KR101439999B1 - The constant production device and method for the coarse mesh reactor core analysis based on three-dimensional transport calculation - Google Patents

Abstract

Disclosed are a device and a method for producing constants for reactor core analysis based on three-dimensional reactor core transport analysis. The method comprises the steps of: generating detailed cross-section information for neutron reaction by nuclide, based on an evaluated nuclear data file (ENDF); generating the entire neutron spectrum and distribution of a reactor core by performing a three-dimensional reactor core transport calculation, based on the generated detailed cross-section information for neutron reaction by nuclide; generating a constant by coarse mesh by editing constants for coarse meshes based on the generated entire neutron spectrum and distribution of the reactor core; and generating a three-dimensional neutron flux distribution and an effective multiplication factor by performing a three-dimensional reactor core coarse mesh diffusion calculation, based on the generated constant by coarse mesh.

Description

Technical Field [0001] The present invention relates to a constant production device and method for generating a constant for a lattice girder core based on a nuclear reactor core transport analysis,

The present invention relates to a constant production apparatus and method including a neutron nuclear reactor cross sectional area used as input information of a reactor core analysis program, and more particularly, to a constant water production apparatus and method based on three-dimensional reactor core transportation analysis .

In order to analyze the nuclear properties of the core, it is necessary to divide the core into a nuclear fuel assembly unit (or 4-minute fuel assembly unit) in the radial direction and a vacancy grid consisting of 10 to 24 nodes in the axial direction, To calculate the multiplication factor and the neutron flux distribution of the core. The various constants representing the characteristics of the lattice lattice, such as the cross-sectional area of the neutron and the boundary discontinuity factor, are analyzed by analyzing the shape of the two-dimensional fuel assembly corresponding to each lattice lattice by neutron transport analysis and combustion calculation, do.

Neutron analysis of reactor core requires neutron cross-sectional area for each nuclide. Generally, the analysis for the core design is based on dividing the core into a homogenized three - dimensional excitation grid. The neutron cross - section of each grid is obtained by two - dimensional neutron transport analysis which precisely simulates the node inside the fuel assembly unit. At this stage, the structure assuming an infinite arrangement of single fuel assemblies is interpreted, so that the effect of the neutron spectrum due to the shape of the three-dimensional core or the constituent material can not be considered, which is a fundamental factor causing uncertainty in the core analysis.

Korean Patent Publication No. 2011-0123646

The problem to be solved by the present invention is to replace the production of constants by the transportation analysis of the two-dimensional fuel assemblies with the three-dimensional core transportation analysis and to eliminate the errors that can be caused by assuming the two-dimensional infinite arrangement in the energy and spatial distribution analysis of neutrons in the constant calculation .

In addition, the combustion calculation is performed in the 3 - core core transportation calculation, and the purpose of simplifying the construction of the transportation analysis is to remove the parameter of the combustion degree in the 3 - dimensional core grating diffusion calculation.

In a constant water production apparatus for three-dimensional core analysis,

An input unit for receiving data for generating detailed neutron reaction cross-sectional information of a nuclear type, and a nuclear type detailed neutron reaction cross-sectional information based on the input data, And a control unit for analyzing core transportation.

In the method of producing a constant for three-dimensional core analysis,

Generating nuclear neutron response cross-sectional information based on an ENDF (Evaluated Nuclear Data File) nuclear data, performing a three-dimensional core transport calculation based on the generated nuclear type detailed neutron reaction cross-sectional information, And generating a distribution, generating a constant for the excitation lattice by editing the constant for the excitation lattice based on the generated total core neutron spectrum and distribution, and generating a constant for each excitation lattice based on the generated constant for each excitation lattice, And performing a lattice diffusion calculation to generate a three-dimensional neutron flux distribution and an effective multiplication factor.

According to the apparatus and method for producing core constants for core analysis based on the nuclear reactor core transport analysis according to the present invention, it is possible to estimate the energy and spatial distribution of neutrons in the constants calculation by assuming a two- The accuracy of the core analysis can be improved.

