KR101129953B1 - Storage medium storing core analysis program - Google Patents

Abstract

An object of the present invention is to provide a core analysis program which can easily calculate the nuclear constant according to the operating conditions of the core and can accurately evaluate the nuclear characteristics in the core using the calculated nuclear constant.
In a core analysis program executable in hardware for evaluating nuclear characteristics in a core, a calculation point is set for each of a plurality of parameters depending on the nuclear reaction in the fuel assembly, and the nuclear constant is calculated at the calculation points set for each parameter. Nuclear constant calculation step S101 and fitting equation derivation step S102 for deriving the fitting constant of the nuclear constant corresponding to the displacement of each parameter from the calculated nuclear constant at the calculated point to have robustness, and the operating conditions of the core The core calculation step S103 which derives the nuclear constant corresponding to the some parameter changed according to this from a fitting formula, and performs core calculation based on the derived nuclear constant is provided.

Description

Storage medium storing the core analysis program {STORAGE MEDIUM STORING CORE ANALYSIS PROGRAM}

The present invention relates to a core analysis program for evaluating nuclear characteristics in a core.

Conventionally, as a method for evaluating nuclear characteristics in a core, a core performance calculation method of a reactor for calculating neutron flux distribution in a core and fuel rod output distribution in an assembly is known (see Patent Document 1, for example). In this core performance calculation method, a neutron is divided by dividing the reactor core into a plurality of small volumes (fuel nodes) and using a nuclear constant for the fuel nodes given by a single aggregate combustion calculation (nuclear constant calculation). The core performance of the reactor is evaluated by knowing the diffusion equation or the transport equation (by performing the core calculation). At this time, the nuclear constant given by the single aggregate combustion calculation is changed by the spectral interference effect, which is the effect of neutrons leaking from the adjacent aggregates. For this reason, in this core performance calculation method, the nuclear constant given by the single aggregate combustion calculation is correct | amended in consideration of the spectral interference effect.

[Patent Document 1] Japanese Patent Application Laid-open No. Hei 8-75891

In general, when evaluating the nuclear characteristics in the core, as in the conventional core performance calculation method, the nuclear constant of the aggregate is calculated to calculate the nuclear constant, and the calculated nuclear constant is set in the fuel node to perform core calculation. Nuclear characteristics within the core are being assessed. Here, when the nuclear characteristics in the core are evaluated with high accuracy, it is preferable to set an optimal nuclear constant corresponding to each of the plurality of fuel nodes. In other words, the nuclear reaction in each fuel node is changed due to different operating conditions of the core, that is, parameters such as the fuel temperature of the aggregate and the temperature of the coolant. For this reason, if the parameters, such as fuel temperature and coolant temperature, are appropriately set for each fuel node, and nuclear constant calculation of an assembly is performed, the optimal nuclear constant corresponding to each fuel node can be calculated. In this case, however, the number of calculations for the calculation of the nuclear constant must be performed by the number of fuel nodes, and in this case, the calculation time is enormous. Therefore, the calculation of the nuclear constant for each fuel node in accordance with the operating conditions of the core consumes a great amount of calculation time, and there is a fear that the calculation cost increases.

Therefore, an object of the present invention is to provide a core analysis program which can easily calculate the nuclear constant according to the operating conditions of the core and can accurately evaluate the nuclear characteristics in the core using the calculated nuclear constant.

The core analysis program of the present invention performs nuclear constant calculation for calculating the nuclear constant of the fuel assembly loaded on the core, and performs core calculation based on the nuclear constant calculated by the nuclear constant calculation. In a core analysis program executable on hardware for evaluating characteristics, a nucleus constant for calculating a nucleus constant at a calculation point set for each parameter by setting a calculation point for each of a plurality of parameters depending on the nuclear reaction in the fuel assembly. A fitting equation derivation step for deriving the fitting constant of the nuclear constant corresponding to the displacement of each parameter from the calculated nucleus constant at the calculated calculation point and the calculated point to have robustness; A furnace for deriving the nuclear constant corresponding to the parameter of the equation from the fitting equation and performing core calculation based on the derived nuclear constant It is characterized by having a seam calculation step.

