JP5557998B2 - Core analysis program - Google Patents

Description

  The present invention relates to a core analysis program capable of analyzing a neutron flux in a core.

  Conventionally, as a method of analyzing the neutron flux in the core, a method of analyzing using a neutron transport equation by a method of characteristics (MOC) is known (for example, see Non-Patent Document 1). This analysis method is a deterministic calculation method based on the integral type neutron transport equation. On the neutron path (neutron range) drawn in detail for the analysis target area consisting of multiple divided detailed areas, By solving using the basic equation of the characteristic curve method, it is possible to calculate the behavior of neutrons (specifically angular neutrons) on the neutron path, that is, annihilation and generation.

Shinya Kosaka, “Development and Application of Efficient Transport Calculation Method Based on Characteristic Method in Light Water Reactor Core Calculation”, 2005

  By the way, when calculating each detailed area and neutron flux divided by the conventional characteristic curve method, an error accompanying discretization occurs. At this time, if the discretization parameters such as the path width of the neutron path and the angle division of the neutron path are made detailed in order to improve the calculation accuracy, the discretization error can be reduced. However, if the discretization parameters are made detailed, the calculation amount increases and the calculation time becomes long. For this reason, in order to reduce the error accompanying discretization, the length of the neutron path passing through each detailed region is normalized by a volume correction factor. The volume correction factor is expressed as a ratio between the volume of the detailed region obtained analytically in advance and the volume of the detailed region obtained by calculation.

  Although the error due to discretization can be reduced by standardizing with the volume correction factor, volume correction will be performed if the discretization parameter is further coarsened (especially, the path width of the neutron path is widened). Correction due to factors is not sufficient, and errors due to discretization occur again.

  Therefore, an object of the present invention is to provide a core analysis program capable of reducing errors accompanying discretization and shortening the calculation time of neutron flux.

  The core analysis program of the present invention is executed on hardware and a cell storing fuel rods in a core analysis program capable of analyzing a neutron flux in a core by solving a neutron transport equation based on a characteristic curve method A cell group composed of a plurality of cells is set as an analysis target region, and a ray tracing step for generating a plurality of neutron paths on the analysis target region divided into a plurality of detailed regions, and each neutron path generated by the ray tracing step A neutron flux calculating step for solving the neutron transport equation based on the characteristic curve method and calculating the neutron flux of each detailed region, and in the neutron flux calculating step, each detailed region calculated based on the neutron path If each corresponding calculated solution is far from the analytical solution corresponding to each detailed area obtained analytically in advance, And a discharge neutron flux emitted from the incident neutron flux and the details area incident on the fine region and is equal to, and calculates the neutron flux of each detail area.

  In this case, it is preferable to use a volume correction factor that normalizes the length of the neutron path as an index for determining whether the calculated solution is far from the analytical solution.

  Also, in this case, if each volume correction factor corresponding to each detailed area is larger than a predetermined upper limit value set in advance, it is determined that the calculated solution is too small for the analytical solution, and the incident light is incident on each detailed area. It is preferable to make the neutron flux equal to the emitted neutron flux emitted from each detailed region.

  In this case, if each volume correction factor corresponding to each detailed region is smaller than a predetermined lower limit value set in advance, it is determined that the calculated solution is too large for the analytical solution, and the incident light enters each detailed region. It is preferable to make the neutron flux equal to the emitted neutron flux emitted from each detailed region.

  According to the core analysis program of the first aspect, the neutron flux can be calculated with high accuracy even if the path width of the neutron path is widened. That is, if the analytical solution corresponding to each detailed region and the calculated solution are far from each other, for example, even if normalized by a volume correction factor, an error due to discretization occurs in the calculated neutron flux. At this time, it is considered that the reason why the analytical solution and the calculated solution are separated from each other is that the detailed region is a very small region. As a result, when the detailed region is a small region, the error generated in the calculated neutron flux can be reduced by making the incident neutron flux incident on each detailed region equal to the emitted neutron flux emitted from each detailed region. Can be reduced. Therefore, even if the path width of the neutron path is wide, the neutron flux can be calculated with high accuracy, and the calculation time can be shortened as much as the path width of the neutron path can be widened. It becomes possible. Note that the analysis solution and the calculation solution corresponding to the detailed region include not only the volume of the detailed region but also the length of the neutron path in the detailed region.

