CN114203270A - Subgroup parameter calculation method and system suitable for dispersed granular fuel - Google Patents

Subgroup parameter calculation method and system suitable for dispersed granular fuel Download PDF

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CN114203270A
CN114203270A CN202111445233.9A CN202111445233A CN114203270A CN 114203270 A CN114203270 A CN 114203270A CN 202111445233 A CN202111445233 A CN 202111445233A CN 114203270 A CN114203270 A CN 114203270A
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张乾
黄冬
李颂
梁越超
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Harbin Engineering University
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Abstract

The invention discloses a subgroup parameter calculation method and a subgroup parameter calculation system suitable for dispersed granular fuel, wherein the method comprises the following steps: establishing a variant of a dispersed particle fuel cell by adjusting the density of the fuel and the moderator; processing each energy group of each variant through a Monte Carlo method to obtain an effective resonance section of each type; mixing the dispersed particle fuel in each variant with a matrix, sequentially calculating all standard fluxes after fuel areas are homogenized by a characteristic line method, and further sequentially calculating all corresponding background sections; the effective resonance sections and the background sections are in one-to-one correspondence to establish a resonance section table under the dual non-uniform system; the subgroup parameters are solved using pade approximation according to the resonance table. The method does not need to change a model of a traditional fuel assembly calculation program, and simplifies the neutron calculation difficulty of the dispersed particle fuel.

Description

Subgroup parameter calculation method and system suitable for dispersed granular fuel
Technical Field
The invention relates to the technical field of nuclear reactor core design and safety, in particular to a subgroup parameter calculation method and a subgroup parameter calculation system suitable for dispersed granular fuel.
Background
The condition that Fuel particles are randomly dispersed in a matrix exists in spherical nuclear Fuel and full Ceramic micro encapsulated Fuel (FCM) of a high-temperature gas cooled reactor, and a dual heterogeneous system different from a traditional reactor heterogeneous system is presented. Because the self-resonance self-shielding effect exists in the dispersed fuel particles, the fuel particles and the matrix cannot be simply mixed during the neutron calculation, and special treatment needs to be adopted for double non-uniformity.
For resonance processing of dispersed particulate fuels, there is a general need for improvements in resonance calculation methods to existing fuel assembly calculation programs, including the following two: (1) correcting the resonance section by using a Danco factor, wherein the Danco factor is obtained by calculating the collision probability of the spherical shell of the dispersed particles; (2) and the Sanchez-Pomraning method is adopted to improve the transfer calculation flow called in the resonance calculation. However, these improvements all require a programmed modification of the existing fuel assembly calculation program, which makes the neutronics calculation of dispersed particles difficult.
If the fuel assembly calculation program is not changed, modeling conversion can be performed on the dispersed particle fuel by adopting an equivalent-Physical Transformation (RPT), for example, fuel nuclei in the fuel particles are compressed to the inner ring of the fuel rod, and the dual heterogeneous model is equivalently processed into the single heterogeneous model. However, the method needs to adopt a Monte Carlo program to carry out fine modeling on different working conditions of dispersed particle fuel, and the equivalent inner ring radius is obtained through searching, so that the efficiency is low; meanwhile, as the method changes the fuel structure, the radial power distribution of the fuel ball, the rod and the plate can not be calculated.
Therefore, a method capable of simplifying the neutron calculation difficulty of the dispersed particle fuel and calculating the radial power distribution of the fuel spheres, rods and plates without changing the fuel elements is needed to be awaited.
Disclosure of Invention
The present invention is directed to solving, at least to some extent, one of the technical problems in the related art.
To this end, it is an object of the present invention to propose a subgroup parameter calculation method suitable for dispersed particulate fuels, which simplifies the neutronics calculation of dispersed particulate fuels, and which allows the calculation of the radial power distribution of the fuel spheres, rods, plates without changing the shape and dimensions of the fuel elements.
It is another object of the present invention to provide a subgroup parameter calculation system suitable for use with dispersed particulate fuel.
