JP6522368B2 - Nuclear constant calculation program and analyzer - Google Patents

Description

  The present invention relates to a nuclear constant calculation program and an analysis device for executing a neutron transport calculation to calculate a nuclear constant of a fuel assembly.

  Conventionally, as a core analysis program for analyzing the inside of a core, a nuclear constant calculation code (grid calculation code) for calculating a nuclear constant of a fuel assembly loaded in the core using a neutron transport equation is known (for example, Patent Document 1). In the nuclear constant calculation code, a square geometry having a cross section obtained by cutting the fuel assembly in a plane orthogonal to the axial direction is used as a two-dimensional analysis target area.

JP, 2012-168015, A

  The nuclear constant calculated by the nuclear constant calculation code of Patent Document 1 is used as an input value of the core calculation code, and a plurality of sets are prepared according to the calculation condition in consideration of the condition in the core. Then, in the core calculation code, nuclear characteristics in the core are analyzed based on the input nuclear constant.

By the way, in core analysis, analysis based on a wide range of calculation conditions is required in consideration of various conditions in the core such as severe accident. For example, as the condition in the core, it is assumed that the condition in which the inside of the core is filled with the moderator (cooling water) to the condition in which the moderator in the core is lost. In this case, the water density in the core is applied as one of the parameters of the calculation conditions, and it is necessary to consider the calculation conditions in which the water density is 0 g / cm ³ from the calculation conditions in which the water density is 1 g / cm ³ There is. However, under a wide range of calculation conditions, since the calculation conditions increase, the number of sets of nuclear constants calculated by the lattice calculation code increases, so the calculation load increases and the calculation cost increases.

  Therefore, an object of the present invention is to provide a nuclear constant calculation program and an analysis device capable of suppressing an increase in calculation load without deteriorating calculation accuracy even under a wide range of calculation conditions.

  A nuclear constant calculation program according to the present invention is a nuclear constant calculation program for causing a calculation unit of an analysis apparatus to execute neutron transport calculation to calculate a nuclear constant of a fuel assembly loaded in a core storing fuel rods. A parameter set to handle the fuel assembly system is set as a simple parameter, and a simple transport calculation step of executing a neutron transport calculation based on a large number of energy groups, and after execution of the simple transport calculation step A detailed transport calculation step of performing a neutron transport calculation, setting the parameter to a detailed parameter as compared to the simple transport calculation step and based on a small number of energy groups obtained by reducing the multiple energy groups; And the like.

  According to this configuration, the simplified transportation calculation step uses a large number of energy groups, while the parameters set for handling the fuel assembly system can be simplified, thereby shortening the calculation time. it can. Further, in the detailed transportation calculation step, while the parameters are made detailed, since the number of energy groups can be reduced, the calculation time can be shortened. And in the simplified transportation calculation step and the detailed transportation calculation step, since a large number of energy groups and detailed parameters are taken into consideration, the nuclear constant is made accurate while suppressing an increase in the calculation load even under a wide range of calculation conditions. It can be calculated well. In addition, as a system of a fuel assembly, there is an analysis object area which is a cell group configured by collecting a plurality of cells storing fuel rods. Further, as parameters set to handle the fuel assembly system, the number of divisions of the detailed area obtained by dividing the analysis target area into a plurality, the flight direction of neutrons on the analysis target area, anisotropic scattering of neutrons, etc. is there. For this reason, the simple parameter is to roughly handle the fuel assembly system, and specifically, the number of divisions of the detailed area is small, or the flight direction of the neutron set on the analysis target area (neutron path) In some cases, the anisotropic scattering of neutrons is simple (that is, isotropic scattering). On the other hand, the detailed parameter is to handle the fuel assembly system in detail. Specifically, the number of divisions of the detailed area is large, and the flight direction of the neutron set on the analysis target area (neutron path) Or neutrons are anisotropically scattered.

  In this case, the large number of energy groups is the number of groups obtained as a calculation result of resonance calculation for calculating an effective cross section serving as an input value of the neutron transport calculation, and the small number of energy groups are calculation of the resonance calculation. It is preferred that the resulting resonance peaks be a contraction of the plurality of energy groups so that they fall within the predetermined energy group.

  According to this configuration, a large number of energy groups can be reduced to a small number of energy groups while leaving the characteristics of the resonance peak without loss. For this reason, since the calculated nuclear constant can be made into consideration of the resonance peak, the nuclear constant can be accurately calculated.