According to this configuration, the fitting formula of the nuclear constant can be derived based on the nuclear constant calculated at the calculation point set for each parameter. For this reason, the nuclear constant according to the operating conditions of a core can be computed simply by inputting the value according to the operating conditions of a core to each parameter to the derived fitting formula. Thereby, since core calculation can be performed based on the nuclear constant according to operating conditions of a core, the nuclear characteristic in a core can be analyzed with high precision. In addition, since the fitting formula of the nuclear constant has robustness, the derived fitting formula has the concept of quality engineering developed from the experimental design method, and the nuclear constant derived by the fitting formula is highly reliable and has high precision. It can be set as a value of degree. In addition, robustness means the property which is hard to be affected even if there exists a fluctuation error, such as noise in each parameter, and has stability.

In this case, in the nuclear constant calculation step, it is preferable that the calculation point set for each parameter is selected based on the orthogonal table, and the nuclear constant calculation is performed based on the calculation point of each selected parameter.

According to this configuration, since the calculation point can be selected based on the orthogonal table, the nuclear constant at each calculation point can be derived while reducing the number of calculations of the nuclear constant. For example, in the case where the calculation points of the three parameters are each 6 points, it is necessary to perform 216 calculations (6x6x6 types) when the numbers are calculated in all cases. However, in the case of using the L36 orthogonal table, the calculation results are almost the same as in the case of the 216 calculations by performing 36 calculations.

In this case, the fitting derivation step sets, for each parameter, a term indicating the behavior of the nuclear constant when one parameter is displaced and another parameter is fixed based on the nuclear constant calculated by the nuclear constant calculation step. It is provided with the fitting formula making step which combines the term setting step to combine, the term which shows the behavior of the nuclear constant set for each parameter, and the robustness evaluation step which evaluates the robustness of the created fitting formula. desirable.

According to this configuration, the fitting equation used in the core calculation step can be made to have robustness by evaluating the created fitting equation. In the robustness evaluation step, when the created fitting formula does not have robustness, it is preferable to create a new fitting formula in the fitting formula preparation step until it has robustness.

In this case, in the robustness evaluation step, it is preferable that the fitting-type robustness is evaluated using Akaike Information Criterion.

According to this configuration, since the fitting equation can be evaluated using the Akaike information quantity criterion, the resulting fitting equation can be made highly reliable.

In this case, the plurality of parameters include the fuel temperature of the fuel assembly, the temperature of the coolant introduced into the core, the density of the coolant introduced into the core, and among these, it is preferable to use two parameters including at least the density of the coolant. .

According to this configuration, the fitting equation can be derived from three combinations of fuel temperature, coolant density and coolant temperature, fuel temperature and coolant density, coolant density and coolant temperature. As a result, the dependence on the nuclear reaction is high, and an optimal parameter can be selected in deriving a fitting equation.

According to the core analysis program of the present invention, the nuclear constant according to the operating conditions of the core can be simply calculated, and the calculated nuclear constant can be used to accurately evaluate the nuclear characteristics in the core.

BRIEF DESCRIPTION OF THE DRAWINGS The structural diagram which showed typically the reactor in which the core used for core design was installed.
FIG. 2 is a cross-sectional view taken along the line AA ′ of the core in FIG. 1. FIG.
3 is a cross-sectional view of the fuel assembly.
4 is an explanatory diagram of a core divided into a plurality of fuel nodes;
5 is a flowchart for evaluating nuclear properties in a core using a core analysis program.
Fig. 6 is a table showing calculation points set for each parameter.
7 is a table showing a calculation point when one parameter is displaced and another parameter is fixed.
8 is a table in the form of terms showing the behavior of nuclear constants set for each parameter.
9 is a flowchart for deriving a fitting equation for nuclear constants.
Fig. 10 is a table showing the reactivity coefficients obtained using the core analysis program of the present example and their maximum errors.

