CN111584019B - Method for obtaining response of detector outside reactor based on first collision source-Monte Carlo coupling - Google Patents
Method for obtaining response of detector outside reactor based on first collision source-Monte Carlo coupling Download PDFInfo
- Publication number
- CN111584019B CN111584019B CN202010382546.3A CN202010382546A CN111584019B CN 111584019 B CN111584019 B CN 111584019B CN 202010382546 A CN202010382546 A CN 202010382546A CN 111584019 B CN111584019 B CN 111584019B
- Authority
- CN
- China
- Prior art keywords
- neutron
- source
- collision
- flux density
- detector
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Active
Links
Classifications
-
- G—PHYSICS
- G16—INFORMATION AND COMMUNICATION TECHNOLOGY [ICT] SPECIALLY ADAPTED FOR SPECIFIC APPLICATION FIELDS
- G16C—COMPUTATIONAL CHEMISTRY; CHEMOINFORMATICS; COMPUTATIONAL MATERIALS SCIENCE
- G16C20/00—Chemoinformatics, i.e. ICT specially adapted for the handling of physicochemical or structural data of chemical particles, elements, compounds or mixtures
- G16C20/70—Machine learning, data mining or chemometrics
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C17/00—Monitoring; Testing ; Maintaining
- G21C17/10—Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain
- G21C17/108—Measuring reactor flux
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Landscapes
- Engineering & Computer Science (AREA)
- Physics & Mathematics (AREA)
- Evolutionary Computation (AREA)
- Software Systems (AREA)
- High Energy & Nuclear Physics (AREA)
- Health & Medical Sciences (AREA)
- Artificial Intelligence (AREA)
- Computer Vision & Pattern Recognition (AREA)
- Data Mining & Analysis (AREA)
- Databases & Information Systems (AREA)
- Plasma & Fusion (AREA)
- General Health & Medical Sciences (AREA)
- Medical Informatics (AREA)
- General Engineering & Computer Science (AREA)
- Chemical & Material Sciences (AREA)
- Crystallography & Structural Chemistry (AREA)
- Life Sciences & Earth Sciences (AREA)
- Bioinformatics & Cheminformatics (AREA)
- Bioinformatics & Computational Biology (AREA)
- Computing Systems (AREA)
- Theoretical Computer Science (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Measurement Of Radiation (AREA)
Abstract
Method for obtaining response of detector outside reactor based on first collision source-Monte Carlo coupling, according to material and geometry of reactorThe method is characterized in that a reactor region is divided into an in-core region containing a reactor core, a reflecting layer and a shielding layer and an out-of-core region containing an out-of-core detector, and the neutron flux density at the out-of-core detector is divided into non-collision neutron flux density phi according to a primary collision source technology u (r, E) and collision neutron flux density phi c (r, E); calculating non-collision neutron flux density phi from external neutron point source by using semi-analytic ray tracing method u (r, E) with primary collision source Q c (r, Ω, E); from the first collision source Q using the Monte Carlo method c (r, omega, E) calculating to obtain collision neutron flux density phi c (r, E); adding the two parts of neutron flux densities at the position of the out-of-pile detector to obtain the neutron flux density at the position of the out-of-pile detector, and calculating the detector response according to the neutron flux density at the position of the out-of-pile detector; in the method, based on the first collision source technology, the advantages of the Monte Carlo method and the semi-analytic ray tracing method are respectively utilized, and the statistical efficiency and the calculation accuracy of the deep penetration problem are improved.
Description
Technical Field
The invention relates to calculation of a nuclear reactor core deep penetration problem, in particular to a method for acquiring a reactor external detector response based on first collision source-Monte Carlo coupling.
