CN115455796B - Transport equivalent method for dispersed boron particles, electronic device and storage medium - Google Patents

Transport equivalent method for dispersed boron particles, electronic device and storage medium Download PDF

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CN115455796B
CN115455796B CN202211157746.4A CN202211157746A CN115455796B CN 115455796 B CN115455796 B CN 115455796B CN 202211157746 A CN202211157746 A CN 202211157746A CN 115455796 B CN115455796 B CN 115455796B
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赵文博
蔡云
彭星杰
刘勇
倪东洋
樊兴
徐飞
娄磊
陈长
于颖锐
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Abstract

The application discloses a transportation equivalent method, electronic equipment and storage medium for dispersing boron particles, which specifically comprises the following steps: according to the spherical characteristics of particles in a fuel core, a plurality of particle type models of different types are established, all particle type models of all energy groups are circulated, flux adverse factors of each energy group are calculated, a microcosmic transport section is corrected according to the flux adverse factors, a plurality of groups of macroscopic transport sections in a core matrix are calculated, a plurality of groups of Boltzmann neutron transport equations are solved, a plurality of groups of neutron fluxes of the matrix are obtained, the absorption reaction rate is updated, and burnup calculation is carried out on each particle type model to obtain the nuclear density in the particles at the next moment. According to the scheme, 2 independent steps are added before the original transportation module is executed, 1 step is added before the fuel consumption module is calculated, and the whole program is less modified. The scheme has higher calculation precision, and can effectively describe the self-shielding effect of the boron particles.

