculates the correction factor (step S5). The nuclear constant calculation unit 12 calculates the correction factor f ig Next, the nuclear constant calculation unit 12 calculates the correction factor f ig It is determined whether or not the correction factor f has converged (step S6). ig and the previously calculated correction factor fig The difference between the current and previous values ​​is calculated for each position in the core axis direction, and if the difference between the current and previous values ​​at all positions is within a predetermined threshold, the correction factor f ig is determined to have converged, otherwise the correction factor f ig is not converged. Correction factor f ig When it is determined that the values ​​of the homogeneous nuclear constants (base nuclear constants) input in the previous three-dimensional homogeneous single-assembly calculation have not converged (step S6; No), the nuclear constant calculation unit 12 corrects each of the homogeneous nuclear constants (base nuclear constants) input in the previous three-dimensional homogeneous single-assembly calculation by the Simultaneous-SPH method (step S7). The nuclear constant calculation unit 12 corrects the correction factor f by the reaction rate ratio preservation method calculated in step S5, instead of the correction factor (SPH factor) by the SPH method, which is obtained by preserving the reaction rate by the three-dimensional heterogeneous transport calculation and the reaction rate by the three-dimensional homogeneous transport calculation, as in the general Simultaneous-SPH method. ig For example, as explained in equation (5) above, the transport cross section is corrected by 1 / f ig For scattering and production cross sections, ig The self-group scattering cross section is corrected by multiplying by . The homogeneous nuclear constant is calculated for each of the upper shield, upper blanket, fuel region, lower blanket, and lower shield, so the correction factor f ig The correction factor f for the homogeneous nuclear constant depending on the position of ig For example, the transport cross section of the upper shield Σ tr If so, the homogenized transport cross section of the upper shield Σ tr , the correction factor f calculated for each position of the upper shield ig 1 / f using ig Multiplying by the transport cross section Σ tr Correct the following.

[0038] correction factor f ig When it is determined that the correction factor f has converged (step S6; Yes), the nuclear constant calculation unit 12 calculates the correction factor f igBy correcting each of the homogeneous nuclear constants (base nuclear constants) by the Simultaneous-SPH method using the above, the nuclear constants that take into account the effect of neutron flow due to heterogeneous media in the axial direction of the core are calculated (step S8). As described above, the nuclear constant calculation unit 12 calculates the nuclear constants that take into account the effect of neutron flow due to heterogeneous media in the axial direction of the core by applying the correction factor f ig The nuclear constants at each position of the lower shielding are calculated by multiplying the nuclear constants by the Simultaneous-SPH method described using Equations (5) and (6). The correction method for the homogeneous nuclear constants is the same as that for the other layers (upper shielding, upper blanket, fuel region, and lower blanket). Next, the core calculation unit 13 performs a three-dimensional whole-core transport calculation (or diffusion calculation) using the core calculation code 143 (step S9). The core calculation unit 13 acquires the nuclear constants calculated in step S8 and inputs them into the core calculation code 143. The core calculation code 143 performs a three-dimensional whole-core transport calculation (or diffusion calculation) by applying the nuclear constants corresponding to each position in the core axial direction, and calculates core nuclear characteristics such as the effective multiplication factor, reactivity coefficient, power distribution, and sodium void reactivity.

[0039] (effect) As described above, according to this embodiment, a correction factor distribution in the core axis direction is calculated so that the reaction rate distribution in the core axis direction obtained by the heterogeneous single-assembly calculation can be reproduced by the homogeneous single-assembly calculation. Then, by multiplying the homogeneous nuclear constant by the correction factor, a nuclear constant that takes into account the effect of neutron inflow due to heterogeneous media in the core axis direction is calculated. By performing a core calculation using this nuclear constant, the prediction accuracy of the overall nuclear characteristics of a fast reactor (particularly the effective multiplication factor, sodium void reactivity, and axial power distribution) can be improved. The correction factor calculation method of this embodiment can be applied to fast reactors in general, including axially homogeneous cores and axially heterogeneous cores, as exemplified in FIG. 3 . Furthermore, the application of the correction factor calculation method of this embodiment is not limited to fast reactors.

[0040] Furthermore, if it is confirmed that fluctuations in core conditions (e.g., the plutonium content in the fuel region) do not have much effect on the distribution of correction factors, the calculation results of the axial distribution of correction factors obtained by this embodiment using a representative core as a model can also be applied to core analyses of other fast reactors.

