Calculation method, calculation device, and program

Abstract

Provided is a method for calculating nuclear constants accounting for effects of heterogeneous media in the reactor core axial direction. Provided is a calculation method that is executed by a computer, the calculation method comprising: a step of calculating the distribution of a first neutron flux in the axial direction of a reactor core through a three-dimensional heterogeneous single-assembly calculation; a step of calculating the distribution of a second neutron flux in the axial direction of the reactor core through a three-dimensional homogeneous single-assembly calculation; a step of calculating an axial distribution of correction factors, which are based on the ratio between the first neutron flux and the second neutron flux, that enables a three-dimensional homogeneous single-assembly calculation with nuclear constants corrected using the correction factors to reproduce the reaction rate based on the distribution of the first neutron flux; a step of updating the distribution of the second neutron flux through a three-dimensional homogeneous single-assembly calculation reflecting the correction factors; a step of calculating the axial distribution of the correction factors until the correction factors converge; and a step of repeating the step of updating the distribution of the second neutron flux.

Description

Calculation method, calculation device, and program

[0001] This disclosure claims priority to Japanese Patent Application No. 2024-101988, filed on June 25, 2024, the contents of which are incorporated herein by reference.

[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 heterogeneous neutron transport calculation is performed on 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 geometric 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 an 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 a homogeneous neutron diffusion or transport calculation on the entire three-dimensional core system homogenized within the assembly, and core nuclear characteristics 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 first stage of calculating the axially infinite nuclear constants. In particular, in fast reactor systems, the mean free path of neutrons (the average flight distance from when 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 handling of the axially infinite in the first stage becomes a source of error in the axial reaction rate distribution and the associated overall core nuclear characteristics.

[0004] Patent Literature 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 Literature 1 discloses the reaction-rate ratio preservation method as a method for preserving reaction rates obtained from three-dimensional heterogeneous transport calculations in homogeneous calculations, and Non-Patent Literature 2 discloses the Simultaneous-SPH method, which is an extension of the SPH method to enable it to be used in homogeneous transport calculations.

[0005] Japanese Patent Application Laid-Open No. 2003-66179

[0006] S.Kosaka, et al., New Control Rod Homogenization Method forFast Reactors, J.Nucl.Sci.Technol., 31, 7, pp.647-653, 1994.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.

[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-mentioned problems.

[0009] According to one aspect of the present disclosure, a calculation method is a calculation method executed by a computer, comprising the steps of: calculating a first neutron flux distribution in an axial direction of a reactor core by a three-dimensional non-homogeneous 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 as to reproduce a 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.

[0010] According to one aspect of the present disclosure, a calculation device includes: means for calculating a first neutron flux distribution in an axial direction of a reactor core by a three-dimensional heterogeneous single-assembly calculation; means for calculating a second neutron flux distribution in the axial direction of the reactor core by a three-dimensional homogeneous single-assembly calculation; 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 a reaction rate based on the first neutron flux distribution can be reproduced by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and 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] According to one aspect of the present disclosure, a program 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 non-homogeneous 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 applying a nuclear constant correction factor based on a ratio between the first neutron flux and the second neutron flux so as to reproduce a 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.

[0012] According to the above-described calculation method, calculation device, and program, by calculating the correction factor of the nuclear constant at each position in the axial direction of the core, it is possible to calculate the nuclear constant taking into account the effect of neutron flow due to heterogeneous media in the axial direction of the core.

[0013] Fig. 1 is a block diagram showing an example of a core analysis device according to an embodiment; Fig. 2 is a diagram showing an example of a core cross section in the radial direction according to an embodiment; Fig. 3 is a diagram showing an example of a core cross section in the axial direction according to an embodiment; Fig. 4 is a diagram showing an example of a correction factor according to an embodiment; Fig. 5 is a flowchart showing an example of a core analysis process according to an embodiment; Fig. 6 is a diagram showing an example of a hardware configuration of a core analysis device according to an embodiment;