EMBODIMENT OF THE INVENTION Hereinafter, the core analysis program which concerns on this invention is demonstrated with reference to attached drawing. In addition, this invention is not limited by the following example. In addition, the component in a following example includes the thing which can be replaced by those skilled in the art, and what is easy or substantially the same.

[Example]

The core analysis program according to the present embodiment is a program for analyzing the distribution of neutron flux in the core, and predicts and evaluates the distribution and behavior of neutrons that mediate the nuclear reaction in the core. And core design is performed based on the analysis result obtained by this core analysis program. In addition, the core design is performed to replace the fuel loaded in the core in consideration of safety, combustion efficiency, fuel arrangement, and the like.

1 is a structural diagram which shows typically the reactor in which the core used for core design was installed. FIG. 2 is a cross-sectional view taken along the line AA ′ of the core in FIG. 1. FIG. 3 is a cross-sectional view of the fuel assembly. 4 is an explanatory diagram of a core divided into a plurality of fuel nodes.

As shown in FIG. 1, the reactor 5 stores a core 5 to be subjected to core design. The core 5 is composed of a plurality of fuel assemblies 6, and the plurality of fuel assemblies 6 are arranged in a geometric shape so as to maintain symmetry of 90 degrees as shown in FIG. 2. In addition, the fuel is exchanged in units of the fuel assembly 6.

As shown in FIG. 3, each fuel assembly 6 includes a plurality of fuel rods 10, a plurality of covering tubes 11 covering each of the fuel rods 10, and an unillustrated grid that bundles the plurality of covering tubes 11. The inside of the fuel assembly 6 is filled with the coolant 13, and the control rod 14 is comprised so that insertion is possible.

The fuel assembly 6 is formed in a rectangular cross section, and is composed of, for example, a cell 20 of 17 × 17. The control rods 14 are inserted into the twenty-four cells 20 of the cells 20 of 17x17, respectively, and an in-house nuclear instrument is inserted into the cells 20 at the center of the assembly. At this time, the cell 20 into which the control rod 14 is inserted is referred to as the control rod guide tube and the cell 20 into which the furnace nuclear instrument is inserted. In addition, the fuel rods 10 are inserted into the other cells 20, respectively. In addition, when the fuel assembly 6 is used for the boiling water reactor BWR, the fuel assembly 6 is covered with a channel box on the outside thereof. On the other hand, when the fuel assembly 6 is used in the pressurized water reactor PWR, the fuel assembly 6 is open outside. In the case of BWR, the inter-assembly gap 12 exists outside the channel box and outside the fuel assembly 6 in the case of PWR.

Next, the core analysis program will be described. This core analysis program is a program executable in hardware, and calculates nuclear characteristics in the core 5 based on the nuclear constant calculation code for calculating the nuclear constant of the fuel assembly loaded on the core and the calculated nuclear constant. It has a core calculation code.

As shown in FIG. 3, the nuclear constant calculation code uses a rectangular geometric shape that forms a cross section obtained by cutting the fuel assembly 6 into a plane orthogonal to the longitudinal direction as a two-dimensional analysis target region 30. It is a code that can calculate the nuclear constant in the analysis target area 30. In addition, the nuclear constant is input data used for core calculation, and the nuclear constant includes a diffusion coefficient, an absorption cross section, a removal cross section, and a generated cross section. That is, the nuclear constant is calculated using the nuclear constant calculation code to generate the nuclear constant, which is input data for core calculation.

The core calculation code evaluates the core characteristics in the core such as the critical boron concentration, the power distribution, the reactivity coefficient by performing core calculation by setting the calculated nuclear constants in the plurality of fuel nodes 35 shown in FIG. 4, respectively. Possible code is: The fuel node 35 is a small volume of a rectangular parallelepiped shape which divided the core 5 into a plurality.

When the above core analysis program is executed on hardware, the hardware calculates the nuclear constant in the analysis target region 30 of the fuel assembly 6 using the nuclear constant calculation code, and uses the core calculation code. The nucleus characteristics of the core 5 are evaluated by setting the calculated nuclear constant at each fuel node 35 to perform core calculation.