Description

Transport equivalent method for dispersed boron particles, electronic device and storage medium
Technical Field
The application relates to the field of transportation equivalence in nuclear reactor physical numerical computation, in particular to a transportation equivalence method, electronic equipment and storage medium for dispersed boron particles.
Background
Boron is a species widely used for burnable poisons. For dispersive fuel elements, boron is often dispersed in the core in the form of compound particles. When the geometry of the boron particles is greater than or comparable to the thermal neutron free path (-0.1 mm), the particles will have a spatially self-shielding, i.e., the outer region flux level is higher than the inner region flux level.
If particles are directly averaged into the core during reactor neutron calculation, the absorption of boron can be overestimated, which affects the calculation accuracy of the reactivity release during burnup, and thus affects the reactivity control in the design. The larger the radius of the particles, the more pronounced the self-shielding phenomenon.
Neutron computing methods in the reactor core are divided into two main categories, namely a definite theory method and a probabilistic theory method. The probability theory method has strong geometric processing capability, but has large consumption of computing resources. The deterministic method, especially the two-step method widely used for design calculation, adopts a large amount of space union regions and energy union groups, performs more approximation on geometric modeling, reduces the degree of freedom of calculation and has high calculation efficiency.
For the deterministic method, there is a great difficulty if the randomly distributed particles are directly modeled explicitly in the calculation. Firstly, the particles themselves present complex geometries, which can be approximated by spheres, which can in principle be overcome; secondly, the randomness of the distribution exists, and the specific distribution of the particles inside each product which is specifically and practically manufactured is unknown, and the probability of the distribution of the particles in the matrix is generally assumed to meet the uniform distribution and can be overcome in principle; again, the storage bottleneck caused by the burnup process, the small particle size, and thus the huge number, even with one fuel assembly, is not affordable by the cost of calculating burnup and storing each particle independently, with the most prominent being the storage bottleneck. One way is to merge the particles, without using each particle as a separate burnup zone. Finally, the geometric handling capability of transport calculations is generally not provided with the computational capability of handling spherical geometries inside the cell. Adding such direct geometric processing capability requires a significant development effort.
Disclosure of Invention
The technical problem to be solved by the application is as follows: in the prior art, the transportation calculation of the dispersed particles has large consumption of calculation resources, and the method aims to provide a transportation equivalent method, electronic equipment and storage medium for the dispersed boron particles, adopts a one-dimensional sphere model to directly carry out numerical solution, does not need to additionally call a reference program (such as a Monte Carlo program) for explicit modeling, adopts a plurality of groups of self-shielding factors to correct a plurality of groups of cross sections, can process the space self-shielding effect of the dispersed boron particles which are randomly distributed, has a calculation result consistent with the explicit modeling result, and has small change on the existing neutron transportation calculation program.
The application is realized by the following technical scheme:
in a first aspect, a transport equivalent method for dispersing boron particles, comprising: step 1: according to the spherical characteristics of particles in the fuel core, establishing a plurality of different particle type models; step 2: cycling all particle type models of all energy groups, and calculating flux adverse factors of each energy group; step 3: cycling all particle type models of all energy groups and correcting microscopic transport sections according to the flux adverse factors; step 4: calculating a plurality of groups of macroscopic transport sections in the reactor core matrix according to the corrected microscopic transport sections; step 5: solving a multi-group Boltzmann neutron transport equation according to a multi-group macroscopic transport section in the reactor core matrix to obtain multi-group neutron flux of the matrix; step 6: and updating the absorption reaction rate according to the neutron flux of the plurality of groups, and executing burnup calculation on each particle type model to obtain the nuclear density in the particles at the next moment.
Further, calculating the flux adverse factor of each energy group is to calculate the volume average flux of the particle type model, and the volume average flux of the particle type model is obtained by solving a neutron transport equation of single-group one-dimensional sphere geometry.
Further, in the calculation of the flux adverse factors, scattering between energy groups and scattering from the energy groups to the energy groups are ignored, a single group one-dimensional sphere geometric transportation equation is solved, and a boundary condition is selected as a cosine incidence boundary condition of neutron flux on the surface of the particle sphere.
Further, solving the single group one-dimensional sphere geometric transportation equation by adopting a discrete longitudinal standard method.
Further, the single group one-dimensional sphere geometric transportation equation is as follows:
Figure BDA0003859563950000031
wherein the variable r is the space radius coordinate under the spherical coordinate system, the variable mu is the cosine of the included angle between the r direction and the neutron flight direction, the energy variable adopts the energy group approximation as the subscript g, Σ g (r) is a macroscopic cross section and subscript a is an absorption cross section.
Further, the step 1 includes: the boron particles are approximated as spheres and all particle type models of different types are built according to the radius, volume fraction and density of the boron particles.
Further, when the probability distribution exists in the radius of the boron particles, the particles with a plurality of radii are adopted to respectively model according to the volume proportions corresponding to the probability distribution, and are used as different particle materials dispersed in the same core body to be processed.
Further, the method further comprises the following steps: step 7: normal burnup calculations are performed on the burnup zone of the non-substrate.
In a second aspect, an electronic device comprises a memory and a processor, the memory having stored thereon a computer program which, when executed by the processor, performs a transport equivalent method for diffusing boron particles as described above.
