CN107145721A - A kind of mixing computational methods for obtaining the few group cross-section parameter of fast neutron reactor - Google Patents
A kind of mixing computational methods for obtaining the few group cross-section parameter of fast neutron reactor Download PDFInfo
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Abstract
A kind of mixing computational methods for obtaining the few group cross-section parameter of fast neutron reactor, by the way that monte carlo method is combined with determining by method, the resonance effects of fast pile component is finely considered using monte carlo method, calculate the microcosmic total cross section of the accurate multigroup of each nucleic, fission cross section, elastic scattering cross-section, simultaneously using determining that opinion method solves each rank elastic scattering cross-section of fast pile component and each rank neutron flux square, and using this by multigroup cross section merger for group cross-section less；This mixing computational methods highly versatile, it is applied widely, the few group cross-section of high-precision fast pile component can be produced, accurate reliable cross section parameter is provided for the core design objective of reactor core.
Description
Technical field
It is a kind of acquisition fast neutron reaction the present invention relates to nuclear reactor design and nuclear reactor physical computing field
The mixing computational methods of the few group cross-section parameter of heap.
Background technology
In order to quickly and accurately solve reactor core neutronics parameter, the method based on " two-step method " turns into fast reactor engineering calculation
Main method.So-called " two-step method ", the first step is that heap in-core various assemblies material is modeled and calculated, and is obtained in component
Netron-flux density distribution so as to which merger goes out few group's homogenization group cross-section；Second step is to calculate obtained homogenization by back
Parameter carries out the solution of few group's neutron-transport equation to reactor core, obtains the physics such as reactor core Effective multiplication factor, core power distribution
Amount.
Monte carlo method can be divided into and determine to discuss method by calculating the method for few group cross-section use.Monte carlo method is
One kind is based on statistical method, and it carries out the simulation of neutronics characteristic by a large amount of particles of sampling, so as to obtain neutronics phase
The parameter of pass.It is determined that being the math equation group that neutronics characteristic is described by various theoretical or Numerical Methods Solves by method, lead to
Cross and obtain the parameter that solution of equations obtains neutronics correlation.
Because monte carlo method can use the database of Continuous Energy, at the same to geometrical model can with Accurate Model,
Therefore its result of calculation has high precision.In Monte Carlo calculations, the calculating of few group cross-section need to nuclear reaction rate, in
The information such as sub- flux density distribution are counted.But when these numerical value very littles, it can be difficult to obtain accurate result, such as high-order
Neutron flux square.Due to needing to use high-order neutron flux square in the calculating in high-order scattering section, and high-order neutron flux square
It is difficult to obtain again, therefore is introduced approximately in monte carlo method, it is using zeroth order neutron flux square to high-order scattering section
Calculated.And it is determined that by method, it is possible to use numerical calculations go out accurate high-order neutron flux square, it is to avoid number
The problem of result of calculation is inaccurate when being worth very little.But it is determined that by method, the accuracy of few group cross-section result of calculation is depended on
The result of calculation of multigroup cross section, and the calculating of multigroup cross section is related to resonance computational methods, therefore during using determining opinion method
Must be based on the higher resonance computational methods of precision.
Due to monte carlo method is used alone or determines have the shortcomings that to be difficult to overcome by method, it is therefore necessary to by this
Two methods are combined, and invent a kind of mixing computational methods of the few group cross-section of high-precision fast reactor.
The content of the invention
In order to overcome the shortcoming of monte carlo method and determination opinion method in the few group cross-section calculating of fast reactor, the present invention is in meter
Method is discussed using monte carlo method and determination simultaneously in the flow for calculating the few group cross-section of fast reactor, core is produced using monte carlo method
The microcosmic total cross section of element, fission cross section, elastic scattering cross-section, calculate other cross section informations by method using determination and neutron are logical
Metric density is distributed, and the few group cross-section of high-precision fast pile component is obtained with this.