  Preferably, the plurality of energy groups is 172 groups and the small number of energy groups is 22 groups.

  According to this configuration, it is possible to set the number of energy groups that can appropriately suppress an increase in calculation load while suppressing deterioration in calculation accuracy of the calculated nuclear constant.

  An analysis apparatus of the present invention is characterized in that it can execute the above-described nuclear constant calculation program.

  According to this configuration, even under a wide range of calculation conditions, it is possible to suppress an increase in calculation load without deteriorating the calculation accuracy of the nuclear constant.

FIG. 1 is a schematic configuration diagram schematically showing an analysis device according to the present embodiment. FIG. 2 is an explanatory view schematically showing a core to be analyzed by the analysis apparatus according to the present embodiment. FIG. 3 is a cross-sectional view of the fuel assembly to be analyzed, which is taken along a plane orthogonal to the axial direction. FIG. 4 is an explanatory diagram of a lattice calculation code stored in the analysis device. FIG. 5 is an explanatory view showing an analysis target area divided into a plurality of detailed areas. FIG. 6 is an explanatory view showing an analysis target area in which a neutron path is drawn. FIG. 7 is a flowchart for calculating the neutron flux by the transport calculation program. FIG. 8 is a graph showing the effective cross sections of the 172 group obtained by resonance calculation. FIG. 9 is an enlarged graph of a part surrounded by a dotted line in FIG. FIG. 10 is a diagram comparing the calculation accuracy of the lattice calculation code of the present embodiment with the calculation accuracy of the conventional lattice calculation code. FIG. 11 is a diagram comparing the calculation speed of the grid calculation code of this embodiment with the calculation speed of the conventional grid calculation code.

  Hereinafter, embodiments according to the present invention will be described in detail based on the drawings. The present invention is not limited by this embodiment. In addition, constituent elements in the following embodiments include those that can be easily replaced by persons skilled in the art or those that are substantially the same. Furthermore, the components described below can be combined as appropriate, and when there are a plurality of embodiments, it is also possible to combine each embodiment.

[Embodiment]
The nuclear constant calculation program according to the present embodiment is incorporated into a core analysis program for evaluating fuel assemblies in a core. The core analysis program is stored in the analysis device, and the core analysis program is executed in the analysis device to calculate the neutron flux in the core and predict the distribution and behavior of neutrons that mediate nuclear reactions in the core, Evaluate nuclear characteristics in the core. Then, core design is performed based on the analysis result obtained by the core analysis program. The core design is performed to replace the fuel loaded in the core in consideration of safety, combustion efficiency, fuel arrangement, and the like. First, an analysis apparatus will be described with reference to FIG.

  FIG. 1 is a schematic block diagram schematically showing an analysis apparatus. As shown in FIG. 1, the analysis apparatus 40 includes an operation unit 41 capable of executing various programs and performing calculations, a storage unit 42 for storing various programs and data, and an input unit 43 configured by an input device such as a keyboard. And an output unit 44 configured of an output device such as a monitor. The analysis device 40 may be configured as a single device, or may be configured as a plurality of devices combining an arithmetic device, a data server, and the like, and is not particularly limited.

  The core analysis program P is stored in the storage unit 42 as various programs, and as a data, a cross-sectional area library D1 in which effective cross-sectional areas are summarized and nuclear constant data D2 in which nuclear constants forming a plurality of sets are integrated. And are stored. Further, the core analysis program P includes a lattice calculation code (nuclear constant calculation program) C1 for calculating a nuclear constant, and a core calculation code C2 for analyzing nuclear characteristics in the core based on the nuclear constant. There is.

  Here, the core 5 to be analyzed by the core analysis program P will be described with reference to FIGS. 2 and 3. FIG. 2 is an explanatory view schematically showing a core to be analyzed by the analysis apparatus according to the present embodiment. FIG. 3 is a view in which a fuel assembly to be an analysis target area is cut in a plane orthogonal to the axial direction FIG. As shown in FIG. 2, in the nuclear reactor, a core 5 to be subjected to core design is stored. The core 5 is composed of a plurality of fuel assemblies 6. The fuel replacement is performed in units of six fuel assemblies.