  According to the core analysis program of claim 2, a new index is provided by using a volume correction factor that normalizes the length of the neutron path as an index for determining whether or not the calculated solution is far from the analytical solution. It is possible to easily determine whether or not the calculated solution is far from the analytical solution.

  According to the core analysis program of claim 3, it is possible to determine whether or not the calculated solution is far from the analytical solution by determining whether or not the volume correction factor is larger than the predetermined upper limit value. .

  According to the core analysis program of claim 4, it is possible to determine whether or not the calculated solution is far from the analytical solution by determining whether or not the volume correction factor is smaller than the predetermined lower limit value. .

  Hereinafter, a core analysis program according to the present invention will be described with reference to the accompanying drawings. The present invention is not limited to the following examples. In addition, constituent elements in the following embodiments include those that can be easily replaced by those skilled in the art or those that are substantially the same.

Here, FIG. 1 is a structural diagram schematically showing the core that is the target of the core design, and FIG. 2 schematically shows the fuel assembly that is the analysis target region of the core analysis program according to the present embodiment. FIG. FIG. 3 is an explanatory diagram showing an analysis target region divided into a plurality of detailed regions, and FIG. 4 is an explanatory diagram showing an analysis target region to which a neutron path is drawn. Further, FIG. 5 is an explanatory diagram showing a detailed image and a discretized image of the angular neutron flux calculated in the characteristic curve method, and FIG. 6 is a table relating to the sensitivity of the path width of the neutron path. FIG. 7 is a graph showing the distribution of the volume correction factor and the difference in the absorption reaction rate from the reference solution when the path width of the neutron path is 0.05 cm. FIG. 8 shows the path width of the neutron path. FIG. 9 is a graph showing the distribution of the volume correction factor when the neutron path is 0.2 cm and the difference from the reference solution of the absorption reaction rate. FIG. 9 shows the volume correction factor when the path width of the neutron path is 0.4 cm. It is the graph showing the difference from the reference solution of distribution and absorption reaction rate. Further, FIG. 10 is an explanatory diagram showing a discretization error when “V _cal <V _ref ” in a certain detailed region, and FIG. 11 shows “V _cal > V _ref ” in a certain detailed region. It is explanatory drawing showing the discretization error in the case of. FIG. 12 is a table relating to the filtering sensitivity when the path width of the neutron path is fixed, and FIG. 13 is a table relating to the filtering sensitivity optimally set when the path width of the neutron path is varied. .

  The core analysis program according to this embodiment is a program for analyzing the neutron flux distribution in the core 1, and predicts and evaluates the distribution and behavior of neutrons that mediate the nuclear reaction in the core 1. Then, the core design is performed based on the analysis result obtained by the core analysis program. At this time, as the core design, for example, in consideration of safety, combustion efficiency, fuel arrangement, and the like, a part of the fuel loaded in the core is replaced with a new fuel.

  Here, as shown in FIGS. 1 and 2, the core 1 that is the object of the core design is composed of a plurality of fuel assemblies 5, and is arranged in a geometric shape so as to maintain 90 degree symmetry. It is installed. The fuel is exchanged in units of 5 fuel assemblies. Each fuel assembly 5 includes a plurality of fuel rods 10, a plurality of cladding tubes 11 that cover the fuel rods 10, and a grid that bundles the plurality of cladding tubes 11. While being filled with the coolant 13, the some control rod 14 is comprised so that insertion is possible.

  The fuel assembly 5 is formed in a square cross section, and is composed of, for example, 17 × 17 cells 20. Of the 17 × 17 cells 20, the control rods 14 are inserted into 25 cells 20, and the fuel rods 10 are inserted into the other cells 20, respectively. Note that the outside of the fuel assembly used in the boiling water reactor (BWR) is covered with a channel box. In the case of BWR, there is an inter-assembly gap 12 outside the channel box, and in the case of a pressurized water reactor (PWR), outside the fuel assembly.

  Next, the core analysis program will be described. The core analysis program performs inner iterative calculation and outer iterative calculation, and the present invention is applied in the inner iterative calculation. In the inner iterative calculation, the neutron transport equation based on the characteristic curve method is solved, and the region average angle neutron flux in each detailed region i is calculated in the analysis target region of the fuel assembly 5 (details). Will be described later).