Here, the nuclear constant calculated by the nuclear constant calculation code is preferably set for each fuel node 35 in accordance with the operating conditions of the core 5. That is, since each fuel node 35 differs in operating conditions of the core 5 according to the place, the optimum nuclear constant is set according to each fuel node 35, so that the evaluation of the nuclear characteristics of the core 5 can be precisely performed. It becomes possible to do it on the road.

At this time, if the nuclear constants of the plurality of fuel nodes 35 are to be calculated using the nuclear constant calculation code, the nuclear constants should be calculated by the number of fuel nodes 35, so that the calculation time in the core analysis program To be massive. Therefore, in the core analysis program of the present embodiment, by using the concept of quality engineering that has developed the experimental design method, and using the fitting constant of the nuclear constant corresponding to the operating conditions of the core 5, it corresponds to each fuel node 35. It is possible to derive the nuclear constant. Hereinafter, a series of control steps for evaluating the nuclear characteristics in the core 5 using the core analysis program of the present embodiment will be described in detail. In the present embodiment, a case where the neutrons flying in the analysis target region 30 are divided into two groups, a high speed neutron group and a thermal neutron group, will be described. However, the present invention is not limited to this configuration. You may apply about division.

First, before describing a series of control steps in the core analysis program of the present embodiment, a plurality of parameters serving as operating conditions of the core 5 will be described. This parameter changes depending on the operating conditions of the core 5 and is also dependent on the nuclear reaction. As this parameter, it is as showing in Formula (1).

As shown in Equation (1), the macro-sectional area Σ _G created for the energy group G includes the geometric shape Geom, the fuel composition N _f0 , the combustion degree Bu, It can be seen that it depends on the parameters of the boron concentration (C _B ), the fuel temperature (T _f ), the coolant density (ρ _m ), and the coolant temperature (T _m ). Among these parameters, in deriving the fitting formula of the nuclear constant, the parameters that are easy to derive (i.e., easy to use as polynomials) are fuel temperature (T _f ), coolant density (ρ _m ), and coolant temperature (T _m). )to be. For this reason, in this embodiment, as shown in Formula (2), the fitting formula of the nuclear constant is set using the fuel temperature T _f , the coolant density ρ _m , and the coolant temperature T _m . To derive.

Subsequently, with reference to FIG. 5, a series of control steps for evaluating the nuclear characteristics in the core 5 using the core analysis program of the present embodiment will be described in detail. 5 is a flowchart for evaluating nuclear properties in a core using a core analysis program. As shown in Fig. 5, in the core analysis program of the present embodiment, the nuclear constant calculation step (step S101), the fitting-type derivation step (step S102), and the core calculation step (step S103) are performed in largely different ways.

In the nuclear constant calculation step S101, a plurality of calculation points are set in each of the three parameters of the fuel temperature T _f , the coolant density ρ _m , and the coolant temperature T _m , and at the calculated point set for each parameter, Nuclear constants are calculated using the nuclear constant calculation code.

6 is a table showing the calculation points set for each parameter. As shown in Fig. 6, in this embodiment, six calculation points are set in three parameters. At this time, when trying to calculate the nuclear constant by the number of all cases at each calculation point, it is necessary to perform a total of 216 calculations (6x6x6). However, in this embodiment, an orthogonal table is used to reduce the number of calculations. In this embodiment, a total of 36 calculations are performed by using an orthogonal table with three factors and six levels. As a result, by performing 36 calculations using the L36 orthogonal table, the same calculation result as in the case of 216 calculations can be obtained.

When the nuclear constant at the predetermined calculation point is calculated by the nuclear constant calculation step S101, the fitting equation of the nuclear constant is derived by the fitting equation derivation step S102 using the calculated nuclear constant. In fitting type derivation step S102, a term setting step, a fitting type preparation step, and a robustness evaluation step are performed.

In the term setting step, the form of the term indicating the behavior of the nuclear constant calculated by the nuclear constant calculation step is set for each parameter while the one parameter is displaced and the other parameter is fixed.