  As shown in FIG. 3, each fuel assembly 6 is composed of a plurality of fuel rods 10, a plurality of cladding tubes 11 covering the respective fuel rods 10, and a grid (not shown) for bundling the plurality of cladding tubes 11, The inside of the fuel assembly 6 is filled with a moderator (coolant) 13, and a plurality of control rods 14 and in-core nuclear instrumentation 15 can be inserted.

  The fuel assembly 6 is formed in a rectangular cross section, and is configured of, for example, 17 × 17 cells 20. The control rods 14 are inserted into the 24 cells 20 among the 17 × 17 cells 20, and the in-core nuclear instrumentation 15 is inserted into the cell 20 at the center of the assembly. At this time, the cell 20 in which the control rod 14 is inserted is referred to as a control rod guide pipe, and the cell 20 in which the in-core nuclear instrumentation 15 is inserted is referred to as an instrumentation guide pipe. Further, the fuel rods 10 are inserted into the other cells 20 respectively. When the fuel assembly 6 is used in a boiling water reactor (BWR), the outside of the fuel assembly 6 is covered with a channel box. On the other hand, when the fuel assembly 6 is used in a pressurized water reactor (PWR), the fuel assembly 6 is open at the outside. An inter-assembly gap 12 exists outside the channel box in the case of BWR and outside the fuel assembly 6 in the case of PWR.

  Next, the lattice calculation code C1 and the core calculation code C2 of the core analysis program P will be described.

  The lattice calculation code C1 has a rectangular geometric shape which is a cross section obtained by cutting the fuel assembly 6 in a plane orthogonal to the axial direction as a two-dimensional analysis target area 30 (see FIG. 3). It is a code that can calculate the number. The nuclear constant is input data used for the core calculation code C2, and the nuclear constant includes the diffusion coefficient, the absorption cross section, the removal cross section, the generation cross section, and the like. That is, by performing nuclear constant calculation using the lattice calculation code C1, a nuclear constant which is input data for core calculation is generated.

  The core calculation code C2 performs core calculation by respectively setting the calculated nuclear constants to a fuel node (not shown) in which the fuel assembly 6 is divided into a plurality of pieces in the axial direction to form a small volume of rectangular solid shape. . The plurality of fuel nodes represent the core, and the core calculation code C2 is a code capable of evaluating core characteristics in the core such as critical boron concentration, power distribution, reactivity coefficient, etc. by performing core calculation. ing.

  The analysis device 40 causes the operation unit 41 to execute the core analysis program P stored in the storage unit 42 based on the input parameters input from the input unit 43. Then, the analysis device 40 calculates the nuclear constant in the analysis target region 30 of the fuel assembly 6 using the lattice calculation code C1, and sets the calculated nuclear constant to each fuel node using the core calculation code C2. The core characteristics of the core 5 are evaluated by performing core calculation. Then, the analysis device 40 outputs the analysis result by the core analysis program P to the output unit 44.

  Next, the lattice calculation code C1 will be specifically described with reference to FIG. FIG. 4 is an explanatory diagram of a lattice calculation code stored in the analysis device. The lattice calculation code C1 of the present embodiment calculates neutron flux in the fuel assembly 6, performs combustion calculation, and performs nuclear constant calculation.

  The lattice calculation code C1 includes a resonance calculation program 51, a transport calculation program 52, a combustion calculation program 53, and a nuclear constant calculation program 54. The lattice calculation code C1 is based on specification data on the fuel assembly 6 input to the analysis device 40 and the effective cross-sectional area acquired from the cross-sectional area library D1 stored in the storage unit 42 of the analysis device 40. Are doing various calculations. The specification data includes, for example, the radius of the fuel rod, the gap between the assemblies, the fuel composition, the fuel temperature, the moderator temperature and the like.

  FIG. 5 is an explanatory view showing an analysis target area divided into a plurality of detailed areas. As shown in FIG. 5, the analysis target area 30 to be analyzed by the lattice calculation code C1 is an arbitrary calculation system, and is constituted by a plurality of cell areas 31a and 31b corresponding to each cell 20. The cell regions 31a and 31b include, for example, a cell region 31a in which the fuel rod 10 is inserted and a cell region 31b in which the control rod 14 is inserted. The cell areas 31a and 31b are divided into a plurality of detailed areas. A part of the plurality of detailed regions is a resonance region where a resonance phenomenon occurs.