In order to achieve the above object, an embodiment of the present invention provides a subgroup parameter calculation method suitable for dispersed particulate fuel, including the following steps: step S1, establishing a preset number of double non-uniform problem variants by adjusting the density of fuel and the density of moderator in the preset fuel assembly;
step S2, each energy group of each double non-uniform problem variant is processed through a Monte Carlo method, and a microscopic effective resonance section of each type of nuclear reaction with a resonance effect is obtained; step S3, mixing the dispersed particle fuel in each double non-uniform problem variant with a matrix, sequentially calculating all standard fluxes after fuel area homogenization through a characteristic line method, and further sequentially calculating all corresponding background sections; step S4, the microscopic effective resonance sections and the background sections are in one-to-one correspondence to establish a resonance section table under a dual non-uniform system; and step S5, solving subgroup parameters by using Pade approximation according to the resonance section table.
The subgroup parameter calculation method suitable for the dispersed granular fuel in the embodiment of the invention aims at the problem of a reference grid cell formed by fuel balls, rods and plate elements in a calculation target, and establishes a series of variants of the dispersed granular fuel grid cell by adjusting the density of fuel and the density of a moderator; calculating the effective resonance section of each energy group of each variant by a Monte Carlo method, calculating the standard flux of the homogenized material by a characteristic line method, and calculating to obtain a corresponding background section; establishing a resonance section table according to the obtained effective resonance section and the background section; solving subgroup parameters by using Pade approximation according to a resonance cross-section table; the resonance cross-section table and the subgroup parameters jointly form a resonance database suitable for homogenization treatment of the dispersed granular fuel, and by replacing the original multi-group nuclear database, the neutron calculation of the dispersed granular fuel can be realized by simplified modeling of the mixing of the dispersed granular fuel without changing the fuel assembly program of the conventional subgroup method, so that the neutron calculation difficulty of the dispersed granular fuel is simplified; the calculation process allows the calculation of the radial power distribution of the fuel spheres, rods, plates without changing the shape and dimensions of the fuel elements.
In addition, the subgroup parameter calculation method suitable for dispersed particulate fuel according to the above embodiment of the present invention may also have the following additional technical features:
further, in an embodiment of the present invention, the step S3 specifically includes: step S301, mixing the dispersed particle fuel in each double non-uniform problem variant with a matrix to obtain the nuclear density after the fuel area is homogenized; and S302, calling a characteristic line neutron transport calculation method to solve a single group fixed source transport equation of the nuclear density to obtain the standard flux of the fuel area so as to calculate the background section of the g group fuel area relative to the nuclide l.
Further, in one embodiment of the present invention, the homogenized nuclear density of the fuel region is:
Figure BDA0003383907570000021
wherein N isiso,j、
Figure BDA0003383907570000022
Density, V, before and after homogenization of nuclide iso in material number jj、VhomThe volume of material j and fuel zone respectively.
Further, in one embodiment of the present invention, the single group fixed source transport equation is:
Figure BDA0003383907570000031
wherein g is an energy group number, omega is an angle variable of neutron flux density,
Figure BDA0003383907570000032
neutron flux density per space position for g group
Figure BDA0003383907570000033
As a function of the change in the amount of the change,
Figure BDA0003383907570000034
for Laplace gradient operators, λ Σp,gMultiplying the intermediate resonance factor lambda by the g group macroscopic potential scattering cross section sigmap,gFixed source item in the characteristic line neutron transport method; with respect to the fuel region in the homogenization problem,
Figure BDA0003383907570000035
the macroscopic absorption cross section of the g group is equal to the statistical microscopic effective resonance cross section of the absorption reaction of the nuclide l multiplied by the nuclear density of the nuclide l; for the non-fuel regions of the homogenization problem,
Figure BDA0003383907570000036
equal to 0.
Further, in an embodiment of the present invention, the step S302 utilizes
Figure BDA0003383907570000037
Calculating the background section, whereinb,g、σa,gAnd
Figure BDA0003383907570000038
background section, microscopic absorption section and standard flux, σ, of the g groupa,gIs the g-group microscopic effective resonance cross section of the absorption reaction of the nuclide l.
To achieve the above object, another embodiment of the present invention provides a subgroup parameter calculation system suitable for dispersing particulate fuel, comprising: the variant building module is used for building a preset number of double non-uniform problem variants by adjusting the fuel density and the moderator density in a preset fuel assembly; the resonance section module is used for processing each energy group of each double non-uniform problem variant through a Monte Carlo method to obtain a microscopic effective resonance section of each type of nuclear reaction with resonance effect; the background section module is used for mixing the dispersed particle fuel in each double non-uniform problem variant with a matrix, sequentially calculating all standard fluxes after fuel areas are homogenized by a characteristic line method, and further sequentially calculating all corresponding background sections; the resonance section table building module is used for corresponding the microscopic effective resonance sections and the background sections one by one so as to build a resonance section table under a dual non-uniform system; and the solving module is used for solving the subgroup parameters by using the Pade approximation according to the resonance section table.