Also, by performing the combustion calculation in the 3 - core core transportation calculation, the combustion analysis calculation becomes unnecessary in the calculation of the 3 - core core dislocation lattice diffusion, and the construction of the transportation analysis can be simplified.

1 is a block diagram showing components of a constant water production apparatus for three-dimensional core analysis according to an embodiment of the present invention.
2 is a block diagram illustrating components of a control unit, in accordance with an embodiment of the present invention.
3 is a cross-sectional view showing an upper part of a general nuclear reactor core model.
4 is a side view showing a side view of a general nuclear reactor core model in general.
5 is an exemplary view showing a pressurized light water reactor core analysis procedure according to an embodiment of the present invention.
6 is an exemplary view comparing a general output distribution error and an output distribution error of the present invention.
FIG. 7 is a diagram illustrating an example of a comparison between a normal core augmentation factor error and a core effective augmentation factor error of the present invention.
FIG. 8 is a flowchart illustrating a procedure of a method for producing a constant for three-dimensional core analysis according to an embodiment of the present invention.

Hereinafter, embodiments of the present invention will be described in detail with reference to the accompanying drawings. In the drawings, the same reference numerals are used to designate the same or similar components throughout the drawings. In the following description of the present invention, a detailed description of known functions and configurations incorporated herein will be omitted when it may make the subject matter of the present invention rather obvious or understandable to one skilled in the art.

The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of the invention. When referring to " output " or " calculation " or " interpretation " When referring to " generation " or " production ", the above expressions should be interpreted equally to each other. When referring to " information " or " data " Also, when referring to a "database" or "library", the above expressions should be interpreted identically to each other.

1 is a block diagram showing components of a constant water production apparatus for three-dimensional core analysis according to an embodiment of the present invention.

Referring to FIG. 1, a constant water production apparatus 1 for a three-dimensional core analysis can simulate a core in detail through a three-dimensional transportation analysis. The three-dimensional core analysis constant producing apparatus 1 may include at least one of an input unit 110, a control unit 120, and a storage unit 130.

The input unit 110 may receive data for generating nuclear type detailed neutron response cross-sectional information. The data may be an Evaluated Nuclear Data File (ENDF) nuclear material. The ENDF is the information that Brookhaven National Laboratory collects and evaluates the cross-sectional data created by each institution. In addition, ENDF can receive data without any limitation when applying to the Sigma Center in the institute.

The control unit 120 can obtain the output distribution and the effective multiplication factor of the reactor core using the three-dimensional core transportation analysis for reactor core analysis. The control unit 120 can obtain the solutions of the three-dimensional core diffusion equation and the eigenvalues to calculate the neutron flux distribution. Since the control unit 120 performs the calculation of the three-dimensional core diffusion in a state in which the degree of combustion for each of the vacancy lattices is fixed, the degree of combustion may be excluded in the variables constituting the constant database function for the core analysis. The control unit 120 can simplify the constant form due to the excluded combustion, and eliminate the occurrence of the error due to the combustion function.

The storage unit 130 stores the ENDF nuclear data information provided to the input unit 110, the information of the nuclear species detailed neutron reaction cross-sectional area calculated by the controller 120, the equivalent constants for each node calculated by the controller 120, The library to be used in the calculation, and the resultant interpreted value of the core can be stored.

2 is a block diagram illustrating components of a control unit, in accordance with an embodiment of the present invention.

2, the control unit 120 may include at least one of a library production unit 210, a column count production unit 220, a column count editing unit 230, and a core analysis unit 240.

The library production unit 210 can generate nuclear type detailed neutron response cross-sectional information necessary for neutron analysis of the reactor core. The library production unit 210 may use the information of the nuclear type detailed neutron reaction cross-sectional area provided by the input unit 110.