In a third aspect, a storage medium storing a computer program executable by one or more processors is provided for implementing a transport equivalent method for dispersing boron particles as described above.
Compared with the prior art, the application has the following advantages and beneficial effects:
according to the method, the space self-shielding effect of the particle boron can be effectively depicted without explicit modeling treatment, the calculation efficiency is high, the calculation cost of an original program is not increased, and meanwhile, the method is convenient to implement in a specific transportation calculation program and has definite engineering practical value.
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In order to more clearly illustrate the technical solutions of the exemplary embodiments of the present application, the drawings that are needed in the examples will be briefly described below, it being understood that the following drawings only illustrate some examples of the present application and therefore should not be considered as limiting the scope, and that other related drawings may be obtained according to these drawings without inventive effort for a person skilled in the art. In the drawings:
fig. 1 is a flow chart of a method of embodiment 1 of the present application.
Detailed Description
For the purpose of making the objects, technical solutions and advantages of the present application more apparent, the present application will be further described in detail with reference to the examples and the accompanying drawings, and the exemplary embodiments of the present application and the descriptions thereof are only for explaining the present application and are not limiting the present application.
Example 1
Embodiment 1 of the application is a transport equivalent method for dispersed boron particles, as shown in fig. 1, based on the method of embodiment 1, the spatial self-shielding effect of randomly distributed dispersed boron particles can be processed, the calculation result is consistent with the explicit modeling result, and meanwhile, the processing method has small change to the existing neutron transport calculation program. The specific method of example 1 of the present application is as follows:
step 1: assuming that the dispersed particles are spherical in shape, the distribution of the particles satisfies a random uniform distribution, specifying the radius, volume fraction, and density of the particles. When the particle radius has probability distribution, a plurality of radius particles are adopted to respectively model according to volume proportions corresponding to the probability distribution, and are treated as different particle materials dispersed in the same core body.
The particles are classified into different types according to radius, material density and the like, and the following is respectively implemented for each type of particles:
step 2: the boron particle flux adverse factor was calculated. The flux penalty factor is related to neutron incident energy. The energy variation is dispersed into small energy intervals (called energy clusters), i.e. continuous energy ranges [ E min ,E max ]Divided into intervals, each of which is called an energy group. The discrete division of energy according to energy groups is a theoretical neutron transport calculation standard solving mode.
Figure BDA0003859563950000051
Each energy cluster has a corresponding flux penalty factor. The neutron transport calculation program obtains the neutron flux change along with energy, space and angle, wherein the energy degree of freedom is also divided according to energy groups. The energy group division of the flux adverse factors is consistent with neutron transport procedures. In the calculation of flux adverse factors, scattering between energy groups and from the energy groups to the energy groups is ignored, a single group one-dimensional sphere geometric transportation equation is solved, and a boundary condition is selected as a cosine incidence boundary condition of neutron flux on the surface of the particle sphere. By solving the single group one-dimensional problem, the volume average flux of the particle sphere can be obtained.
The neutron transport equation of the single group one-dimensional sphere geometry to be solved is formed as follows:
Figure BDA0003859563950000052
wherein the variable r is the space radius coordinate under the spherical coordinate system, the variable mu is the cosine of the included angle between the r direction and the neutron flight direction, the energy variable adopts the energy group approximation as the subscript g, Σ g (r) is a macroscopic cross section and subscript a is an absorption cross section.
The outer boundary condition of the sphere is a neutron cosine incidence condition.
The equation solving method can adopt a mature collision probability method or a discrete longitudinal label method. After obtaining neutron flux density, further obtaining volume average flux, meeting the following requirements
φ vol,g ＝∫φ g (r,μ)r 2 drdμ/(2∫r 2 dr)
Thereby obtaining the flux adverse factor f of the single group, meeting the following conditions
Figure BDA0003859563950000053
Wherein phi is out,g Is the average flux outside the sphere.
The volume average flux is the average value of flux distribution in the sphere, and meets the ≡phi g dV。
The flux penalty factor for the boron particles referred to in example 1 of the present application is the same amount as the flux penalty factor f for the single population obtained in equation (1).
And (5) circulating all types and all energy groups to finish the flux adverse factor calculation of each energy group.
Step 3: the microscopic transport cross section is modified with flux penalty factors,
σ tr,g ＝f g σ tr,g,old (2)
wherein sigma is a microscopic cross section and subscript tr is a transport cross section. Subscript old is the original cross-section prior to update.
And (3) circulating all types and all energy groups to finish the correction of the microscopic transport section as shown in the formula (2). From the flux penalty factor for each energy group, a modification of the cross section of the particle is obtained according to equation (2).
Step 4: in the reactor core matrix, calculating a plurality of groups of macroscopic transport sections of the matrix, as shown in the following formula,
Figure BDA0003859563950000061
where N is the concentration of the species (including all species in the matrix, various types of boron, and other matrix species), the subscript i traverses all species. The remaining scattering cross section, fission cross section, and the like are identical to the calculation method when the boron particles are not considered (identical to the treatment method in which the boron particle correction is not performed).
Step 5: core or coreAnd calculating neutron transport in the module. Solving the multi-group Boltzmann neutron transport equation by using the section calculated in the formula (3) without any change according to the original calculation mode to obtain the multi-group neutron flux of the matrix