The subgroup parameter calculation system suitable for the dispersed granular fuel of the embodiment of the invention generates the resonance database suitable for the double non-uniform system by establishing a plurality of double non-uniform systems, can realize the neutron calculation of the dispersed granular fuel by replacing the original multi-group nuclear database and adopting the mixing simplified modeling of the dispersed granular fuel without changing the fuel assembly program of the existing subgroup using method, thereby simplifying the neutron calculation difficulty of the dispersed granular fuel; meanwhile, the radial power distribution of the fuel ball, the rod and the plate can be calculated without changing the shape and the size of the fuel element in the calculation process.
In addition, the subgroup parameter calculation system suitable for dispersed particulate fuel according to the above embodiment of the present invention may also have the following additional technical features:
further, in an embodiment of the present invention, the background section module specifically includes: the mixing unit is used for mixing the dispersed particle fuel in each double non-uniform problem variant with the matrix to obtain the nuclear density after the fuel area is homogenized; and the solving unit is used for calling a characteristic line neutron transport calculation method to solve the single group fixed source transport equation of the nuclear density to obtain the standard flux of the fuel area so as to calculate the background section of the nuclide l in the g group fuel area.
Further, in one embodiment of the present invention, the homogenized nuclear density of the fuel region is:
Figure BDA0003383907570000041
wherein N isiso,j、
Figure BDA0003383907570000042
Density, V, before and after homogenization of nuclide iso in material number jj、VhomThe volume of material j and fuel zone respectively.
Further, in one embodiment of the present invention, the single group fixed source transport equation is:
Figure BDA0003383907570000043
wherein g is an energy group number, omega is an angle variable of neutron flux density,
Figure BDA0003383907570000044
neutron flux density per space position for g group
Figure BDA0003383907570000045
As a function of the change in the amount of the change,
Figure BDA0003383907570000046
for Laplace gradient operators, λ Σp,gMultiplying the intermediate resonance factor lambda by the g group macroscopic potential scattering cross section sigmap,gFixed source item in the characteristic line neutron transport method; with respect to the fuel region in the homogenization problem,
Figure BDA0003383907570000047
the macroscopic absorption cross section of the g group is equal to the statistical microscopic effective resonance cross section of the absorption reaction of the nuclide l multiplied by the nuclear density of the nuclide l; for the non-fuel regions of the homogenization problem,
Figure BDA0003383907570000048
equal to 0.
Further, in one embodiment of the invention, the solution unit utilizes
Figure BDA0003383907570000049
Calculating the background section, whereinb,g、σa,gAnd
Figure BDA00033839075700000410
background section, microscopic absorption section and standard flux, σ, of the g groupa,gIs the g-group microscopic effective resonance cross section of the absorption reaction of the nuclide l.
Additional aspects and advantages of the invention will be set forth in part in the description which follows and, in part, will be obvious from the description, or may be learned by practice of the invention.
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The foregoing and/or additional aspects and advantages of the present invention will become apparent and readily appreciated from the following description of the embodiments, taken in conjunction with the accompanying drawings of which:
FIG. 1 is a flow chart of a subgroup parameter calculation method suitable for use with dispersed particulate fuel in accordance with one embodiment of the present invention;
FIG. 2 is a flow chart of the fabrication of a resonance cross-section table according to one embodiment of the present invention;
FIG. 3 is a schematic illustration of the reference cell structure and the internal structure of dispersed particles of a dispersed particulate fuel element, such as FCM and TRISO, in accordance with an embodiment of the present invention;
FIG. 4 is a schematic diagram of the homogenization process of fuel elements, for example FCM and TRISO, according to one embodiment of the present invention;
FIG. 5 is an embodiment of the present invention238A trend change graph of the U microscopic absorption cross section along with the background cross section;
FIG. 6 is a schematic diagram of a subgroup parameter calculation system suitable for use with dispersed particulate fuel in accordance with one embodiment of the present invention.