Here, the forms of the term include x, 1 / x, √x, and 1 / √x. By multiplying them n times, for example, (x) ⁴ , (1 / √x) ⁴ , (x) ³ , (1 / x) ^2, and the like are used. It may indicate the behavior of nuclear constants.

FIG. 7 is a table showing a calculation point when one parameter is displaced and another parameter is fixed. As shown in FIG. 7, when the parameter of the fuel temperature T _f is displaced, the parameters of the coolant temperature T _m and the coolant density ρ _m are fixed at a predetermined coolant temperature and coolant density. Similarly, when the parameter of the coolant temperature T _m is displaced, the parameters of the fuel temperature T _f and the coolant density ρ _m are fixed at a predetermined fuel temperature and coolant density. Similarly, when the parameter of the coolant density rho _m is displaced, the parameters of the fuel temperature T _f and the coolant temperature T _m are fixed to a predetermined fuel temperature and coolant temperature.

8 is a table in the form of terms showing the behavior of nuclear constants set for each parameter. As shown in FIG. 8, as the nuclear constant, the diffusion coefficient D ₁ of the fast neutron group, the absorption cross-sectional area of the fast neutron group Σ _a1 , the removal cross-sectional area of the fast neutron group Σ _r , and the generation cross-sectional area of the fast neutron group (νΣ _f1 ), diffusion coefficient (D ₂ ) of the thermal neutron group, absorption cross-sectional area (Σ _a2 ) of the thermal neutron group, and generation cross-sectional area (νΣ _f2 ) of the thermal neutron group. For example, in the diffusion coefficient D ₁ , when the parameters of the fuel temperature T _f are displaced to fix the parameters of the coolant temperature T _m and the coolant density ρ _m , the diffusion coefficient D _1. ) Can be most appropriately expressed in terms of (√T _f ). Similarly, for seven items of nuclear constants, a total of 21 terms are set by displacing three parameters, respectively. In addition, when there is no dependency on each parameter, it becomes (-). When the form of the term in each item of a nuclear constant is set by a term setting step, it transfers to a fitting type preparation step continuously.

In the fitting formula preparation step, a fitting formula for an arbitrary nuclear constant is created by combining terms representing the behavior of nuclear constants set for each parameter, adding a predetermined term, or deleting a predetermined term. For example, in the case of creating a fitting equation for the absorption cross-sectional area Σ _a2 of the thermal neutron group, multiplying (1 / √T _f ), which is a form of a term depending on the coefficient a ₁ and the fuel temperature, based on the table of FIG. in addition to that, the coefficients a ₂ and as multiplied by the term of the form (√T _m) depending on the coolant temperature, the coefficient a ₃ and as multiplied by the term of the form (1 / ρ _m) depending on the coolant density, the constant term a ₀ , The fitting formula shown in Formula (3) is prepared.

Thereby, the fitting formula of the absorption cross-sectional area (Σ _a2 ) of the created thermal neutron group depends on fuel temperature T _f , coolant temperature T _m , and coolant density ρ _m . When the fitting formula of the nuclear constant is prepared by the fitting formula preparation step, the process shifts to the robustness evaluation stage.

In the robustness evaluation step, it is evaluated using the Akaike information quantity criterion (AIC) shown in equation (4) whether or not the created nucleus constant fitting formula has robustness. The fitting formula evaluated to have robustness based on this Akaike information quantity criterion (AIC) enables stable calculation of the nuclear constant even when there is a variation error such as noise in each parameter. In the Akaike information quantity criterion (AIC), N is a data score (the number of calculation points), δ is a fitting error (the difference between the nuclear constant at each calculation point and the nuclear constant obtained from the fitting equation), and M is a fitting. It is a constant of the formula. And the smaller the AIC, the higher the robustness.

9 is a flowchart for deriving the fitting formula of the nuclear constant. With reference to FIG. 9, the control flow which derives the fitting formula used in core calculation step S103 is demonstrated. In the fitting equation derivation step S102, when a fitting equation having robustness is derived, a fitting equation with high robustness is derived by repeating the fitting equation preparing step and the robustness evaluation step.