[0014] <Embodiments> A reactor core analyzer according to the present disclosure will be described below with reference to Figures 1 to 6. (Configuration) Figure 1 is a block diagram showing an example of a reactor core analyzer according to an embodiment. When performing reactor core calculations for a fast reactor, the reactor core analyzer 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 position in the axial direction of the core, thereby performing reactor core analysis with high accuracy. The reactor core analyzer 10 includes an input receiving unit 11, a nuclear constant calculation unit 12, a reactor core calculation unit 13, and a storage 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 and 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 an axially infinite two-dimensional radial system. (2) The nuclear constant calculation unit 12 calculates the axial distribution of correction factors for the nuclear constants. Specifically, the nuclear constant calculation unit 12 calculates the axial neutron flux distribution using 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 using 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 constants 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 factors using a method that combines 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. 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 storage unit 14 stores various setting information, processing data during calculation, etc. The storage unit 14 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 assemblies 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. Then, the cross-sectional area homogenized for each medium in the assembly 2 is applied to the corresponding media assembly in the core 1 for core calculation. However, this method does not take into account differences in the 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, from bottom to top, the lower shielding, lower blanket, fuel region, upper blanket, and upper shielding layers. 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 above-mentioned general core analysis 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 neutron flux calculated using the homogeneous transport calculation 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 by 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] An example of the axial correction factor distribution is shown in Figure 4. The vertical axis of Figure 4 shows the correction factor, and the horizontal axis shows the distance / 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 the 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 the 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]

[0024] R on the left side of the above formula (1) ｈｅｔ ｉｇrepresents 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 homogenization target 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]

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

[0027]

[0028] Substituting equations (1) and (2) into equation (3), the correction factor f ｉｇ When rearranged, the following equation (4) is obtained.

[0029]

[0030] The correction factor f is calculated by equation (4). ｉｇ Once 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]

[0032] Equation (5) is one of the equations used in homogeneous transport calculations. Ω is a vector representing the flight direction of neutrons, Ψ is the angular neutron flux, and f is a correction factor f for each energy group and position. ｉｇ (the value obtained by equation (4)). 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 ｔｒis corrected by multiplying it by 1 / f, and the scattering cross section Σ ｓ is corrected by multiplying by f, and the production cross section vΣ ｆ The self-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, is ｓｇ→ｇ is corrected using the following equation (6): Σ ~ ｓｇ→ｇ is the corrected self-scattering cross section.

[0033]

[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 constant using the corrected nuclear constant, 3D homogeneous transport calculation using the corrected nuclear constant, and correction factor f ｉｇ Repeat the calculation until the 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 calculated at each position of the lower shielding by the nuclear constant 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 core analyzer 10 will be described using FIG. 5 . 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 analyzer 10 and instructs the core analyzer 10 to execute the core analysis. The input receiving unit 11 receives the input information and an instruction to execute the core analysis. The core analyzer 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 performs a three-dimensional neutron transport calculation using the heterogeneous transport calculation code 141 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, the calculation may be performed using an axially infinite two-dimensional radial heterogeneous calculation. The calculation does not need to be performed after step S2, and may be performed 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. As a result, the homogeneous nuclear constants are 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 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 neutron flux distribution in the core axis direction. Next, the nuclear constant calculation unit 12 calculates the homogeneous neutron flux distribution 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 at each position at a predetermined interval in the core axis direction using equation (4). ｉｇ Next, the nuclear constant calculation unit 12 calculates the correction factor f ｉｇIt is determined whether or not the correction factor f has converged (step S6). ｉｇ and the previously calculated correction factor f ｉｇ The difference between the current and previous values ​​is calculated for each position in the core axis direction, and when 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, and if not, the correction factor f ｉｇ It is determined that the correction factor f ｉｇ When it is determined that the values ​​of the homogeneous nucleon constants (basic nucleon 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 nucleon constants (basic 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 the above equation (5), the nuclear constant is corrected by 1 / f ｉｇ The scattering and production cross sections are corrected by multiplying by f ｉｇ The self-scattering cross section is corrected by multiplying by ∑ ... ｉｇ The correction factor f 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 has converged (step S6; Yes), the nuclear constant calculation unit 12 calculates the correction factor fｉｇ By correcting each of the homogeneous nuclear constants (base nuclear constants) by the Simultaneous-SPH method using the above, the nuclear constant calculation unit 12 calculates the nuclear constants taking into account the effect of neutron flow due to heterogeneous media in the axial direction of the core (step S8). As described above, the nuclear constant calculation unit 12 applies the correction factor f ｉｇ The nuclear constants at each position of the lower shield are calculated by multiplying the values ​​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 shield, 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] (Effects) 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 all fast reactors, including the axially homogeneous core and the axially heterogeneous core illustrated in FIG. 3 . The application of the correction factor calculation method of this embodiment is not limited to fast reactors.