Detailed Description
Reference will now be made in detail to embodiments of the present invention, examples of which are illustrated in the accompanying drawings, wherein like or similar reference numerals refer to the same or similar elements or elements having the same or similar function throughout. The embodiments described below with reference to the drawings are illustrative and intended to be illustrative of the invention and are not to be construed as limiting the invention.
A subgroup parameter calculation method and system for a dispersed particulate fuel according to an embodiment of the present invention will be described below with reference to the accompanying drawings, and first, a subgroup parameter calculation method for a dispersed particulate fuel according to an embodiment of the present invention will be described with reference to the accompanying drawings.
FIG. 1 is a flow chart of a subgroup parameter calculation method suitable for use with dispersed particulate fuel in accordance with one embodiment of the present invention.
As shown in fig. 1, the subgroup parameter calculation method for dispersed particulate fuel comprises the following steps:
in step S1, a preset number of dual non-uniformity problem variants are established by adjusting the fuel density, the moderator density, in the preset fuel assembly.
Specifically, a series of problem variants for making a resonance database are established by changing the fuel density of the moderator and the dispersed particles at a certain temperature according to the type of the dispersed fuel elements (preset fuel components) adopted by a preset target reactor, such as fuel spheres, rods and plates.
In step S2, each energy group of each dual non-uniform problem variant is processed by the monte carlo method to obtain the microscopic effective resonance section of each type of nuclear reaction with resonance effect.
That is, for each problem variant, for each resonance nuclide, for each energy group, and for each nuclear reaction type with resonance effect, the Monte Carlo neutron transport calculation software is used for counting the nuclear reaction rate and the neutron flux density, and the nuclear reaction rate and the neutron flux density are divided to obtain the microscopic effective resonance section.
In step S3, the dispersed particle fuel in each dual non-uniformity problem variant is mixed with the matrix, and all the standard fluxes after the fuel regions are homogenized are sequentially calculated by the characteristic line method, so that all the corresponding background sections are sequentially calculated.
Further, in an embodiment of the present invention, step S3 specifically includes:
step S301, mixing the dispersed particle fuel in each double non-uniform problem variant with a matrix to obtain the nuclear density after the fuel area is homogenized;
step S302, a single group fixed source transport equation of the nuclear density is solved by calling a characteristic line neutron transport calculation method, the standard flux of the fuel area is obtained, and the background section of the g group fuel area relative to the nuclide l is calculated.
Specifically, for each problem variant, the dispersed particle fuel is mixed with the matrix, and the nuclear density after fuel area homogenization is calculated by the formula (1):
Figure BDA0003383907570000061
wherein N isiso,j、
Figure BDA0003383907570000062
Density, V, before and after homogenization of nuclide iso in material number jj、VhomThe volume of material j and fuel zone respectively.
After each problem variant is subjected to homogenization, a characteristic line neutron transport calculation method is called to solve a single group fixed source transport equation expressed by a formula (2):
Figure BDA0003383907570000063
wherein g is an energy group number, omega is an angle variable of neutron flux density,
Figure BDA0003383907570000064
neutron flux density per space position for g group
Figure BDA0003383907570000065
As a function of the change in the amount of the change,
Figure BDA0003383907570000066
for Laplace gradient operators, λ Σp,gMultiplying the intermediate resonance factor lambda by the g group macroscopic potential scattering cross section sigmap,gFixed source item in the characteristic line neutron transport method; with respect to the fuel region in the homogenization problem,
Figure BDA0003383907570000067
the macroscopic absorption cross section of the g group is equal to the statistical microscopic effective resonance cross section of the absorption reaction of the nuclide l multiplied by the nuclear density of the nuclide l; for the non-fuel regions of the homogenization problem,
Figure BDA0003383907570000068
equal to 0.
Finally, after the formula (2) is solved by a characteristic line neutron transport calculation method, the standard flux of the fuel area can be obtained, and the background section of the g group fuel area about the nuclide l is further calculated by the formula (3):
Figure BDA0003383907570000069
wherein σb,g、σa,gAnd
Figure BDA00033839075700000610
background section, microscopic absorption section and standard flux, σ, of the g groupa,gA g-population microscopic effective resonance section of the absorption reaction of the nuclide l counted in step S2;
finally, the background section sigma of each problem, each resonance nuclear species and each energy cluster is calculatedb,g。
In step S4, the microscopic effective resonance sections and the background sections are mapped one-to-one to establish a resonance section table under the dual non-uniform system.