In order to realize the above object the present invention takes following technical scheme to be practiced：
A kind of mixing computational methods for obtaining the few group cross-section parameter of fast neutron reactor, comprise the following steps：
Step 1：For any fast pile component of required calculating, the geological information and respective material component of the component are read
Nucleic information；
Step 2：The geological information and material component information read for step 1, reads each nucleic in all material
Microcosmic inelastic scattering cross section value, every time fission release neutron population, fission spectrum information；
Step 3：The information read according to step 1, by specifying the input information of Monte Carlo calculations to set up corresponding meter
Calculate model and calculated；Multigroup overall reaction rate of each nucleic in regional, fission reaction are counted in calculating process
Rate, elastic scattering reactivity and netron-flux density distribution；
Step 4：The multigroup overall reaction rate of each nucleic that is obtained using step 3, fission reaction rate, elastic scattering reactivity with
And netron-flux density distribution, try to achieve the microcosmic total cross section of multigroup of each nucleic, microscopic fission cross section and microcosmic using formula (1)
Elastic scattering cross-section；
In formula：
σx,g--- the microscopic cross of g groups, subscript x refers to microcosmic total cross section, fission cross section or elastic scattering cross-section
Rx,g--- the microreaction rate of g groups, it is anti-that subscript x refers to microcosmic overall reaction rate, fission reaction rate or elastic scattering
Should rate
φg--- the netron-flux density distribution of g groups
Step 5：:The probability of scattering of each nucleic group to group is calculated using formula (2), calculates what is obtained further according to step 4
The microcosmic elastic scattering cross-section of multigroup of each nucleic and the elastic scattering matrix that each nucleic is calculated using formula (3)；
In formula：
Fl(g → g') --- l ranks are by the g groups of probabilities of scattering for scattering to g ' groups
α——(A‐1)2/(A+1)2, A is the atomic mass of nucleic
μc--- the angle of scattering cosine value under geocentric coordinate system
f(E,μc) --- probability of scattering function
Pl(μs) --- on μsL rank Legnedre polynomials
μs--- the angle of scattering cosine value under laboratory coordinate
ΔEg--- the energy group interval of g groups
In formula：
--- l ranks are by the g groups of elastic scattering cross-sections for scattering to g ' groups
σs,g--- the elastic scattering cross-section of g groups
Step 6：According to the multigroup microscopic cross information of each nucleic obtained by step 2 to step 5, read with reference to step 1
The component geological information and material information taken, carries out the solution based on the neutron-transport equation for determining opinion method, obtains many with this
Each rank neutron flux square distribution of group；
Step 7：The multigroup that multigroup microscopic cross information and step 6 solution obtained based on step 2 to step 5 is obtained
Each rank neutron flux square distribution, and using formula (4), (5) to multigroup microscopic cross progress energy group, the merger in space, so that
To few group cross-section of component；
In formula：
σx,G,I--- G groups, belong to the microscopic cross in I areas, subscript x refers to microcosmic total cross section, microscopic fission cross section
σx,g,i--- g groups, belong to the microscopic cross in the i-thth area, subscript x refers to microcosmic total cross section, microscopic fission cross section
φg,i--- g groups, belong to the i-thth area netron-flux density distribution
Vi--- the volume of the i-th subregion
In formula：
--- l ranks scatter to G ' groups by G groups, belong to the microscopic scattering cross section in I areas
--- l ranks scatter to g ' groups by g groups, belong to the microscopic scattering cross section in the i-thth area
--- g groups, belong to the i-thth area l ranks netron-flux density distribution.
Compared with prior art, the present invention has the advantages that to protrude as follows：
1. in mixing computational methods, the microcosmic total cross section of multigroup, fission cross section, elastic scattering cross-section information are by Monte Carlo
Method is calculated, it is ensured that high accuracy；
2. in mixing computational methods, the elastic scattering matrix of each rank is calculated by determining by method, it is to avoid cover spy
It is approximate in the method for Carlow, it is ensured that scattering section precision；
3. in mixing computational methods, the neutron flux square of each rank is calculated by determining by method, it is to avoid Meng Teka
It is approximate in the method for Lip river；Carry out the merger in energy group and region to high-order scattering section using high-order neutron flux square simultaneously, obtain
More accurately lack group cross-section.
Brief description of the drawings
Fig. 1 is the mixed method flow chart that the few group cross-section of fast reactor is calculated.
Fig. 2 is the error of each 24 groups of microscopic cross of nucleic of uniformity problem.
Fig. 3 is the error of one-dimensional 24 groups of volumic total cross-sections of cylinder problem.