  The resonance calculation program 51 is executed to obtain the effective cross-sectional area of each of the detail areas in consideration of the resonance phenomenon. Here, the resonance phenomenon is a phenomenon in which the cross-sectional area dramatically increases when the energy of neutrons reaches a predetermined energy, and appears as, for example, a resonance peak E shown in FIG. In this resonance calculation program 51, the energy of neutrons is divided into a plurality of energy groups, for example, 172 groups, and an effective cross-sectional area, which is the average cross-sectional area of each divided energy group, is determined. That is, in the resonance calculation program 51, the effective cross-sectional area of the multiple group (172 group) is calculated. Then, the calculated effective cross-sectional area is stored in the cross-sectional area library D1.

  The transport calculation program 52 executes a simple neutron transport calculation and a detailed neutron transport calculation using the calculated effective cross section. Here, FIG. 6 is an explanatory view showing an analysis target area in which a neutron path is drawn. As shown in FIG. 6, the transport calculation program 52 creates a plurality of neutron flight paths s on the analysis target region 30 divided into a plurality of detailed regions. Then, the neutron flux of each detailed region i is calculated for each of the created neutron flight paths s. The simple neutron transport calculation and the detailed neutron transport calculation will be described later.

  The combustion calculation program 53 executes a combustion calculation that tracks the generation and extinction of the nuclide in the core 5. The combustion calculation program 53 evaluates the time change of the atomic number density of each nuclide by solving the combustion equation, and gives an input condition of multi-group neutron transport calculation at each burnup point. Thereby, the fuel calculation program 53 tracks the combustion state (time change of the combustion) by alternately performing the combustion calculation and the transport calculation for each predetermined sampling cycle.

  The nuclear constant calculation program 54 uses the neutron fluxes of multiple groups in the fuel assembly 6 obtained by the transport calculation program 52 as weights, and reduces, homogenizes and homogenizes the effective cross sections of the multiple groups in the fuel assembly 6 Calculate the calculated nuclear constant.

  Subsequently, the transportation calculation program 52 of the present embodiment will be specifically described with reference to FIG. 7. FIG. 7 is a flowchart for calculating the neutron flux by the transport calculation program. As shown in FIG. 7, the computing unit 41 of the analysis device 40 executes the transportation calculation program 52 to execute steps including a simplified transportation calculation step S2 and a detailed transportation calculation step S4.

  First, the calculation unit 41 of the analysis device 40 executes the resonance calculation program 51 to calculate the effective cross-sectional area of the multiple group (172 group), and the calculated effective cross-sectional area is stored in the cross-sectional area library D1 of the storage unit 42. Save (step S1). Thereafter, the calculation unit 41 of the analysis apparatus 40 executes the transport calculation program 52 to execute neutron transport calculation based on the energy group of the multiple group (172 group) (step S2: simplified transport calculation step).

  In the simplified transportation calculation step S2, the parameters set to handle the system (calculation system) of the fuel assembly 6 are set as simple parameters compared to the detailed transportation calculation step S4. Here, the parameters include the division number of the detailed area i obtained by dividing the analysis target area 30 into a plurality of pieces, the flight direction of neutrons (neutron flight path s) on the analysis target area 30, and anisotropic scattering of neutrons. . For this reason, in the simplified transportation calculation step S2, the system of the fuel assembly 6 is roughly handled by setting simple parameters. Specifically, in the simplified transport calculation step S2, the number of divisions of the detailed region i is small (in other words, the area of the detailed region i is large), or the flight direction of neutrons (neutrons) set on the analysis target region 30 The flight path s) is small, or anisotropic scattering of neutrons is simple (that is, isotropic scattering).

  In the simplified transport calculation step S2, the neutron cross-section collision probability method (CCCP method: Current Coupling Collision Probability Method) is used based on the effective cross-sections of the 172 groups which are obtained as the calculation result of the resonance calculation. , Neutron transport calculations are performed. In the simplified transport calculation step S2, it is assumed that the neutron flows among the cells 20 are isotropically equally distributed (DP0).

  As described above, in the simplified transportation calculation step S2, the system of the fuel assembly 6 is roughly handled, and the neutron flux of each group is calculated using the neutron flow coupled collision probability method based on the effective cross section of the 172 group.

  After execution of the simplified transport calculation step S2, the computing unit 41 of the analysis device 40 reduces the effective cross-sections of the 172 energy groups using the neutron flux of the 172 group calculated in the simplified transport calculation step S2 as a weighting. , And 22 create an effective cross-sectional area of the energy group (step S3).