Specifically, as shown in fig. 2, each energy cluster and the microscopic effective resonance section of each nuclear reaction with resonance effect counted by the monte carlo neutron transport calculation software in the step S2 are in one-to-one correspondence with the background section in the step six to form a resonance section table under the dual nonuniform system.
In step S5, the subgroup parameters are solved using pade approximation according to the resonance table.
Specifically, a Pade approximation method is adopted to fit each resonance nuclide, each energy group, each subgroup middle section and subgroup probability of nuclear reaction with resonance effect.
Further, after a plurality of dual heterogeneous systems are established, the embodiment of the invention can realize the neutron calculation of the dispersed particle fuel and can also calculate the radial power distribution of the fuel balls, the rods and the plates by adopting the mixing simplified modeling of the dispersed particle fuel in a mode of replacing the original multi-group nuclear database without changing the existing fuel assembly program using the subgroup method.
The calculation method of subgroup parameters suitable for dispersed particulate fuel according to the embodiment of the present invention is further described below by taking the reference cell structure of the dispersed particulate fuel element and the structure inside the dispersed particulate as examples.
A resonance database is constructed by constructing a series of variant problems of dispersed particle fuel reference cells, and the implementation flow of the invention is specifically described by taking FCM fuel as an example.
1) The FCM fuel and TRISO particles in fig. 3 were targeted, and the temperature was chosen to be 300K. The radius R1 of the cell fuel is 0.6252cm, the radius R2 of an air gap (outer ring) is 0.6337cm, the radius R3 of the zirconium cladding is 0.6907cm, and the side length of the cell is 1.65 cm. The TRISO particles are divided into five layers from inside to outside: the fuel core, the buffer zone, the inner layer pyrolytic carbon, the silicon carbide layer and the outer layer pyrolytic carbon have geometric radiuses of 0.0250cm, 0.0340cm, 0.0380cm, 0.04150cm and 0.0455cm in sequence. Table 1 shows the material and species density information in the unit cell, at which the fuel particle fill rate is 40%.
TABLE 1 FCM Unit cell Material information
Figure BDA0003383907570000071
Figure BDA0003383907570000081
2) Based on the moderator and fuel density data in table 1, 37 different dual non-uniformity problems were designed, as shown in table 2. In table 2, the Relative Moderator Density (RMD) is the design condition moderator density/reference moderator density, and the Relative Fuel Density (RFD) is the design condition fuel density/reference fuel density.
TABLE 2 variation of FCM Fuel cells
Figure BDA0003383907570000082
3) Monte Carlo neutron transport calculation software is adopted to count all the problems in the table 2235U and238each energy group of U, each type of microscopic cross section of nuclear reaction with resonance effect, includes absorption (sigma)a) Fission (sigma)f) Scattering (σ)s) And total cross section (σ)t)。
4) As shown in FIG. 4, the particle-matrix region was volume-mixed using equation (1), and for the homogenization problem after mixing, the characteristic line neutron transport method was used, and equation (2) was used to calculate all the problems in Table 2235U and238the fuel zone flux for each energy group of U is calculated using equation (3)235U and238background section of each energy cluster of U.
5) Mixing the obtained product of 3)235U and238each energy group of U, each type of microscopic effective cross section of nuclear reaction with resonance effect, and 4) thereof235U and238the background cross-sections of each energy group of U are in one-to-one correspondence to form an interpolation table of the resonance cross-sections with respect to the background cross-sections as shown in fig. 5.
6) Using Pade approximation method to solve235U and238middle section of U at 300K temperature of subgroup andthe subgroup probability.
7) By utilizing the subgroup middle section and the subgroup probability manufactured in the step 6), the neutron calculation of the FCM fuel grid cells can be directly carried out by applying the fuel assembly program using the conventional subgroup method, and the accuracy equivalent to that of Monte Carlo neutron transport software can be obtained.