Figure BDA0003859563950000062
Figure BDA0003859563950000063
Wherein the space variable
Figure BDA0003859563950000064
The energy variable adopts the energy group approximation as the subscript g, and the angle variable +.>
Figure BDA0003859563950000065
The subscript tr is a transport section, s is a scattering section, and f is a fissile section.
Figure BDA0003859563950000066
For energy group g' to energy group g, the incident neutron flight angle +.>
Figure BDA0003859563950000067
Angle of flight to exit neutrons
Figure BDA0003859563950000068
Is used for the differential scattering cross section of the optical fiber.
Step 6: and (5) calculating the burnup of the particles. The purpose of the burnup calculation is to obtain the change of the relevant nuclide concentration in a certain time by using the reaction rate obtained by the transportation calculation. Before performing the conventional burnup calculation, this example 1 updates the absorption reaction rate by using the flux penalty factor f, specifically, see the formulas (5) and (6), and performs the burnup calculation for each type of particle, resulting in the nuclear density in the particle at the next time.
σ ab (E)＝f(E)σ ab,old (E) (5)
R ab ＝∫φ(E)σ ab (E)dE (6)
Subscript ab denotes absorption cross section, R is absorption reactivity, and phi (E) is core flux.
Step 7: normal burnup calculations are performed on the burnup zone of the non-substrate.
The difference in reactivity between the heterogeneous and homogeneous blend calculations is selected as a parameter for measuring the self-shielding effect of the boron particles. The reference value is calculated by using a Monte Carlo program RMC. The numerical calculation result shows that the method has higher calculation precision and can effectively describe the self-shielding effect of the boron particles.
The normal program calculation flow in the prior art is as follows: and 5, step 7, finishing the calculation of the fuel consumption of one-time transportation, and then advancing the calculation according to a set time step (fuel consumption step). In the implementation process of embodiment 1 of the present application, step 3 and step 4 are added before the execution of the original transportation module, and these two steps are very independent, and step 6 is added before the calculation of the burnup module, so that the modification of the program as a whole is less.
Example 2
The embodiment 2 is based on the embodiment 1, the specific implementation steps are given based on the component program MANTA, but the transportation method and the transportation program for specific application are not limited to the MANTA program, and can be applied to various deterministic multi-group neutron transportation programs. The specific method of this example 2 is as follows:
step 1: the radius, volume fraction, density of the particles are specified. Particles of different radii, different materials are modeled by different particle types.
After the MANTA procedure completes the resonance process, steps 2 and 3 are performed by cycling through all energy groups and particle types before preparing the component transport calculations.
Step 2: and solving a geometric transportation equation of a single group of one-dimensional spheres by adopting a discrete longitudinal standard method, wherein the boundary condition of the outer surface is a cosine incidence boundary condition, so as to obtain the volume average flux of the spheres, and the flux adverse factor f of the group is obtained, thereby satisfying the formula (1).
Step 3: and correcting the multi-group transport cross sections by using the flux adverse factors, and satisfying the formula (2).
Step 4: and calculating a plurality of groups of macroscopic transport cross sections of each grid of the MANTA program, wherein the macroscopic transport cross sections are shown in a formula (3).
Step 5: the MANTA program performs a multi-cluster transport calculation to obtain a multi-cluster flux distribution for the component.
Step 6: the flux-negative factor is used to update the microscopic absorption reactivity of the particles as shown in formula (5). The MANTA program performs the burnup calculations to complete a burnup step of the component calculations.
In order to verify the calculation accuracy of this embodiment 2, the problem of single cell and 3×3 cell was explicitly modeled using the Monte Carlo program RMC as a reference. The scattered boron particles are randomly generated, and the calculation accuracy of the scheme is verified by comparing the reactivity deviation of the scattered boron particles and the uniform mixing under two modes.
Uniformly mixing: irrespective of the geometry of the particles or the infinitely small radius of the particles, the material in the particles is uniformly distributed in the matrix.
For comparison, two types of dispersion particle radii, 107 microns and 203 microns, respectively, were chosen. Meanwhile, the boron loading of the two calculation objects is the same, but the number of particles is different. The same example is the homogeneous blending example for both as a control. The fuel cell has a radius of 0.2425cm and a height of 0.5cm. 3 examples were constructed as follows:
1) Example 1 (107), single cell example. The particles were randomly and uniformly distributed with a radius of 107 microns and a number of particles of 100.
2) Example 2 (203), single cell example. The particles were randomly and evenly distributed with a radius of 203 microns and a number of 15 particles.
3) Example 3 (3 x 3), 3 x 3 cell example, cell arrangement is as follows,
Figure BDA0003859563950000081
Figure BDA0003859563950000091
wherein cell 1 is the cell of example 1, cell 2 is the cell of example 2, and cell 3 is the uniformly mixed cell. The results of the calculations are shown in the following table.
The reactivity comparison under the three calculation examples shown in the following table shows that the self-shielding effect of the embodiment 2 is consistent with the reference result both at the initial time and in the burnup process, and the calculation accuracy is high.
Figure BDA0003859563950000092
In this embodiment 2, taking the component calculation program MANTA as an example, the scheme of the application is not limited to the MANTA program, and is applicable to neutron transport calculation programs of the deterministic method, such as CASMO, DRAGON, SHARK.
Example 3
Embodiment 3 provides an electronic device including a memory and a processor, the memory having stored thereon a computer program which, when executed by the processor, performs the transport equivalent method for dispersing boron particles as described in embodiment 1.
Example 4
Embodiment 4 may also provide a storage medium storing a computer program executable by one or more processors to implement the transportation equivalent method for dispersing boron particles as described in embodiment 1.
The foregoing embodiments have been provided for the purpose of illustrating the general principles of the present application, and are not meant to limit the scope of the invention, but to limit the scope of the invention.