The military water producer 220 can substitute the core core transportation calculation in the existing two-dimensional nuclear fuel assembly neutron transportation calculation using the nuclear type detailed neutron reaction sectional area information generated in the library production unit 210. In addition, the military water production unit 220 can calculate the combustion for analyzing the core characteristics of the direct fuel combustion in the 3-core core transportation calculation. Since the combustion calculation is performed in the 3-dimensional core transport calculation, the calculation of the combustion in the 3-dimensional core grating diffraction calculation may become unnecessary.

Equation (1) is a mathematical expression for calculating a three-core core transportation.

는 입자의 방향속(angular flux)을 의미하고,

는 소격적 핵반응 단면적을 의미하며,

는 입자의 선원항을 의미한다.

은 입자의 위치를 나타내는 벡터를 의미하고,

은 입자가 움직이는 방향을 나타내는 벡터를 의미하며,

는 입자의 에너지를 의미한다. 여기서,

는 원자로 노심을 구성하는 물질의 연소에 따라 변화하는 값이다. In Equation (1)

Means the angular flux of the particles,

Means the cross-sectional area of the nuclear reaction,

Means the source term of the particle.

Means a vector representing the position of the particle,

Means a vector representing the direction in which particles move,

Means the energy of a particle. here,

Is a value that varies with the combustion of the material that constitutes the reactor core.

을 구하기 위한 것이다. <수학식 1>은 상기 방향속을 기초로 원자로 노심을 구성하고 있는 물질의 연소상태 즉, 입자의 수밀도를 계산할 수 있다. 또한 <수학식 1>은 수밀도를 기반으로 다음 연소단계의 계산을 위한 핵반응 단면적

를 산출할 수 있다.&Lt; EMI ID = 1.0 >

. Equation (1) can calculate the combustion state of the material constituting the reactor core, that is, the number density of particles based on the directional direction. Equation (1) is based on the number density, and the nuclear reaction cross-sectional area

Can be calculated.

The military water production unit 220 produces a restart file for branch calculation at the core combustion point based on the number density of particles and changes the operation conditions in the grid such as boric acid concentration, fuel temperature, coolant temperature, To perform branch calculations and combustion calculations. The re-start file for branch computation may include the number density distribution of all the particles making up the core and the calculated neutron spectra and spatial distribution.

The column count editing unit 230 can produce a database for functioning the constants for the spacing grid calculated by the three-dimensional core transportation calculation of the row count producing unit 220. [

The constellation number editing unit 230 may process at least one database of neutrons, cross-sectional areas, and boundary discontinuity factors for each of the excitation lattices and function constants to edit the constellation database for each excitation lattice.

The core analyzer 240 can divide the reactor core into a plurality of three-dimensionally spaced apart lattice zones and analyze the distribution and change of the neutron flux.

The core analyzer 240 can perform the three-dimensional core grating spreading calculation based on the constant database for each of the grid lines generated by the grid editor 250. The calculation of the three-core core-excitation lattice diffusivity can eliminate the constants necessary for the combustion calculation in the constants for each lattice grid since the combustion calculation becomes unnecessary. Also, the calculation of the three-dimensional core gap grid diffusivity can be made in a simple form since it can consist of constants for a specific combustion degree.

Equation (2) is a mathematical expression for calculating the three-dimensional core dislocation lattice diffusion.

는 중성자속(flux)을 의미하며

는 확산계수(diffusion coefficient)를 의미한다.

는 흡수, 산란, 핵분열 핵반응 단면적을 각각 의미하고,

는 핵분열당 생성되는 중성자수를 의미하며,

는 핵분열 중성자의 에너지 스펙트럼을 의미한다.

은 중성자의 위치를 의미하고,

는 중성자의 에너지를 의미하며,

는 유효증배계수를 의미한다. In Equation (2)

Means a neutron flux

Means a diffusion coefficient.

, Respectively, mean the absorption, scattering, and cross-sectional area of the fission nuclear reaction,

Means the number of neutrons generated per fission,

Means the energy spectrum of fission neutrons.

Is the position of the neutron,

Means the energy of a neutron,

Means the effective multiplication factor.