To sum up, the subgroup parameter calculation method suitable for the dispersed granular fuel provided by the embodiment of the invention is used for establishing a series of variants of the dispersed granular fuel cells by adjusting the density of the fuel and the density of the moderator aiming at the problem of the reference cells formed by fuel balls, rods and plate elements in a calculation target; calculating the effective resonance section of each energy group of each variant by a Monte Carlo method, calculating the standard flux of the homogenized material by a characteristic line method, and calculating to obtain a corresponding background section; establishing a resonance section table according to the obtained effective resonance section and the background section; solving subgroup parameters by using Pade approximation according to a resonance cross-section table; the resonance cross-section table and the subgroup parameters jointly form a resonance database suitable for homogenization treatment of the dispersed granular fuel, and by replacing the original multi-group nuclear database, the neutron calculation of the dispersed granular fuel can be realized by simplified modeling of the mixing of the dispersed granular fuel without changing the fuel assembly program of the conventional subgroup method, so that the neutron calculation difficulty of the dispersed granular fuel is simplified; the calculation process allows the calculation of the radial power distribution of the fuel spheres, rods, plates without changing the shape and dimensions of the fuel elements.
A subgroup parameter calculation system adapted to disperse particulate fuel according to an embodiment of the present invention will be described next with reference to the accompanying drawings.
FIG. 6 is a schematic diagram of a subgroup parameter calculation system suitable for use with dispersed particulate fuel in accordance with one embodiment of the present invention.
As shown in fig. 6, the system 10 includes: a variant building module 100, a resonance section module 200, a background section module 300, a resonance section table building module 400 and a solving module 500.
Wherein the variant building module 100 is configured to build a preset number of dual non-uniformity problem variants by adjusting fuel density, moderator density in a preset fuel assembly. The resonance section module 200 is used for processing each energy group of each dual non-uniform problem variant through a Monte Carlo method to obtain a microscopic effective resonance section of each type of nuclear reaction with resonance effect. The background section module 300 is used for mixing the dispersed particle fuel in each dual non-uniform problem variant with the matrix, sequentially calculating all standard fluxes after fuel area homogenization by a characteristic line method, and further sequentially calculating all corresponding background sections. The resonance section table construction module 400 is used for corresponding the microscopic effective resonance sections and the background sections one by one to establish a resonance section table under the dual non-uniform system. The solving module 500 is configured to solve the subgroup parameters using the Pade approximation according to the resonance table.
Further, in an embodiment of the present invention, the background section module specifically includes: the mixing unit is used for mixing the dispersed particle fuel in each double non-uniform problem variant with the matrix to obtain the nuclear density after the fuel area is homogenized; and the solving unit is used for calling a characteristic line neutron transport calculation method to solve a single group fixed source transport equation of the nuclear density to obtain the standard flux of the fuel area so as to calculate the background section of the nuclide l in the g group fuel area.
Further, in one embodiment of the present invention, the homogenized nuclear density of the fuel region is:
Figure BDA0003383907570000091
wherein N isiso,j、
Figure BDA0003383907570000092
Density, V, before and after homogenization of nuclide iso in material number jj、VhomThe volume of material j and fuel zone respectively.
Further, in one embodiment of the present invention, the single group fixed source transport equation is:
Figure BDA0003383907570000101
wherein g is an energy group number, omega is an angle variable of neutron flux density,
Figure BDA0003383907570000102
neutron flux density per space position for g group
Figure BDA0003383907570000103
As a function of the change in the amount of the change,
Figure BDA0003383907570000104
for Laplace gradient operators, λ Σp,gMultiplying the intermediate resonance factor lambda by the g group macroscopic potential scattering cross section sigmap,gFixed source item in the characteristic line neutron transport method; with respect to the fuel region in the homogenization problem,
Figure BDA0003383907570000105
the macroscopic absorption cross section of the g group is equal to the statistical microscopic effective resonance cross section of the absorption reaction of the nuclide l multiplied by the nuclear density of the nuclide l; for the non-fuel regions of the homogenization problem,
Figure BDA0003383907570000106
equal to 0.
Further, in one embodiment of the present invention, utilization in the solution unit
Figure BDA0003383907570000107
Calculating a background cross-section, where σb,g、σa,gAnd
Figure BDA0003383907570000108
background section, microscopic absorption section and standard flux, σ, of the g groupa,gIs the g-group microscopic effective resonance cross section of the absorption reaction of the nuclide l.
It should be noted that the foregoing explanation of the embodiment of the subgroup parameter calculation method for dispersed particulate fuel is also applicable to the system of this embodiment, and will not be described herein again.