Embodiment
The present invention is described in further details with reference to the accompanying drawings and detailed description：
The present invention finely considers fast using monte carlo method by the way that monte carlo method is combined with determining by method
The resonance effects of pile component, calculates the microcosmic total cross section of the accurate multigroup of each nucleic, fission cross section, elastic scattering cross-section, simultaneously
Using determining that opinion method solves each rank elastic scattering cross-section of fast pile component and each rank neutron flux square, with this by multigroup cross section
Merger is few group cross-section, as shown in figure 1, the invention comprises the following steps：
Step 1：For any fast pile component of required calculating, the geological information and respective material component of the component are read
Nucleic information；
Step 2：The geological information and material component information read for step 1, reads each nucleic in all material
Microcosmic inelastic scattering cross section value, every time fission release neutron population, fission spectrum information；
Step 3：The information read according to step 1, by specifying the input information of Monte Carlo calculations to set up corresponding meter
Calculate model and calculated；Multigroup overall reaction rate of each nucleic in regional, fission reaction are counted in calculating process
Rate, elastic scattering reactivity and netron-flux density distribution；
Step 4：The multigroup overall reaction rate of each nucleic that is obtained using step 3, fission reaction rate, elastic scattering reactivity with
And netron-flux density distribution, try to achieve the microcosmic total cross section of multigroup of each nucleic, microscopic fission cross section and microcosmic using formula (1)
Elastic scattering cross-section；
In formula：
σx,g--- the microscopic cross of g groups, subscript x refers to microcosmic total cross section, microscopic fission cross section or microcosmic elastic scattering
Section
Rx,g--- the microreaction rate of g groups, subscript x refers to microcosmic overall reaction rate, microcosmic fission reaction rate or microcosmic bullet
Property scattering reaction rate
φg--- the netron-flux density distribution of g groups
Step 5：:The probability of scattering of each nucleic group to group is calculated using formula (2), calculates what is obtained further according to step 4
The microcosmic elastic scattering cross-section of multigroup of each nucleic and the elastic scattering matrix that each nucleic is calculated using formula (3)；
In formula：
Fl(g → g') --- l ranks are by the g groups of probabilities of scattering for scattering to g ' groups
α——(A‐1)2/(A+1)2, A is the atomic mass of nucleic
μc--- the angle of scattering cosine value under geocentric coordinate system
f(E,μc) --- probability of scattering function
Pl(μs) --- on μsL rank Legnedre polynomials
μs--- the angle of scattering cosine value under laboratory coordinate
ΔEg--- the energy group interval of g groups
In formula：
--- l ranks are by the g groups of elastic scattering cross-sections for scattering to g ' groups
σs,g--- the elastic scattering cross-section of g groups
Step 6：According to the multigroup microscopic cross information of each nucleic obtained by step 2 to step 5, read with reference to step 1
The component geological information and material information taken, carries out the solution based on the neutron-transport equation for determining opinion method, obtains many with this
Each rank neutron flux square distribution of group；
Step 7：The multigroup that multigroup microscopic cross information and step 6 solution obtained based on step 2 to step 5 is obtained
Each rank neutron flux square distribution, and using formula (4), (5) to multigroup microscopic cross progress energy group, the merger in space, so that
To few group cross-section of component.
In formula：
σx,G,I--- G groups, belong to the microscopic cross in I areas, subscript x refers to microcosmic total cross section, microscopic fission cross section
σx,g,i--- g groups, belong to the microscopic cross in the i-thth area, subscript x refers to microcosmic total cross section, microscopic fission cross section
φg,i--- g groups, belong to the i-thth area netron-flux density distribution
Vi--- the volume of the i-th subregion
In formula：
--- l ranks scatter to G ' groups by G groups, belong to the microscopic scattering cross section in I areas
--- l ranks scatter to g ' groups by g groups, belong to the microscopic scattering cross section in the i-thth area
--- g groups, belong to the i-thth area l ranks netron-flux density distribution
In the present invention, the geometry and material information of any fast pile component are read in by step 1, the letter read according to step 1
During breath must read the microcosmic inelastic scattering cross section value of each nucleic by step 2 in Multi-group data storehouse, fission discharges every time
Subnumber, fission spectrum information.Above- mentioned information is included suitable for the Multi-group data storehouse that reactor physics are calculated, the present invention is to multigroup number
Do not limited according to the selection in storehouse.
In the Monte Carlo calculations of step 3, statistical regions must be made with definition, that is, set the space of statistical regions to be included
Scope, the name information of the bound residing for energy interval and specific statistics nucleic, can cover special using arbitrary during statistics
Carlow statistical method, such as collision count method or track lenth counting method, the present invention or not to reactivity and the statistics of flux
The limitation of method.
Solving for each rank neutron flux square distribution can use the neutron-transport equation solution side of different spaces dimension in step 6
Method, such as collision probability method, Discrete-ordinates method, characteristic line method, Spatial Dimension include uniformity problem, one-dimensional problem.This hair
The bright solution procedure for the distribution of each rank neutron flux square and few group cross-section merger is not by neutron-transport equation method for solving
Limitation.