  FIG. 8 is a graph showing the effective cross sections of the 172 group obtained by resonance calculation, and FIG. 9 is a graph in which a part surrounded by a dotted line in FIG. 8 is enlarged. In the graphs of FIGS. 8 and 9, the horizontal axis represents energy of neutrons, and the vertical axis represents an absorption cross section which is one of the effective cross sections. Here, in the energy group, the right side of FIG. 8 where the neutron energy is large is the first group, and the energy groups of the neutrons are small as the left side of FIG. It becomes a group. As shown in FIGS. 8 and 9, a resonance peak E appears in the effective cross section obtained by the resonance calculation.

  In step S3, when the effective cross section of the 172 group is reduced to the effective cross section of the 22 group, the energy group of the 172 group is selected so that the resonance peak E appearing in the resonance calculation falls within the predetermined energy group. It is contracted. That is, as shown in FIG. 9, among the 172 groups, a predetermined energy group is an energy group corresponding to the top of the resonance peak E. Then, among the 22 groups, a predetermined energy group is an energy group centered on the energy group corresponding to the top of the resonance peak E. In other words, in the energy group of the 22 groups to be contracted, the number of divisions at the resonance peak E is larger than the number of divisions at other than the resonance peak E. Therefore, the energy group of the 22 groups to be contracted has high resolution at the resonance peak E, while the resolution at other than the resonance peak E is low. Therefore, the effective cross-sectional area of the group 22 reduced in step S3 is the one in which the characteristic (characteristic) of the resonance peak E remains.

  After the execution of step S3, the operation unit 41 of the analysis device 40 executes neutron transport calculation based on the reduced energy group of the minority group (22 group) (step S4: detailed transport calculation step).

  In the detailed transportation calculation step S4, the parameters set to handle the system of the fuel assembly 6 are set as detailed parameters in comparison with the simplified transportation calculation step S2. That is, in the detailed transportation calculation step S4, the system of the fuel assembly 6 is handled in detail by setting detailed parameters. Specifically, in the detailed transport calculation step S4, the number of divisions of the detailed region i is large (in other words, the area of the detailed region i is small) or the flight direction of neutrons (neutrons) set on the analysis target region 30 There are many flight paths s), or neutrons are anisotropically scattered.

  Further, in the detailed transport calculation step S4, neutron transport calculation is performed using the characteristic curve method (MOC: Method of Characteristics) based on the reduced effective cross section of the reduced 22 groups.

  Thus, in the detailed transport calculation step S4, the system of the fuel assembly 6 is handled in detail, and the neutron flux of each group is calculated using the characteristic curve method based on the reduced effective cross section of the 22 group. .

  Then, after executing the detailed transport calculation step S4, the operation unit 41 of the analysis device 40 expands the neutron flux of group 22 into 172 groups by using a predetermined calculation formula (step S5). The neutron flux is calculated (step S6), and the calculation based on the transport calculation program 52 is ended.

  The calculation unit 41 of the analysis device 40 executes the nuclear constant calculation program 54 to calculate the nuclear constants of the respective groups based on the neutron flux of the 172 group calculated by the transport calculation program 52.

  Next, with reference to FIGS. 10 and 11, the calculation accuracy and calculation speed for neutron transport calculation using the lattice calculation code of the present embodiment, and the calculation accuracy and calculation for neutron transport calculation using the conventional lattice calculation code Compare with speed. FIG. 10 is a diagram comparing the calculation accuracy of the grid calculation code of the present embodiment with the calculation accuracy of the conventional grid calculation code, and FIG. 11 is the calculation speed of the grid calculation code of the present embodiment and the conventional It is the figure which compared with the calculation speed of the lattice calculation code. Here, the conventional lattice calculation code is to perform neutron transport calculation using the characteristic curve method (MOC) based on the effective cross sections of multiple groups which are 172 groups obtained as a calculation result of resonance calculation. .