According to the subgroup parameter calculation system suitable for the dispersed granular fuel provided by the embodiment of the invention, the resonance database suitable for the double non-uniform system is generated by establishing a plurality of double non-uniform systems, the neutron calculation of the dispersed granular fuel can be realized by adopting the mixing simplified modeling of the dispersed granular fuel in a mode of replacing an original multi-group nuclear database without changing the existing fuel assembly program using a subgroup method, and the neutron calculation difficulty of the dispersed granular fuel is simplified; meanwhile, the radial power distribution of the fuel ball, the rod and the plate can be calculated without changing the shape and the size of the fuel element in the calculation process.
Furthermore, the terms "first", "second" and "first" are used for descriptive purposes only and are not to be construed as indicating or implying relative importance or implicitly indicating the number of technical features indicated. Thus, a feature defined as "first" or "second" may explicitly or implicitly include at least one such feature. In the description of the present invention, "a plurality" means at least two, e.g., two, three, etc., unless specifically limited otherwise.
In the description herein, references to the description of the term "one embodiment," "some embodiments," "an example," "a specific example," or "some examples," etc., mean that a particular feature, structure, material, or characteristic described in connection with the embodiment or example is included in at least one embodiment or example of the invention. In this specification, the schematic representations of the terms used above are not necessarily intended to refer to the same embodiment or example. Furthermore, the particular features, structures, materials, or characteristics described may be combined in any suitable manner in any one or more embodiments or examples. Furthermore, various embodiments or examples and features of different embodiments or examples described in this specification can be combined and combined by one skilled in the art without contradiction.
Although embodiments of the present invention have been shown and described above, it is understood that the above embodiments are exemplary and should not be construed as limiting the present invention, and that variations, modifications, substitutions and alterations can be made to the above embodiments by those of ordinary skill in the art within the scope of the present invention.

Claims (10)

1. A subgroup parameter calculation method for a dispersed particulate fuel, comprising the steps of:
step S1, establishing a preset number of double non-uniform problem variants by adjusting the density of fuel and the density of moderator in the preset fuel assembly;
step S2, each energy group of each double non-uniform problem variant is processed through a Monte Carlo method, and a microscopic effective resonance section of each type of nuclear reaction with a resonance effect is obtained;
step S3, mixing the dispersed particle fuel in each double non-uniform problem variant with a matrix, sequentially calculating all standard fluxes after fuel area homogenization through a characteristic line method, and further sequentially calculating all corresponding background sections;
step S4, the microscopic effective resonance sections and the background sections are in one-to-one correspondence to establish a resonance section table under a dual non-uniform system;
and step S5, solving subgroup parameters by using Pade approximation according to the resonance section table.
2. The method of claim 1, wherein said step S3 specifically includes:
step S301, mixing the dispersed particle fuel in each double non-uniform problem variant with a matrix to obtain the nuclear density after the fuel area is homogenized;
and S302, calling a characteristic line neutron transport calculation method to solve a single group fixed source transport equation of the nuclear density to obtain the standard flux of the fuel area so as to calculate the background section of the g group fuel area relative to the nuclide l.
3. A subgroup parameter calculation method for a dispersed particulate fuel as claimed in claim 2, wherein said homogenized nuclear density of said fuel region is:
Figure FDA0003383907560000011
wherein N isiso,j、
Figure FDA0003383907560000012
Density, V, before and after homogenization of nuclide iso in material number jj、VhomThe volume of material j and fuel zone respectively.
4. A subgroup parameter calculation method for a dispersed particulate fuel as claimed in claim 2 wherein said single group stationary source transport equation is:
Figure FDA0003383907560000013
wherein g is an energy group number, omega is an angle variable of neutron flux density,
Figure FDA0003383907560000014
neutron flux density per space position for g group
Figure FDA0003383907560000015
As a function of the change in the amount of the change,
Figure FDA0003383907560000016
for Laplace gradient operators, λ Σp,gMultiplying the intermediate resonance factor lambda by the g group macroscopic potential scattering cross section sigmap,gFixed source item in the characteristic line neutron transport method; with respect to the fuel region in the homogenization problem,
Figure FDA0003383907560000021
the macroscopic absorption cross section of the g group is equal to the statistical microscopic effective resonance cross section of the absorption reaction of the nuclide l multiplied by the nuclear density of the nuclide l; for the non-fuel regions of the homogenization problem,
Figure FDA0003383907560000022
equal to 0.
5. A subgroup parameter calculation method applicable to dispersed particulate fuel as claimed in claim 2, wherein said step S302 uses
Figure FDA0003383907560000023
Calculating the background section, whereinb,g、σa,gAnd
Figure FDA0003383907560000024
background section, microscopic absorption section and standard flux, σ, of the g groupa,gIs the g-group microscopic effective resonance cross section of the absorption reaction of the nuclide l.