For checking effectiveness of the invention, Fig. 2 illustrates the uniform fast reactor component problem obtained using present invention calculating
24 microcosmic total cross sections of each nucleic and the relative error with reference to solution.Result of calculation shows, utilizes lacking that present invention calculating is obtained
Group cross-section with reference to compared with solution, few group cross-section has high precision, each nucleic each can group cross-section error 1% with
It is interior.The ability of authentic component is calculated for the checking present invention, for the fast pile component of one-dimensional cylindrical geometry, Monte Carlo side is respectively adopted
Method, determination opinion method and mixing computational methods are calculated, and using monte carlo method as with reference to solving, compare other two methods
24 groups of few group's volumic total cross-sections error.As shown in figure 3, compared to determining to discuss method, mixing section produced by computational methods
Face has higher precision, and cross-section error is within 1%.
Utilize accurate group cross-section less, it is possible to carry out the various Neutronics calculations of fast reactor reactor core, such as stable state is calculated, burnup
Calculating, transient state calculating etc..The present invention can obtain the few group cross-section of fast pile component of degree of precision, can be applied to actual engineering meter
Among calculating.
Claims (1)
1. a kind of mixing computational methods for obtaining the few group cross-section parameter of fast neutron reactor, it is characterised in that：Comprise the following steps：
Step 1：For any fast pile component of required calculating, the geological information of the component and the core of respective material component are read
Prime information；
Step 2：Each nucleic is microcosmic in the geological information and material component information read for step 1, reading all material
Inelastic scattering cross section value, every time fission release neutron population, fission spectrum information；
Step 3：The information read according to step 1, by specifying the input information of Monte Carlo calculations to set up corresponding calculating mould
Type is simultaneously calculated；Multigroup overall reaction rate of each nucleic in regional, fission reaction rate, bullet are counted in calculating process
Property scattering reaction rate and netron-flux density distribution；
Step 4：The multigroup overall reaction rate of each nucleic that is obtained using step 3, fission reaction rate, elastic scattering reactivity and in
Sub- flux density distribution, the microcosmic total cross section of multigroup, microscopic fission cross section and the microcosmic elasticity of each nucleic are tried to achieve using formula (1)
Scattering section；
In formula：
σx,g--- the microscopic cross of g groups, subscript x refers to microcosmic total cross section, microscopic fission cross section or microcosmic elastic scattering cross-section
Rx,g--- the microreaction rate of g groups, subscript x refers to microcosmic overall reaction rate, microcosmic fission reaction rate or microcosmic elasticity and dissipated
Penetrate reactivity
φg--- the netron-flux density distribution of g groups
Step 5：The probability of scattering of each nucleic group to group is calculated using formula (2), obtained each nucleic is calculated further according to step 4
The microcosmic elastic scattering cross-section of multigroup and the elastic scattering matrix of each nucleic is calculated using formula (3)；
In formula：
Fl(g → g') --- l ranks are by the g groups of probabilities of scattering for scattering to g ' groups
α——(A‐1)2/(A+1)2, A is the atomic mass of nucleic
μc--- the angle of scattering cosine value under geocentric coordinate system
f(E,μc) --- probability of scattering function
Pl(μs) --- on μsL rank Legnedre polynomials
μs--- the angle of scattering cosine value under laboratory coordinate
ΔEg--- the energy group interval of g groups
In formula：
--- l ranks are by the g groups of elastic scattering cross-sections for scattering to g ' groups
σs,g--- the elastic scattering cross-section of g groups
Step 6：According to the multigroup microscopic cross information of each nucleic obtained by step 2 to step 5, read with reference to step 1
Component geological information and material information, carry out the solution based on the neutron-transport equation for determining opinion method, multigroup are obtained with this
Each rank neutron flux square distribution；
Step 7：The multigroup microscopic cross information and step 6 obtained based on step 2 to step 5 solves each rank of obtained multigroup
Neutron flux square is distributed, and using formula (4), (5) to multigroup microscopic cross progress energy group, the merger in space, so as to obtain group
Few group cross-section of part；
In formula：
σx,G,I--- G groups, belong to the microscopic cross in I areas, subscript x refers to microcosmic total cross section, microscopic fission cross section
σx,g,i--- g groups, belong to the microscopic cross in the i-thth area, subscript x refers to microcosmic total cross section, microscopic fission cross section
φg,i--- g groups, belong to the i-thth area netron-flux density distribution
Vi--- the volume of the i-th subregion
In formula：
--- l ranks scatter to G ' groups by G groups, belong to the microscopic scattering cross section in I areas
--- l ranks scatter to g ' groups by g groups, belong to the microscopic scattering cross section in the i-thth area
--- g groups, belong to the i-thth area l ranks netron-flux density distribution.