[0041] FIG. 6 is a diagram illustrating an example of a hardware configuration of a reactor core analyzer. The computer 900 includes a CPU 901 , a main memory device 902 , an auxiliary memory device 903 , an input / output interface 904 , and a communication interface 905 . The above-described reactor core analyzer 10 is implemented in a computer 900. Each of the above-described functions is stored in the auxiliary storage device 903 in the form of a program. The CPU 901 reads the program from the auxiliary storage device 903, loads it into the main storage device 902, and executes the above-described processing in accordance with the program. The CPU 901 also allocates a storage area in the main storage device 902 in accordance with the program. The CPU 901 also allocates a storage area in the auxiliary storage device 903 for storing data being processed in accordance with the program.

[0042] A program for implementing all or part of the functions of the reactor core analyzer 10 may be recorded on a computer-readable recording medium, and the program may be loaded into a computer system and executed to perform processing by each functional unit. The term "computer system" as used herein includes hardware such as an OS and peripheral devices. If a WWW system is used, the term "computer system" also includes a homepage providing environment (or display environment). The term "computer-readable recording medium" refers to portable media such as CDs, DVDs, and USBs, as well as storage devices such as hard disks built into the computer system. If the program is distributed to the computer 900 via a communication line, the computer 900 may load the program into the main storage device 902 and execute the processing described above. The program may be for implementing part of the functions described above, or may be capable of implementing the functions described above in combination with a program already stored in the computer system.

[0043] As described above, several embodiments according to the present disclosure have been described, but all of these embodiments are presented as examples and are not intended to limit the scope of the invention. These embodiments can be implemented in various other forms, and various omissions, substitutions, and modifications can be made without departing from the spirit of the invention. These embodiments and their modifications are included in the scope of the invention and its equivalents as defined in the claims, as well as in the scope and spirit of the invention.

[0044] <Additional Notes> The calculation method, calculation device, and program described in the embodiments can be understood, for example, as follows.

[0045] (1) A calculation method according to a first aspect is a calculation method executed by a computer, and includes the steps of: calculating a first neutron flux distribution in the axial direction of a core by a three-dimensional heterogeneous single-assembly calculation; calculating a second neutron flux distribution in the axial direction of the core by a three-dimensional homogeneous single-assembly calculation; calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation to which a correction factor based on a ratio between the first neutron flux and the second neutron flux is applied, so that the reaction rate based on the first neutron flux distribution (Equation (1)) can be reproduced; updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and repeating the steps of calculating the axial distribution of the correction factor and updating the second neutron flux distribution until the correction factor converges. By calculating the distribution of the correction factor in the axial direction of the core, it is possible to calculate the nuclear constant taking into account the effect of neutron inflow due to heterogeneous media in the axial direction of the core.

[0046] (2) A core analysis method according to a second aspect is a core analysis method according to (1), further comprising the steps of correcting the axial nuclear constant of the core by correcting the nuclear constant of the core using the converged correction factor, and performing a three-dimensional whole core calculation by applying the corrected nuclear constant. This allows the core nuclear characteristics to be calculated with high accuracy.

[0047] (3) A calculation method according to a third aspect is the calculation method of (2), wherein the core is composed of multiple layers of different media in the axial direction, and in the step of correcting the nuclear constant, the nuclear constant is corrected by applying the correction factor for each axial position of the layer to the nuclear constant homogenized for each layer. By correcting the homogeneous nuclear constants with the correction factors, the core nuclear characteristics can be calculated with high accuracy.

[0048] (4) A calculation method according to a fourth aspect is the calculation method of (1) to (3), wherein in the step of calculating the axial distribution of the correction factor, the correction factor is calculated by a reaction rate ratio conservation method. By using the reaction rate ratio conservation method, it is possible to avoid the correction factor not converging, and to stably calculate the correction factor.

[0049] (5) A calculation method according to a fifth aspect is the calculation method of (1) to (4), wherein in the step of updating the second neutron flux distribution, the nuclear constant is corrected by the Simultaneous-SPH method using the correction factor, and the corrected nuclear constant is used as an input to perform the three-dimensional homogeneous single-assembly calculation. By correcting the nuclear constants using the simultaneous-SPH method with the correction factor of the reaction rate ratio conservation method, the nuclear constants in the axial direction of the core can be calculated (corrected) with high accuracy.

[0050] (6) A calculation method according to a sixth aspect is the calculation method of any one of (1) to (5), wherein the core is a core of a fast reactor. This allows for accurate calculation of the core nuclear characteristics of fast reactors.