는 원자로 노심을 구성하는 물질의 연소에 따라 변화하는 값으로 연소계산을 수행하여 구해야 하는 변수이지만 상수생산 과정에서 미리 연소계산을 수행하여 산출된 값이므로 소격격자 확산계산에서는 연소에 따른 변화를 고려할 필요가 없게 되었다. In Equation (2)

Is a variable that must be obtained by carrying out combustion calculation at a value that changes according to the combustion of the material constituting the reactor core. However, since it is the value calculated by performing the combustion calculation in advance in the production process of the constant, .

과 유효증배계수

를 구할 수 있다. 노심해석부(240)는 산출된 중성자속 분포 및 증배계수를 기초로 노심의 출력분포 및 유효증배계수를 산출할 수 있다.
Equation (2)

And effective multiplication factor

Can be obtained. The core analyzing unit 240 can calculate the output distribution and effective multiplication factor of the core based on the calculated neutron flux distribution and the multiplication factor.

FIG. 3 is a sectional view showing an upper part of a general nuclear reactor core model, and FIG. 4 is a side view showing a side view of a general nuclear reactor core model.

Referring to FIGS. 3 to 4, the reactor core 10 may include a nuclear fuel bundle as a central portion of the reactor. The reactor core 10 may further include at least one of a moderator for decelerating the high-speed neutrons emitted from the nuclear fuel assembly to thermal neutrons, and a coolant for eliminating the generated excessive heat. Here, at least one of water and graphite may be used as the moderator depending on the type of reactor.

The reactor core 10 can analyze the neutron flux in the core through the core analyzer 240. The core analyzer 240 can divide the inside of the core into a plurality of three-dimensional grids and analyze the distribution and change of the neutron flux in the core.

5 is an exemplary view showing a pressurized light water reactor core analysis procedure according to an embodiment of the present invention.

5, the pressurized light water reactor core analysis can be performed using at least one of the library production unit 210, the column water production unit 220, the column constant editing unit 230, and the core analysis unit 240.

The input unit 110 may receive data for generating nuclear type detailed neutron response cross-sectional information. The data may be ENDF nuclear material.

The library production unit 210 can generate the nuclear type detailed neutron reaction cross-sectional area 520 required for analyzing neutrons of the reactor core using the data input from the input unit 110.

The library production unit 210 can generate a library of Muti-group 530 based on the generated nuclear species detailed neutron reaction cross-section 520. [ The multidomain cross-sectional library 530 is a library for analyzing the diffusion and transport phenomena of neutrons based on the multidomain theory.

The MRC generator 220 can simulate the entire core using the multidimensional cross-sectional library 530 generated in the library production unit 210 into a precise three-dimensional core model 540. The three-dimensional core model 540 can be confirmed that the fuel assembly 545 is composed of a detailed model including the fuel rod 541 in the enlarged image 540 '.

The headquarters production unit 220 can perform the three-core core transport calculation 550 using the simulated three-dimensional core model 540. The 3-core core transport calculation (550) can also calculate the combustion to analyze the core characteristics of direct fuel combustion.

The tower water producing section 220 can generate the core whole neutron spectrum and distribution 560 using the calculated three-dimensional core transportation calculation 550. The whole neutron spectrum and distribution (560) of the core can reflect the effect of the geometry of the entire core.

The constellation editor 230 may generate a Few-group library 570 using the generated core full neutron spectrum and distribution 560. [ The minority library 570 may include a constant for the excitation grid. The constellation number editor 230 can edit constants for the exciton lattice for the functioning of the excitation lattice using the created library of prime groups 570. [

The string constellation editor 230 can generate constants 580 for each excitation matrix using the edited constants for the spacing matrix. The ruling number editing unit 230 can generate a constant 580 for each of the nodes by functioning the equivalent constants for each node by the boron concentration, the fuel temperature, and the coolant density.