Specifically, in the fitting formula preparation step, a fitting formula of the absorption cross-sectional area Σ _a2 of the thermal neutron group of formula (3) is prepared (step S1). In the fitting formula preparation step, the fitting formula of the formula (3) is (1 / √T _f ), which is a term depending on the fuel temperature, (√T _m ), which is a term depending on the coolant temperature, based on the table of FIG. depending on the coolant density anti - Human (1 / ρ _m) an optionally create a fitting expression add new wherein combining (step S2). Here, as the new term, (1 / √T _f ) ² , (1 / √T _f ? √T _m ), (1 / √T _f ? 1 / ρ _m ), (√T _m ) ² , (√T _m ? 1 / ρ _m ), (1 / ρ _m ) ^2, and the like. In addition, the fitting equation to which the new term is added is illustrated in equation (5).

Then, the fitting formula of step S1 and the fitting formula of step S2 which were created in the fitting formula preparation step are evaluated in a robustness evaluation step. That is, in the robustness evaluation step, a fitting AIC of the formula (4) and a plurality of fitting formulas to which new terms are added are derived, and the smallest AIC is selected among these fitting formulas (step S3). ).

In step S3, if a fitting equation is selected, it will transfer to a fitting equation preparation step again. In the fitting formula preparation step, any one of the terms of the selected fitting formula is deleted (step S4). Also, a fitting equation in which one term is omitted is shown in Equation 6.

Thereafter, the fitting equation of the selected step S3 and the fitting equation of step S4 created in the fitting equation preparing step are evaluated in the robustness evaluation step. That is, in the robustness evaluation step, the AIC of the selected fitting equation and the AIC of the fitting equation from which any one is omitted are derived, and the one with the smallest AIC is selected from these fitting equations (step S5).

In the robustness evaluation step, it is determined whether or not the AIC is minimized (step S6), and steps S5 to S5 are repeatedly executed until the AIC becomes minimum. As a result, in the fitting equation derivation step, a fitting equation with minimum AIC is derived. In addition, although the case where the fitting formula of the absorption cross-sectional area (( _sigma ) _a2 ) of a thermal neutron group was derived was demonstrated, the same also applies to the case where the other fitting formula in a nuclear constant is derived. In addition, the fitting formula in each item of the nuclear constant derived by the above-mentioned control flow is shown in equation (7).

Here, equation (5) is a fitting equation for the diffusion coefficient D ₁ of the fast neutron group. Equation (6) is a fitting equation of the absorption cross-sectional area (Σ _a1 ) of the fast neutron group. Equation (7) is a fitting equation for the generation cross-sectional area (νΣ _f1 ) of the fast neutron group. Equation (8) is a fitting equation for the removal cross-sectional area (Σ _r ) of the fast neutron group. Equation (9) is a fitting equation for the diffusion coefficient D ₂ of the thermal neutron group. Formula (10) is a fitting formula of the absorption cross-sectional area (Σ _a2 ) of the thermal neutron group. Equation (11) is a fitting equation for the generation cross-sectional area (νΣ _f2 ) of the thermal neutron group. When the fitting formula of the nuclear constant is derived in the fitting formula derivation step S102, the routine proceeds to the core calculation stage S103.

In the core calculation step S103, the nuclear constants are set for the plurality of fuel nodes 35 in accordance with the operating conditions of the core 5, respectively. At this time, the nuclear constant is derived by substituting the fuel temperature T _f , the coolant temperature T _m , and the coolant density ρ _m , which are the operating conditions of the core 5, into the fitting equation of the derived nuclear constant. The derived nuclear constant is set in correspondence with each fuel node 35, and the core characteristics are evaluated by performing core calculation using a core calculation code.

According to the above structure, in the core analysis program, the nuclear constant suitable for the plurality of fuel nodes 35 can be derived without calculating the nuclear constants for the number of the plurality of fuel nodes 35. Thereby, the number of calculations of the nuclear constant calculation in this embodiment can be made smaller than the number of calculations of the nuclear constant calculation by the number of fuel nodes 35, thereby significantly reducing the calculation time. can do. In addition, since the nuclear constant suitable for the plurality of fuel nodes 35 can be set, the core calculation can be performed with high accuracy, and the nuclear characteristics in the core can be appropriately evaluated. In addition, since the nuclear constants are derived using a fitting formula created using the concept of quality engineering, the nuclear constants calculated by the fitting formulas can be regarded as highly reliable and highly accurate, and furthermore, the nucleus in the core 5 is further developed. Evaluation of a characteristic can be made highly reliable.