Claims (10)

1. A transport equivalent method for dispersing boron particles, comprising:
step 1: according to the spherical characteristics of particles in the fuel core, establishing a plurality of different particle type models;
step 2: cycling all particle type models of all energy groups, and calculating flux adverse factors of each energy group;
step 3: cycling all particle type models of all energy groups and correcting microscopic transport sections according to the flux adverse factors;
step 4: calculating a plurality of groups of macroscopic transport sections in the reactor core matrix according to the corrected microscopic transport sections;
step 5: solving a multi-group Boltzmann neutron transport equation according to a multi-group macroscopic transport section in the reactor core matrix to obtain multi-group neutron flux of the matrix;
step 6: and updating the absorption reaction rate according to the neutron flux of the plurality of groups, and executing burnup calculation on each particle type model to obtain the nuclear density in the particles at the next moment.
2. The transportation equivalent method for dispersing boron particles according to claim 1, wherein the flux adverse factor of each energy group is calculated, namely the volume average flux of a particle type model is calculated, and the volume average flux of the particle type model is obtained by solving a neutron transportation equation of single-group one-dimensional sphere geometry.
3. The transportation equivalent method for dispersing boron particles according to claim 2, wherein in the calculation of flux adverse factors, scattering between energy groups and from the group to the group is ignored, a single group one-dimensional sphere geometric transportation equation is solved, and a boundary condition is selected as a cosine incidence boundary condition of neutron flux on the surface of the particle sphere.
4. A transport equivalent method for dispersed boron particles according to claim 3, characterized in that a discrete longitudinal scale method is used to solve the single group one-dimensional sphere geometric transport equation.
5. The transport equivalent method for dispersed boron particles of claim 4, wherein said single population of one-dimensional sphere geometry transport equations is as follows:
Figure FDA0003859563940000011
wherein the variable r is the space radius coordinate under the spherical coordinate system, the variable mu is the cosine of the included angle between the r direction and the neutron flight direction, the energy variable adopts the energy group approximation as the subscript g, Σ g (r) is a macroscopic cross section and subscript a is an absorption cross section.
6. The transport equivalent method for dispersed boron particles according to claim 1, wherein said step 1 comprises: the boron particles are approximated as spheres and all particle type models of different types are built according to the radius, volume fraction and density of the boron particles.
7. The transportation equivalent method for dispersing boron particles according to claim 6, wherein when the probability distribution exists in the radius of the boron particles, a plurality of radius particles are adopted to respectively model according to the volume share corresponding to the probability distribution, and the particles are treated as different particle materials dispersed in the same core body.
8. The transport equivalent method for dispersing boron particles of claim 1, further comprising:
step 7: normal burnup calculations are performed on the burnup zone of the non-substrate.
9. An electronic device comprising a memory and a processor, the memory having stored thereon a computer program which, when executed by the processor, performs a transport equivalent method for dispersing boron particles as claimed in any one of claims 1 to 8.
10. A storage medium storing a computer program executable by one or more processors for implementing a transport equivalent method for dispersing boron particles as claimed in any one of claims 1 to 8.
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Citations (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
EP3816842A1 (en) * 2019-10-30 2021-05-05 Dassault Systemes Simulia Corp. Computer system for simulating physical process using lattice boltzmann based scalar transport enforcing galilean invariance for scalar transport

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* Cited by examiner, † Cited by third party
Title
弥散燃料与可燃毒物双重非均匀性分析;娄磊;姚栋;柴晓明;于颖锐;彭星杰;刘琨;田超;王晨琳;刘勇;谢运利;;中国核电(第01期);全文 *
核数据对球床式高温气冷堆燃料元件中子学特性的影响;王黎东;郭炯;李富;Jason Hou;Kostadin Ivanov;;原子能科学技术(第09期);全文 *

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