The core analyzer 240 may perform the 3-D core disruption grid diffusion calculation 590 based on the constant 580 for each of the excitation gratings. The 3-D core grating diffraction calculation (590) eliminates the need for combustion calculations and excludes the necessary constants in the combustion calculation among the constants for each lattice spacing. The three-dimensional core excitation lattice diffusion calculation 590 may consist solely of constants for a particular degree of combustion. In other words, since the three-dimensional core grating diffusion calculation (590) is performed in a state in which the degree of combustion is determined for each of the lattice grids, the degree of burning can be excluded from the parameters, the constant form can be simplified, The factor can be eliminated.

) 및 유효증배계수(

)(600)를 산출할 수 있다.
The core analyzing unit 240 calculates the neutron flux distribution by using the calculated neutron diffusion calculation (590)

) And effective multiplication factor (

) &Lt; / RTI &gt;

6 is an exemplary view comparing a general output distribution error and an output distribution error of the present invention. The output distribution error in Fig. 6 is the result of a comparative analysis of 1 / 8th symmetric core of one cycle of Yeonggwang Nuclear Power Plant No. 3.

Referring to FIG. 6, the output distribution refers to the thermal output spatial distribution within the reactor core.

FIG. 6 shows that the core analysis is performed using constants (two-dimensional constants) produced by the conventional method and constants (three-dimensional constants) produced by the method of the present invention. Each constant represents the percent difference of the radial power distribution for each fuel assembly by performing core analysis. FIG. 6 shows the output distribution error of the three-dimensional constant calculation is smaller than the output distribution error of the two-dimensional constant calculation.

FIG. 7 is a diagram illustrating an example of a comparison between a normal core augmentation factor error and a core effective augmentation factor error of the present invention. 7 is a result of a comparative analysis of a 1/8 symmetric core of the first cycle of Yonggwang Nuclear Power Plant No.3.

Referring to FIG. 7, the effective multiplication factor is a ratio of the total number of neutrons generated to the total number of neutrons lost due to absorption, leakage, or the like within a certain time.

FIG. 7 shows that the core analysis is performed using constants (two-dimensional constants) produced by the conventional method and constants (three-dimensional constants) produced by the method of the present invention. FIG. 7 shows a smaller value of the effective multiplication error of the three-dimensional constant calculation than the effective multiplication error of the two-dimensional constant calculation. That is, FIG. 7 means that the effective multiplication factor of the three-dimensional constant calculation is smaller than the actual effective multiplication factor of the two-dimensional constant calculation.

8 is a flowchart illustrating a method for producing a constant for a three-dimensional core analysis according to an embodiment of the present invention.

Referring to FIG. 8, the method for producing a constant for three-dimensional core analysis accurately calculates the output distribution and effective multiplication factor of a core by a three-dimensional core analysis.

The library production unit 210 generates nuclear type detailed neutron reaction cross-sectional information 520 (S110). The library production unit 210 can generate the nuclear type detailed neutron response cross-sectional information 520 using the ENDF nucleus data 510.

The head count generation unit 220 performs a three-core core transport calculation 550 based on the nuclear type detailed neutron response cross-sectional information 520 generated in the library production unit 210 (S120). The military water producer 220 can perform the combustion calculation for analyzing the core characteristics according to the combustion of the direct fuel in the three-core core transportation calculation 550. [ Also, the tower water producing section 220 generates a core whole neutron spectrum and distribution 560 based on the calculated three-dimensional core transportation calculation 550. The neutron spectra and distribution (560) used in the editing of the lattice constants such as the neutron response cross-sectional area and the boundary discontinuity factor by using the three-dimensional core transportation calculation (550) Can be reflected.

The constellation number editing unit 230 edits constants for the excitation lattice using the generated core whole neutron spectrum and distribution 560 (S130). The controller 120 can edit the constants for the excitation lattice to function as a branch calculation for changing the operation conditions in the lattice, such as the boric acid concentration, the fuel temperature, the coolant temperature, and the control rod insertion, and the constants for the excitation lattices. The constellation number editing unit 230 generates constants 580 for each excitation grid using the edited constellation for the excitation grid.