6. A subgroup parameter calculation system adapted for use with a dispersed particulate fuel, comprising:
the variant building module is used for building a preset number of double non-uniform problem variants by adjusting the fuel density and the moderator density in a preset fuel assembly;
the resonance section module is used for processing each energy group of each double non-uniform problem variant through a Monte Carlo method to obtain a microscopic effective resonance section of each type of nuclear reaction with resonance effect;
the background section module is used for mixing the dispersed particle fuel in each double non-uniform problem variant with a matrix, sequentially calculating all standard fluxes after fuel areas are homogenized by a characteristic line method, and further sequentially calculating all corresponding background sections;
the resonance section table building module is used for corresponding the microscopic effective resonance sections and the background sections one by one so as to build a resonance section table under a dual non-uniform system;
and the solving module is used for solving the subgroup parameters by using the Pade approximation according to the resonance section table.
7. A subgroup parameter calculation system adapted for use with a dispersed particulate fuel as claimed in claim 6, wherein said background section module comprises in particular:
the mixing unit is used for mixing the dispersed particle fuel in each double non-uniform problem variant with the matrix to obtain the nuclear density after the fuel area is homogenized;
and the solving unit is used for calling a characteristic line neutron transport calculation method to solve the single group fixed source transport equation of the nuclear density to obtain the standard flux of the fuel area so as to calculate the background section of the nuclide l in the g group fuel area.
8. A subgroup parameter calculation system adapted for use with a dispersed particulate fuel as recited in claim 7, wherein said homogenized nuclear density of said fuel region is:
Figure FDA0003383907560000025
wherein N isiso,j、
Figure FDA0003383907560000026
Density, V, before and after homogenization of nuclide iso in material number jj、VhomThe volume of material j and fuel zone respectively.
9. A subgroup parameter calculation system adapted for use with a dispersed particulate fuel as recited in claim 7, wherein said single group stationary source transport equation is:
Figure FDA0003383907560000031
wherein g is an energy group number, omega is an angle variable of neutron flux density,
Figure FDA0003383907560000032
neutron flux density per space position for g group
Figure FDA0003383907560000033
As a function of the change in the amount of the change,
Figure FDA0003383907560000034
for Laplace gradient operators, λ Σp,gMultiplying the intermediate resonance factor lambda by the g group macroscopic potential scattering cross section sigmap,gFixed source item in the characteristic line neutron transport method; with respect to the fuel region in the homogenization problem,
Figure FDA0003383907560000035
the macroscopic absorption cross section of the g group is equal to the statistical microscopic effective resonance cross section of the absorption reaction of the nuclide l multiplied by the nuclear density of the nuclide l; for the non-fuel regions of the homogenization problem,
Figure FDA0003383907560000036
equal to 0.
10. A subgroup parameter calculation method for dispersed particulate fuel as claimed in claim 7, wherein said solving means utilizes
Figure FDA0003383907560000037
Calculating the background section, whereinb,g、σa,gAnd
Figure FDA0003383907560000038
background section, microscopic absorption section and standard flux, σ, of the g groupa,gIs the g-group microscopic effective resonance cross section of the absorption reaction of the nuclide l.
CN202111445233.9A 2021-11-30 2021-11-30 Subgroup parameter calculation method and system suitable for dispersed granular fuel Pending CN114203270A (en)

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CN115455796A (en) * 2022-09-22 2022-12-09 中国核动力研究设计院 Transport equivalence method for dispersed boron particles, electronic device and storage medium
CN115565617A (en) * 2022-09-16 2023-01-03 上海交通大学 Method for rapidly evaluating production efficiency of plutonium isotope based on energy spectrum environment

Cited By (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN115565617A (en) * 2022-09-16 2023-01-03 上海交通大学 Method for rapidly evaluating production efficiency of plutonium isotope based on energy spectrum environment
CN115565617B (en) * 2022-09-16 2024-02-13 上海交通大学 Method for rapidly evaluating production efficiency of plutonium isotope based on energy spectrum environment
CN115455796A (en) * 2022-09-22 2022-12-09 中国核动力研究设计院 Transport equivalence method for dispersed boron particles, electronic device and storage medium

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