  Here, the calculation procedure of the area average angle neutron flux by the core analysis program will be specifically described. In the core analysis program, the analysis target region 30 is a square region having a cross-sectional area obtained by cutting the fuel assembly 5 in the radial direction. In addition, the core analysis program generates a plurality of neutron paths s on the analysis target region 30 divided into a plurality of detailed regions i, and a characteristic curve method for each neutron path s created by the ray tracing step. And a neutron flux calculating step of calculating a region average angle neutron flux of each detailed region i by solving a neutron transport equation based on

  As shown in FIG. 3, the analysis target region 30 that is a part of the fuel assembly 5 is an arbitrary system, and includes a plurality of cell regions 31 a and 31 b, and each cell region includes a plurality of detailed regions. divided into i. Each cell region 31a, 31b corresponds to the 17 × 17 cell 20 described above, and the number of divisions of the detailed region i of the cell region 31a in which the fuel rod 10 is inserted is the cell region in which the control rod 14 is inserted. The number of divisions is smaller than the number of divisions of the detailed area i of 31b. That is, the cell region 31b in which the control rod 14 is inserted has a more detailed division than the cell region 31a in which the fuel rod 10 is inserted.

  As shown in FIG. 4, the ray tracing step creates a plurality of neutron paths s in the analysis target region 30 as described above. That is, the neutron path s is created not in units of 20 cells but in units of 5 fuel assemblies. Here, the neutron path s is a line on which neutrons fly. In this ray tracing step, so-called cyclic ray tracing is performed. Cyclic ray tracing is to create the neutron path s so that the neutron path s started from the starting point repeats reflection in the analysis target region 30 and returns to the original starting point again.

（１） The neutron flux calculation step calculates the area average neutron flux of each detailed area i by solving the neutron transport equation based on the characteristic curve method for each neutron path s created by the ray tracing step as described above. Yes. Here, on the neutron path s, the relationship between the incident angle neutron flux incident on a certain detailed area i and the emission angle neutron flux emitted from the certain detailed area i is expressed by the following basic equation (1) of the characteristic curve method: Can be obtained from

(1)

  That is, the first term on the right side of the above equation (1) indicates how much the incident angle neutron flux incident on the detailed region i attenuates while traveling on the neutron path s, and the second term on the right side. Indicates how much neutrons generated in the detailed region i are added.

（２）
（３） Further, the line segment average angle neutron flux in the entire line segment of the neutron path s can be obtained by the following equation (2). By substituting the equation (1) into the equation (2), the equation (3) Can be obtained.

(2)
(3)

（４） The area average angle neutron flux in a certain detailed area i is averaged by weighting the line average angle neutron flux obtained on the line segment existing in the detailed area i by the volume of the calculated detailed area i. It can be obtained from the following equation (4).

(4)

（５） Incidentally, as shown in FIG. 5, in the characteristic curve method, the volume (area) of the detailed region i is the sum of the volume (area) of the segment Seg calculated on each neutron path s. It is expressed as

(5)

Here, ΔS ₀ (Ω) = (S _out −S _in ) is the length of the neutron path s in a certain detailed region i, and ΔA (Ω) is the path width of the neutron path s. For this reason, the volume (area) of each detailed area i has an error due to discretization. At this time, if the discretization parameters such as the path width of the neutron path s and the angular direction of the neutron path s are detailed, this discretization error can be reduced. However, the more detailed the discretization parameters are set, the more the number of calculation steps increases, leading to a longer calculation time. For this reason, the length of the neutron path s passing through the detailed region i is normalized by the volume correction factor C in order to reduce the error associated with the discretization.

（６）
（７）
（８） In the following formula (6), in a certain detailed region i, the true region volume (analytical solution) V _{ref, i} obtained analytically and the region volume (calculated solution) V _{cal, i} calculated by the characteristic curve method The volume correction factor C defined as the ratio of The volume correction factor C can be defined for each azimuth angle of the neutron path s as shown in the equation (7). However, for a minute detailed region i where the neutron path s does not pass in a certain angle direction, Since the correction factor C diverges, the angle average volume is adopted as shown in the equation (8).

(6)
(7)
(8)

（９） After normalization by normalizing the volume correction factor C obtained by substituting the expression (8) into the expression (6) on the length ΔS ₀ (Ω) of the neutron path s in a certain detailed region i, The length of the neutron path s is expressed by the following equation (9).