The core analyzer 240 performs a three-dimensional core grating diffusion calculation 590 using the generated constants 580 for each of the excitation gratings (S140). The core analyzer 240 can calculate the three-dimensional neutron flux and the effective multiplication factor 600 of the core through the three-dimensional core dislocation lattice diffusion calculation 590. The core analyzing unit 240 may exclude the parameters necessary for the calculation of the combustion that is unnecessary in the three-dimensional core dislocation grid diffusion calculation 590. [ In addition, the core analysis unit 240 can be constructed in a simpler form since it can be configured only with a constant for a specific combustion degree instead of the excluded variables.

The present invention can also be embodied as computer-readable codes on a computer-readable recording medium. A computer-readable recording medium includes all kinds of recording apparatuses in which data that can be read by a computer apparatus is stored. Examples of the computer-readable recording medium include a ROM, a RAM, a CD-ROM, a magnetic tape, a floppy disk, an optical data storage device, and the like, and may be implemented in the form of a carrier wave (for example, transmission via the Internet) . The computer-readable recording medium may also be distributed to networked computer devices so that computer readable code can be stored and executed in a distributed manner.

While the present invention has been particularly shown and described with reference to exemplary embodiments thereof, it is clearly understood that the same is by way of illustration and example only and is not to be taken by way of limitation in the embodiment in which said invention is directed. It will be understood by those skilled in the art that various changes in form and detail may be made therein without departing from the scope of the appended claims.

1: Constant production device for three-dimensional core analysis 10: reactor
110: input unit 120:
130: storage unit 210: library production unit
220: Military water production section 230: Military water section editor
240: Core analysis section

Claims (7)

In a constant water production apparatus for three-dimensional core analysis,
An input unit for receiving data for generating nuclear type detailed neutron response cross-sectional information;
A library production unit for generating the nuclear type detailed neutron response cross-sectional information based on the input data;
A three-dimensional core transport calculation based on the generated nuclear type detailed neutron reaction cross-sectional information, to generate a nuclear whole neutron spectrum and distribution;
A ruling number editing unit for editing constants for the excitation lattice based on the generated whole core neutron spectra and distribution to generate constants for each excitation lattice; And
A core analysis unit for performing diffusion calculation for each three-dimensional core excitation lattice on the basis of the generated constellation for each excitation lattice to generate a three-dimensional neutron flux distribution and an effective multiplication factor,
Wherein the reactor is provided with a reactor core and a reactor core.
The method according to claim 1,
Wherein the input data is an ENDF (Evaluated Nuclear Data File) nuclear material. The apparatus for producing a constant for a vacancy lattice core analysis based on a nuclear reactor core transport analysis.
delete The method according to claim 1,
Wherein the three-dimensional core transport calculation is to calculate the number density of particles for analyzing the core characteristics of the nuclear fuel according to the combustion of the nuclear fuel.
The method according to claim 1 or 4,
The three-dimensional core dislocation lattice diffusivity calculation is characterized in that the combustion degree variable is changed to a constant through the combustion calculation and the error according to the functioning of the combustion degree parameter is eliminated. Constant water production equipment.
The method according to claim 1,
Wherein the nuclear neutron spectrum and distribution of the core are reflected in characteristics of the geometry of the core as a whole.
In the method of producing a constant for three-dimensional core analysis,
Generating nuclear neutron specific cross-sectional area information based on an Evaluated Nuclear Data File (ENDF) nuclear data;
Performing three-dimensional core transport calculations based on the generated nuclear type detailed neutron reaction cross-sectional information to generate a nuclear whole neutron spectrum and distribution;
Generating a constellation for each excitation lattice by editing constants for the excitation lattice based on the generated whole core neutron spectrum and distribution; And
And generating a three-dimensional neutron flux distribution and an effective multiplication factor by performing a three-dimensional core dislocation lattice diffusion calculation based on the generated constant for each of the excitation lattices. Constant production method for lattice core analysis.

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