In FIG. 10, "Fuel Type" is a type including fuel and cell types contained in the fuel rod 10. As fuel types, uranium oxide fuel (UO ₂ ), gadolinia (Gd ₂ O ₃ ), uranium oxide fuel (Gd) And mixed oxide fuel (MOX), erbia-containing uranium oxide fuel (Er), etc., and cell types include “14 × 14”, “15 × 15”, “17 × 17” and the like. “Avg” is an average of the difference between the reactivity of the fuel assembly 6 and the reference solution, and “Stdev” is a standard deviation of the difference of the reactivity of the fuel assembly 6 from the reference solution. The reference solution is the reactivity of the fuel assembly obtained by neutron transport calculation based on the continuous energy Monte Carlo method. Comparing the reactivity of the fuel assembly 6 calculated using the lattice calculation code of the present embodiment with the reactivity of the fuel assembly 6 calculated using the conventional lattice calculation code, the reactivity is almost equivalent. It has been confirmed that the calculation accuracy is maintained.

  In FIG. 11, the right-hand column is the calculation speed of the present embodiment divided by the conventional calculation speed (the calculation speed of the present embodiment / the conventional calculation speed), and the calculation speed of the present embodiment is the conventional It is about 3 times faster than the calculation speed of.

  As described above, according to the present embodiment, while the large number of energy groups are used in the simplified transportation calculation step S2, the system of the fuel assembly 6 is roughly handled, so that the calculation time can be shortened. In addition, while the system of the fuel assembly 6 is handled in detail in the detailed transportation calculation step S4, the number of energy groups can be reduced, so that the calculation time can be shortened. Then, in the simplified transportation calculation step S2, while using a large number of energy groups, the system of the fuel assembly 6 is roughly handled, and in the detailed transportation calculation step S4, the number of energy groups is small. Since the system of (1) is handled in detail, the calculation accuracy of the simplified transport calculation step S2 is compensated, and even under a wide range of calculation conditions, the nuclear constant can be accurately calculated while suppressing the increase of the calculation load.

  Moreover, according to the present embodiment, it is possible to reduce a large number of energy groups into a small number of energy groups while leaving the characteristic of the resonance peak E without losing it. For this reason, since the calculated nuclear constant can be set in consideration of the resonance peak E, the nuclear constant can be accurately calculated.

  Further, according to the present embodiment, since a large number of energy groups can be set to 172 groups and a small number of energy groups can be set to 22 groups, as shown in FIG. 10 and FIG. The increase of the calculation load can be suitably suppressed while suppressing the deterioration of.

  In addition, this embodiment may be applied to a boiling water reactor, may be applied to a pressurized water reactor, and is not particularly limited.

  In the present embodiment, the CCCP method is used in the simplified transportation calculation step S2, and the MOC method is used in the detailed transportation calculation step S4, but the present invention is not limited to this configuration.

Reference Signs List 5 core 6 fuel assembly 10 fuel rod 11 cladding tube 12 inter-assembly gap 13 moderator 14 control rod 15 in-core nuclear instrumentation 20 cell 30 analysis target area 31a, 31b cell area 40 analysis device 41 calculation unit 42 storage unit 43 input Part 44 Output part 51 Resonance calculation program 52 Transport calculation program 53 Combustion calculation program 54 Nuclear constant calculation program P Core analysis program D1 Cross section library D2 Nuclear constant data C1 Lattice calculation code C2 Core calculation code s Neutron flight path i Detailed area E Resonance peak

Claims (4)

In a nuclear constant calculation program for causing a calculation unit of an analysis apparatus to execute neutron transport calculation to calculate a nuclear constant of a fuel assembly loaded in a core storing fuel rods,
Setting a parameter set to handle the system of the fuel assembly to a simple parameter, and performing a neutron transport calculation based on multiple energy groups;
After the execution of the simplified transport calculation step, the neutron transport calculation is performed based on a small number of energy groups in which the parameter is set to a detailed parameter compared to the simplified transport calculation step and the large number of energy groups are contracted. And a detailed transport calculation step to be executed. In resonance calculation for calculating an effective cross section serving as an input value of the neutron transport calculation, energy of neutrons is divided into the plurality of energy groups, and the effective cross sections of the divided energy groups are calculated.
The multiple energy groups are the number of groups divided in the resonance calculation,
The small number of energy groups are characterized in that the large number of energy groups are contracted so that resonance peaks obtained as a calculation result of the resonance calculation fall within a predetermined energy group. Nuclear constant calculation program described in. The plurality of energy groups are 172 groups,
The nuclear constant calculation program according to claim 1 or 2, wherein the small number of energy groups is 22 groups.   An analysis apparatus characterized by being capable of executing the nuclear constant calculation program according to any one of claims 1 to 3.

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