10 is a table which shows the standard deviation and the maximum error of the reactivity coefficient obtained using the core analysis program of a present Example. As shown in FIG. 10, in the operation of the core 5 including a boiling state, the reactivity coefficient has a standard deviation of 0.056% ΔK / K (56pcm), and the maximum error is 0.235% ΔK / K. (235pcm). In the operation of the core 5 that does not include a boiling state, the reactivity coefficient has a standard deviation of 0.028% ΔK / K (28pcm) and a maximum error of 0.107% ΔK / K (107pcm). It is. For this reason, since the reactivity coefficient obtained using the core analysis program of a present Example can be made into high precision, it turns out that the core analysis program of a present Example is highly reliable.

In addition, in this core analysis program, nuclear constant calculation in the some calculation point set to each parameter is selected and performed using an orthogonal table. As a result, it is possible to reduce the number of calculations as compared with the case of calculating the nuclear constant with the number of all cases at each calculation point, while obtaining the same calculation result as the case of calculating the nuclear constant with the number of all cases. It can be shortened.

In this core analysis program, the created fitting equation can be evaluated by using the Akaike information quantity criterion (AIC), so that the robustness can be made high, so that the fitting equation can be made highly reliable.

As described above, the core analysis program according to the present invention is useful in the case of highly accurate evaluation of nuclear characteristics in the core while shortening the calculation time, and in particular in the case of performing highly reliable core design incorporating the concept of quality engineering. Suitable for

1: reactor
5: core
6: fuel assembly
10: fuel rod
11: cladding tube
12: gap between aggregates
13: coolant
14: control rod
20: cell
30: analysis target area
35: fuel node

Claims (5)

A hardware constant for evaluating nuclear characteristics in the core by performing a nuclear constant calculation for calculating a nuclear constant of a fuel assembly loaded in the core and performing a core calculation based on the nuclear constant calculated by the nuclear constant calculation. A storage medium storing a core analysis program executable in the
A nuclear constant calculation step of setting a calculation point on each of a plurality of parameters depending on the nuclear reaction in the fuel assembly, and calculating the nuclear constant at the calculation point set for each parameter;
A fitting equation deriving step of deriving a fitting equation of the nuclear constant corresponding to the displacement of each parameter from the calculated nuclear constant at the calculated point to have robustness;
And a core calculation step of deriving the nuclear constant corresponding to the plurality of parameters changed in accordance with the operating conditions of the core from the fitting equation and performing the core calculation based on the derived nuclear constant. A storage medium that stores a core interpretation program. The nuclear constant calculation step according to claim 1, characterized in that in the nuclear constant calculation step, the calculation point set for each parameter is selected based on an orthogonal table, and the nuclear constant calculation is performed based on the calculation point of each selected parameter. The storage medium which stored the core analysis program. The method of claim 1 or 2, wherein the fitting derivation step,
A term setting step of setting, for each parameter, a term representing a behavior of the nuclear constant when the one parameter is displaced and the other parameter is fixed based on the nuclear constant calculated by the nuclear constant calculation step; ,
A fitting equation preparation step of creating the fitting equation by combining the terms representing the behavior of the nuclear constants set for each of the parameters;
And a robustness evaluation step of evaluating the created robustness of the fitting type. 4. The storage medium according to claim 3, wherein in the robustness evaluation step, the robustness of the fitting equation is evaluated using an akaike information quantity criterion. The fuel cell according to claim 1 or 2, wherein the plurality of parameters include a fuel temperature of the fuel assembly, a temperature of the coolant introduced into the core, and a density of the coolant introduced into the core, among which at least the density of the coolant is included. A storage medium storing a core analysis program, characterized by using two parameters.

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