(9)

As described above, by standardizing the length ΔS ₀ (Ω) of the neutron path s in a certain detailed area i using the volume correction factor C, it is possible to reduce the error due to discretization. Here, the sensitivity of the path width of the neutron path s will be examined based on the experimental results shown in FIG. Note that the neutron path s having a path width of 0.02 cm is a reference solution, and the neutron path s has a path width of 0.05 cm, 0.1 cm, 0.15 cm, 0.2 cm, and 0.3 cm. , And 0.4 cm are considered.

  As shown in FIG. 6, from the left side, the path width of the neutron path s (Path Width (cm)), effective multiplication factor (k-effective), effective multiplication factor difference (Difference (pcm)), output of the fuel rod 10 The mean square difference (Pin power RMS Difference) and the calculation time (CPU time). The effective multiplication factor is the ratio of the number of neutrons generated per unit time in the core 1 to the number of neutrons consumed per unit time. When the effective multiplication factor is 1.0, the core 1 has an increase or decrease in neutrons. There is no critical state. In FIG. 6, in particular, attention is paid to the path width of the neutron path s, the effective multiplication factor, and the output distribution of the fuel rod 10, and the effective multiplication factor is used as an index for determining whether or not the analysis accuracy of the neutron flux is good.

  As can be seen from FIG. 6, when the path width of the neutron path s is 0.05 cm, a result almost equivalent to the reference solution is obtained. However, the wider the path width of the neutron path s, the more the reference solution of the effective multiplication factor. And the difference from the reference solution of the power distribution of the fuel rod 10 also increases. In other words, it can be seen that the analysis accuracy decreases as the path width of the neutron path s increases.

Next, referring to FIGS. 7 to 9, the volume when the path width of the neutron path s is 0.05 cm, 0.1 cm, 0.15 cm, 0.2 cm, 0.3 cm, and 0.4 cm. Consider the distribution of the correction factor C and the difference from the reference solution of the absorption response rate. Here, in particular, only cases where the path width of the neutron path s is 0.05 cm, 0.2 cm, and 0.4 cm are illustrated. In FIG. 7 to FIG. 9, the horizontal axis is obtained by numbering a plurality of divided detailed areas i and arranged in numerical order, and the right vertical axis is a volume correction factor C (V _ref / V _cal ). (V _cal / V _ref ), and the left vertical axis represents the difference from the reference solution of the absorption reaction rate.

Here, the volume correction factor C will be described. The volume V _ref of each detailed region i is an analytical solution obtained in advance, and the volume V _cal of each detailed region i is a segment calculated on the neutron path s. This is a calculation solution calculated from the sum of Seg volumes. When the volume correction factor C is 1, V _cal and V _ref are equal, and when the volume correction factor C is far from 1, V _cal is far away from V _ref .

As can be seen from FIGS. 7 to 9, as the path width of the neutron path s becomes wider, the reciprocal of the volume correction factor C diverges, that is, the reciprocal of the volume correction factor C deviates from 1. As shown in FIG. 10, when the neutron path s is drawn only in a part of the detailed area i, the volume V _cal of the calculated detailed area i is that of the detailed area i obtained analytically. The volume correction factor C becomes smaller than the volume V _ref and becomes larger than 1 (the reciprocal of the volume correction factor C is smaller than 1). On the other hand, as shown in FIG. 11, when the width of the detailed region i is smaller than the path width of the neutron path s, the calculated volume V _cal of the detailed region i is the volume of the detailed region i obtained analytically. It becomes larger than V _ref and the volume correction factor C becomes smaller than 1 (the reciprocal of the volume correction factor C becomes larger than 1).

  Here, it can be seen that the area where the volume correction factor C diverges is the cell area 31b in which the control rod 14 is inserted, and the area where the number of divisions of the detailed area i is large, in other words, each detailed area i becomes minute. It is an area. In the cell region 31b, it can be seen that the difference in the absorption reaction rate from the reference solution is also large. From the above, it can be seen that when the path width of the neutron path s is wide and the detailed area i is small, the volume correction factor C is far from 1, and an error occurs in the calculation accuracy of the area average angle neutron flux.