[0051] (7) A calculation device according to a seventh aspect includes: a means for calculating a first neutron flux distribution in the axial direction of a core by a three-dimensional heterogeneous single-assembly calculation; a means for calculating a second neutron flux distribution in the axial direction of the core by a three-dimensional homogeneous single-assembly calculation; a means for calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation to which a correction factor based on a ratio between the first neutron flux and the second neutron flux is applied, so that the reaction rate based on the first neutron flux distribution (Equation (1)) can be reproduced; a means for updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and a means for repeating the steps of calculating the axial distribution of the correction factor and updating the second neutron flux distribution until the correction factor converges.

[0052] (8) A program according to an eighth aspect causes a computer to execute the following steps: calculating a first neutron flux distribution in the axial direction of a core by a three-dimensional heterogeneous single-assembly calculation; calculating a second neutron flux distribution in the axial direction of the core by a three-dimensional homogeneous single-assembly calculation; calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation to which a correction factor based on a ratio between the first neutron flux and the second neutron flux is applied, so that the reaction rate based on the first neutron flux distribution (Equation (1)) can be reproduced; updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and repeating the steps of calculating the axial distribution of the correction factor and updating the second neutron flux distribution until the correction factor converges. [Explanation of symbols]

[0053] 10. Core analysis equipment 11 Input reception section 12...Nuclear constant calculation part 13. Core Calculation Section 14...Storage section 141...Heterogeneous Transport Calculation Code 142 Homogeneous Transport Calculation Code 143 Core Calculation Code 900···Computer 901 CPU 902...Main memory 903...Auxiliary storage device 904 Input / Output Interface 905···Communication Interface

Claims

1. 1. A computer-implemented computational method comprising: calculating a first neutron flux distribution in the axial direction of the core by a three-dimensional heterogeneous single-assembly calculation; calculating a second neutron flux distribution in the axial direction of the core by a three-dimensional homogeneous single-ensemble calculation; calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation to which a nuclear constant correction factor based on the ratio of the first neutron flux to the second neutron flux is applied, so that the reaction rate based on the distribution of the first neutron flux can be reproduced; updating the distribution of the second neutron flux by the three-dimensional homogeneous single ensemble calculation reflecting the correction factor; repeating the steps of calculating the axial distribution of the correction factor and updating the distribution of the second neutron flux until the correction factor converges; A calculation method having the following.

2. correcting the nuclear constant of the core in the axial direction by correcting the nuclear constant of the core using the converged correction factor; applying the corrected nuclear constants to perform a three-dimensional whole core calculation; 10. The method of claim 1 further comprising:

3. The core is composed of a plurality of layers of different media in the axial direction, In the step of correcting the nuclear constant, correcting the homogenized nuclear constant for each layer with the correction factor for each axial position of the layer; The calculation method according to claim 2.

4. In the step of calculating the axial distribution of the correction factor, the correction factor is calculated by a reaction rate ratio conservation method. The calculation method according to claim 1 or claim 2.

5. In the step of updating the distribution of the second neutron flux, a nuclear constant is corrected by a Simultaneous-SPH method using the correction factor, and the corrected nuclear constant is used as an input to perform the three-dimensional homogeneous single-ensemble calculation. The calculation method according to claim 1 or claim 2.

6. The core is a fast reactor core. The calculation method according to claim 1 or claim 2.

7. means for calculating a first neutron flux distribution in an axial direction of the reactor core by a three-dimensional heterogeneous single-assembly calculation; means for calculating a second neutron flux distribution in the axial direction of the core by a three-dimensional homogeneous single-ensemble calculation; means for calculating an axial distribution of a nuclear constant correction factor based on the ratio of the first neutron flux to the second neutron flux, by the three-dimensional homogeneous single-assembly calculation to which the correction factor is applied, so that a reaction rate based on the distribution of the first neutron flux can be reproduced; means for updating the distribution of the second neutron flux by the three-dimensional homogeneous single ensemble calculation reflecting the correction factor; means for repeating the steps of calculating the axial distribution of the correction factor and updating the distribution of the second neutron flux until the correction factor converges; A computing device having:

8. On the computer, calculating a first neutron flux distribution in the axial direction of the core by a three-dimensional heterogeneous single-assembly calculation; calculating a second neutron flux distribution in the axial direction of the core by a three-dimensional homogeneous single-ensemble calculation; calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation to which a nuclear constant correction factor based on the ratio of the first neutron flux to the second neutron flux is applied, so that the reaction rate based on the distribution of the first neutron flux can be reproduced; updating the distribution of the second neutron flux by the three-dimensional homogeneous single ensemble calculation reflecting the correction factor; repeating the steps of calculating the axial distribution of the correction factor and updating the distribution of the second neutron flux until the correction factor converges; A program that executes the following.

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