[0040] If it is confirmed that fluctuations in core conditions (e.g., the plutonium content in the fuel region) do not significantly affect 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] 6 is a diagram showing an example of the hardware configuration of a reactor core analyzer. A computer 900 includes a CPU 901, a main storage device 902, an auxiliary storage device 903, an input / output interface 904, and a communication interface 905. The above-described reactor core analyzer 10 is implemented in the computer 900. The above-described functions are 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 allocates a storage area in the main storage device 902 in accordance with the program. The CPU 901 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 website provision environment (or display environment). The term "computer-readable recording medium" refers to portable media such as CDs, DVDs, and USBs, and 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, comprising 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 of 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 (Equation (1)); 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 a nuclear constant that takes into account the effect of neutron flow due to heterogeneous media in the axial direction of the core.

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

[0048] (4) A fourth aspect of the calculation method is the calculation method of any one of (1) to (3), wherein the step of calculating the axial distribution of the correction factor calculates the correction factor using 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 any one 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 constant by the Simultaneous-SPH method using the correction factor of the reaction rate ratio conservation method, the nuclear constant 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 makes it possible to calculate the core neutronics characteristics of a fast reactor with high accuracy.

[0051] (7) A calculation device according to a seventh aspect includes: means for calculating a first neutron flux distribution in the axial direction of a reactor core by a three-dimensional heterogeneous single-assembly calculation; means for calculating a second neutron flux distribution in the axial direction of the reactor core by a three-dimensional homogeneous single-assembly calculation; 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; means for updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and 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.

[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.

[0053] According to the above-described calculation method, calculation device, and program, by calculating the correction factor of the nuclear constant at each position in the axial direction of the core, it is possible to calculate the nuclear constant taking into account the effect of neutron flow due to heterogeneous media in the axial direction of the core.

[0054] DESCRIPTION OF SYMBOLS 10: Reactor core analysis device 11: Input reception unit 12: Nuclear constant calculation unit 13: Reactor core calculation unit 14: Storage unit 141: Heterogeneous transport calculation code 142: Homogeneous transport calculation code 143: Reactor core calculation code 900: Computer 901: CPU 902: Main storage unit 903: Auxiliary storage unit 904: Input / output interface 905: Communication interface

Claims

1. A calculation method executed by a computer, comprising: a step of calculating a first neutron flux distribution in the axial direction of a reactor core by a three-dimensional non-homogeneous single-assembly calculation; a step of calculating a second neutron flux distribution in the axial direction of the reactor core by a three-dimensional homogeneous single-assembly calculation; a step of calculating an axial distribution of the correction factor by the three-dimensional homogeneous single-assembly calculation applying a nuclear constant correction factor based on the ratio of the first neutron flux to the second neutron flux so as to reproduce the reaction rate based on the first neutron flux distribution; a step of updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation reflecting the correction factor; and a step of repeating the steps of calculating the axial distribution of the correction factor and updating the second neutron flux distribution until the correction factor converges.

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

3. The calculation method according to claim 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 homogenized for each layer is corrected by the correction factor for each axial position of the layer.

4. The calculation method according to claim 1 or 2, 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.

5. The calculation method according to claim 1 or 2, wherein in the step of updating the distribution of the second neutron flux, 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-ensemble calculation.

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

7. A calculation device comprising: means for calculating a first neutron flux distribution in the axial direction of a reactor core by a three-dimensional non-homogeneous single-assembly calculation; means for calculating a second neutron flux distribution in the axial direction of the reactor core by a three-dimensional homogeneous single-assembly calculation; 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 the ratio of the first neutron flux to the second neutron flux is applied, so that a reaction rate based on the first neutron flux distribution can be reproduced by the three-dimensional homogeneous single-assembly calculation to which the correction factor is applied; means for updating the second neutron flux distribution by the three-dimensional homogeneous single-assembly calculation that reflects the correction factor; and 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.

8. A program causing a computer to execute the following steps: calculating a first neutron flux distribution in the axial direction of a reactor core by a three-dimensional non-homogeneous 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 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 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.

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