（１０） Therefore, in this embodiment, when the volume correction factor C is far from 1, the region average angle neutron flux is calculated in the detailed region i corresponding to the volume correction factor C by the basic equation (1) of the characteristic curve method. It is assumed that the incident angle neutron flux incident on the detailed area i is equivalent to the emission angle neutron flux emitted from the detailed area i. The region average angle neutron flux in the detailed region i is equivalent to the incident angle neutron flux and the emission angle neutron flux, and this relationship is expressed by the following equation (10). That is, when the volume correction factor C is far from 1, Equation (10) is used, and when the volume correction factor C is not far from 1, Equation (1) is used. This change of the expression is called filtering.

(10)

Thereby, in the detailed area i to which the formula (10) is applied, the area average angle neutron flux is calculated using the basic expression of the characteristic curve method, assuming that there is no neutron reaction in the detailed area i. Here, based on the experimental results shown in FIG. 12, the analysis results when the formula (10) is applied will be examined. Here, the path width of the neutron path s is fixed to 0.2 cm, the upper limit value and the lower limit value of the volume correction factor C are set, and the upper limit value and the lower limit value are used as the filtering factor. That is, when the volume correction factor C is larger than the upper limit value, it is determined that the volume V _cal is too small (is far away in the decreasing direction) with respect to the volume V _ref . On the other hand, when the volume correction factor C is smaller than the lower limit value, it is determined that the volume V _cal is too large with respect to the volume V _ref (away from the increasing direction).

As shown in FIG. 12, from the left side, the path width of the neutron path s (Path Width (cm)), the lower limit value of the filtering factor, the upper limit value of the filtering, the effective multiplication factor (k-effective), and the difference between the effective multiplication factors (Difference (pcm)). Note that the neutron path s having a path width of 0.02 cm is a reference solution. As can be seen from FIG. 12, when the lower limit value is fixed to 0 and the upper limit value is changed in the increasing direction, the effective multiplication factor is also changed in the increasing direction. On the other hand, if the lower limit value is changed in the increasing direction with the upper limit value fixed at 2.15, the effective multiplication factor hardly changes. From the above, with respect to the upper limit value where the volume correction factor C is greater than 1, the sensitivity of filtering is strong, and the analysis accuracy can be improved by performing filtering. This is because if the volume correction factor C is much larger than 1, it is considered that the detailed area i is very small as shown in FIG. 10, so that it can be regarded as ψ (S _out ) = ψ (S _in ). It is thought that it is from. In addition, for the lower limit value where the volume correction factor C is less than 1, the sensitivity of filtering is weak, and the improvement in analysis accuracy is negligible by performing filtering. This is considered to be because when the volume correction factor C is much smaller than 1, ψ (S _out ) ≈ψ (S _in ) in equation (1), which approximates equation (10). For this reason, at least filtering by the upper limit value can improve the analysis accuracy.

  Next, with reference to FIG. 13, the analysis result when the filtering factor is set optimally when the path width of the neutron path s is varied will be examined. Note that only the upper limit is applied as the filtering factor.

（１１） As shown in FIG. 13, from the left side, the path width of neutron path s (Path Width (cm)), upper limit value of filtering, virtual neutron leakage (Imaginary Leak), effective multiplication factor (k-effective), effective increase A difference in magnification (Difference (pcm)) and a mean square difference (Pin power RMS Difference) from the reference solution of the output distribution of the fuel rod 10 are obtained. Also in this case, the reference solution is the one having a neutron path s having a path width of 0.02 cm. In addition, when the left side and the right side of Equation (11) shown below do not match, it is defined as virtual neutron leakage. That is, due to the presence of virtual neutron leakage, the effective multiplication factor obtained by the aggregate calculation and the effective multiplication factor calculated from the uniform cross-sectional area of the assembly are inconsistent.

(11)

  As can be seen from FIG. 13, even if the path width of the neutron path s is widened, by filtering, the effective multiplication factor and the output distribution of the fuel rod 10 maintain the same value as the reference solution. It was confirmed that the accompanying error was reduced.

  According to the above configuration, even if the path width of the neutron path s is wide, the neutron flux can be calculated with high accuracy. In addition, since the calculation time can be shortened by the extent that the path width of the neutron path s can be widened, the core analysis can be performed at high speed. In this embodiment, the volume correction factor C is a filtering factor, but the length of the neutron path s in each detailed area i may be a filtering factor. That is, the length of the neutron path s is obtained analytically in advance, and if the length of the neutron path s obtained by calculation is far from the length of the neutron path s obtained analytically, filtering is performed. You may comprise.