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Cited By (6)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN108733903A (en) * | 2018-05-08 | 2018-11-02 | 西安交通大学 | A kind of neutron transport method for numerical simulation carrying out personalisation process by neutron energy |
| CN109493924A (en) * | 2018-11-13 | 2019-03-19 | 西安交通大学 | A method of obtaining the effective multigroup cross section of FCM fuel |
| CN110704806A (en) * | 2019-09-19 | 2020-01-17 | 西安交通大学 | Rapid online calculation method for one-dimensional cylindrical geometric collision probability |
| CN112966428A (en) * | 2021-03-31 | 2021-06-15 | 清华大学 | Cross-section treatment system for reactor core |
| CN113868591A (en) * | 2021-09-28 | 2021-12-31 | 西安交通大学 | Method for obtaining high-precision total reaction section of indistinguishable resonance region |
| CN117150829A (en) * | 2023-10-31 | 2023-12-01 | 西安交通大学 | Method for generating stack-used few-group cross-section energy group structure for neutron analysis of intermediate energy spectrum |
Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN103116667A (en) * | 2013-01-24 | 2013-05-22 | 西安交通大学 | Method of obtaining fusion reactor experimental covering module neutronics parameters |
| CN106128518A (en) * | 2016-06-24 | 2016-11-16 | 西安交通大学 | A kind of method obtaining the few group cross-section of high-precision fast neutron reaction pile component |
-
2017
- 2017-04-20 CN CN201710269812.XA patent/CN107145721B/en active Active
Patent Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN103116667A (en) * | 2013-01-24 | 2013-05-22 | 西安交通大学 | Method of obtaining fusion reactor experimental covering module neutronics parameters |
| CN106128518A (en) * | 2016-06-24 | 2016-11-16 | 西安交通大学 | A kind of method obtaining the few group cross-section of high-precision fast neutron reaction pile component |
Non-Patent Citations (7)
| Title |
|---|
| DU XIANAN等: ""THE GENERATION OF FEW-GROUP CONSTANTS FOR FAST REACTOR ANALYSIS"", 《24TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING》 * |
| FAEZEH ABBASI等: ""Monte Carlo Based Modeling and Simulation of Neutron Flux Distribution and Activity Map of the German Research Reactor FRJ-2"", 《DIESE DISSERTATION IST AUF DEN INTERNETSEITEN DER HOCHSCHULBIBLIOTHEK ONLINE VERFÜGBAR》 * |
| ZHENG ZHENG等: ""A deterministic and Monte Carlo coupling method for PWR cavity radiation streaming calculation"", 《ANNALS OF NUCLEAR ENERGY》 * |
| 李满仓: ""连续能量蒙特卡罗方法组件均匀化研究"", 《中国博士学位论文全文数据库 工程科技II辑》 * |
| 杜夏楠等: ""基于点截面的快堆组件均匀化参数计算方法"", 《强激光与粒子束》 * |
| 杜夏楠等: ""蒙特卡罗方法在快堆组件参数计算中的应用"", 《核动力工程》 * |
| 郑友琦等: ""快堆中子学计算程序NECP-SARAX的改进与验证"", 《强激光与粒子束》 * |
Cited By (9)
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| CN108733903A (en) * | 2018-05-08 | 2018-11-02 | 西安交通大学 | A kind of neutron transport method for numerical simulation carrying out personalisation process by neutron energy |
| CN109493924A (en) * | 2018-11-13 | 2019-03-19 | 西安交通大学 | A method of obtaining the effective multigroup cross section of FCM fuel |
| CN110704806A (en) * | 2019-09-19 | 2020-01-17 | 西安交通大学 | Rapid online calculation method for one-dimensional cylindrical geometric collision probability |
| CN110704806B (en) * | 2019-09-19 | 2021-03-02 | 西安交通大学 | Rapid online calculation method for one-dimensional cylindrical geometric collision probability |
| CN112966428A (en) * | 2021-03-31 | 2021-06-15 | 清华大学 | Cross-section treatment system for reactor core |
| CN113868591A (en) * | 2021-09-28 | 2021-12-31 | 西安交通大学 | Method for obtaining high-precision total reaction section of indistinguishable resonance region |
| CN113868591B (en) * | 2021-09-28 | 2022-12-09 | 西安交通大学 | Method for obtaining high-precision total reaction cross section of unresolved resonance region |
| CN117150829A (en) * | 2023-10-31 | 2023-12-01 | 西安交通大学 | Method for generating stack-used few-group cross-section energy group structure for neutron analysis of intermediate energy spectrum |
| CN117150829B (en) * | 2023-10-31 | 2024-01-23 | 西安交通大学 | Method for generating stack-used few-group cross-section energy group structure for neutron analysis of intermediate energy spectrum |
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