  As described above, the core analysis program according to the present invention is useful for analyzing the neutron flux in the core, and is particularly suitable for shortening the calculation time.

1 is a structural diagram schematically showing a core that is a target of core design. FIG. 2 is a structural diagram schematically showing a fuel assembly that is an analysis target region of a core analysis program according to the present embodiment. It is explanatory drawing showing the analysis object area | region divided | segmented into the several detailed area | region. It is explanatory drawing showing the analysis object area | region where the neutron path was drawn. It is explanatory drawing showing the discretization image of the detailed area | region and angle neutron flux which are calculated in the characteristic curve method. It is a table | surface regarding the sensitivity of the path width of a neutron path. It is a graph showing the difference from the reference solution of the distribution of the volume correction factor and the absorption reaction rate when the path width of the neutron path is 0.05 cm. It is a graph showing the difference from the reference solution of the distribution of the volume correction factor and the absorption reaction rate when the path width of the neutron path is 0.2 cm. It is a graph showing the difference from the reference solution of the distribution of the volume correction factor and the absorption reaction rate when the path width of the neutron path is 0.4 cm. It is explanatory drawing showing the discretization error at the time of becoming " _Vcal < _Vref " in a certain detailed area _| region. It is explanatory drawing showing the discretization error at the time of becoming " _Vcal > _Vref " in a certain detailed area | region. It is a table | surface regarding the sensitivity of filtering in the state which fixed the path | pass width of the neutron path. It is the table | surface regarding the sensitivity of the filtering set optimally when the path width of a neutron path is made variable.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 Core 5 Fuel assembly 10 Fuel rod 11 Cladding tube 12 Gap between assemblies 13 Coolant 14 Control rod 20 Cell 30 Analysis object region 31a Cell region (fuel rod)
31b Cell area (control rod)
i Detailed region s Neutron path Seg Segment C Volume correction factor V _ref Detailed region volume (analysis solution)
Volume of V _cal detailed area (calculated solution)

Claims (4)

In the core analysis program that can be executed on hardware and can analyze the neutron flux in the core by solving the neutron transport equation based on the characteristic curve method,
A ray tracing step of creating a plurality of neutron paths on the analysis target region divided into a plurality of detailed regions, with a cell group configured by assembling a plurality of cells storing fuel rods as the analysis target region;
For each of the neutron paths created by the ray tracing step, a neutron flux calculation step for solving a neutron transport equation based on a characteristic curve method and calculating a neutron flux of each detailed region, and
In the neutron flux calculating step, as a ratio of a true region volume corresponding to each detailed region obtained analytically in advance and each region volume corresponding to each detailed region calculated based on the neutron path When each defined volume correction factor is larger than a predetermined upper limit value set in advance, the incident neutron flux incident on each detail region is assumed to be equal to the emitted neutron flux emitted from each detail region. A core analysis program characterized by calculating the neutron flux in the detailed region. In the neutron flux calculating step, the each volume correction factor, is smaller than a preset predetermined lower limit value, before Symbol and emission flux emitted from the incident neutron flux and the respective area of detail incident on each detail area The core analysis program according to claim 2 , wherein the neutron flux of each of the detailed regions is calculated on the assumption that they are equal .   In the core analysis program that can be executed on hardware and can analyze the neutron flux in the core by solving the neutron transport equation based on the characteristic curve method,
  A ray tracing step of creating a plurality of neutron paths on the analysis target region divided into a plurality of detailed regions, with a cell group configured by assembling a plurality of cells storing fuel rods as the analysis target region;
  For each of the neutron paths created by the ray tracing step, a neutron flux calculation step for solving a neutron transport equation based on a characteristic curve method and calculating a neutron flux of each detailed region, and
  In the neutron flux calculating step, as a ratio of a true region volume corresponding to each detailed region obtained analytically in advance and each region volume corresponding to each detailed region calculated based on the neutron path When each defined volume correction factor is smaller than a predetermined lower limit set in advance, it is assumed that the incident neutron flux incident on each detail area is equal to the emitted neutron flux emitted from each detail area, A core analysis program characterized by calculating the neutron flux in the detailed region.   In the neutron flux calculating step, when each volume correction factor is larger than a predetermined upper limit value set in advance, an incident neutron flux incident on each detailed region and an emitted neutron flux emitted from each detailed region are The core analysis program according to claim 3, wherein the neutron flux of each detailed region is calculated as equal.

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