Background
The monte carlo method is also called a probability theory method, and firstly a model is established, so that the solution of a problem is the mathematical expectation of a certain random variable or the quantity related to the mathematical expectation, and then the arithmetic mean value of a plurality of specific observed values of the random variable is statistically estimated through an experimental method, namely the solution of the problem. Solving a neutron transport equation by using a Monte Carlo method, establishing a model according to the physical process of neutron transport, and calculating the contribution of the model to neutron flux density or other response quantities by simulating the motion history of a large number of neutrons so as to obtain the statistical estimation value of the neutron flux density or other response quantities.
The Monte Carlo method is used for directly simulating from an external neutron point source, namely the Monte Carlo method is used for simulating in the whole process, so that the problems of large calculated amount and low statistical efficiency of a target detector exist; the flight direction of neutrons generated by a point source is biased by using a source bias skill, neutrons with larger deviation from a target direction are abandoned, the weight of the rest neutrons is improved, the unbiased property is ensured, and the simulation number of a target area is improved at the same time, so that the variance and the total simulation calculated amount are reduced. The source bias is used for simulating from a single point source, only the direction of the point source is biased, in the simulation process of the rest neutrons, a part of neutrons collide and fly to gradually deviate from the position of a target detector, the statistical contribution of the part of neutrons to the neutron flux density at the target position is very small, and a lot of calculated amount is increased when the part of neutrons are tracked by a Monte Carlo method; it can be known from the definition of the quality factor that the quality factor is inversely proportional to the product of the calculation time and the square of the statistical relative error, and when the statistical relative error is guaranteed to be constant, the longer the calculation time is, the lower the quality factor is, the lower the statistical efficiency of the target position is.
Disclosure of Invention
The invention aims to provide a method for obtaining the response of an out-of-reactor detector based on the first collision source-Monte Carlo coupling, which is characterized in that the first collision source technology in a determinacy method is combined with the Monte Carlo method to realize the coupling of the determinacy-Monte Carlo method, the advantage of rapid calculation in the deep penetration problem is fully utilized to process the advantage of the complex geometric problem of the out-of-reactor detector part in the deep penetration problem by the determinacy method, the Monte Carlo calculation part in the first collision source optimization shielding calculation is generated by combining the first collision source technology, an outer neutron source only distributed in an in-reactor area is converted into the first collision source distributed in the whole reactor area, compared with the Monte Carlo method, the calculation is directly started from the outer neutron source, the number of particles reaching the out-of-reactor detector is more, the biased neutron sampling operation can be carried out on the spatial distribution of the first collision source, the number of the particles reaching the out-of-reactor detector is further improved, the calculation efficiency of the flux density of the neutron source at the out-reactor detector can be improved, and the neutron flux density at the target detector is more accurate and the neutron flux density can be calculated more rapidly;
in order to realize the purpose, the invention adopts the following technical scheme to implement:
a method for obtaining a reactor-external detector response based on a first collision source-monte carlo coupling, the method comprising the steps of:
step 1: according to the material and geometric characteristics of the reactor, the reactor area is divided into an in-core area containing a reactor core, a reflecting layer and a shielding layer and an out-of-core area containing an out-of-core detector, and the neutron flux density is divided into non-collision neutron flux density phi according to the first collision source technology u (r, E) and collision neutron flux density phi c (r, E); dividing the whole reactor region into three-dimensional rectangular coordinate system grids, taking an external neutron source in an actual reactor as a source, converting the external neutron source into an external neutron source located at the center of the source region grid, wherein the conversion mode is as follows: integrating the distribution of the external neutron source intensity density with the volume of the grid to obtain the neutron source intensity at the center of the grid, namely, the source in the whole grid is equivalent to a point source at the center of the grid; starting from an external neutron point source, calculating by adopting a semi-analytic ray tracing method to obtain the density phi of the non-collision flux in the whole reactor area u (r, E); for an external neutron source, calculating the secondary neutron source r by using a semi-analytic ray tracing method p Angular flux density of non-colliding neutrons φ at a given position r in arrival space u (r, Ω, E), the calculation formula is:
from non-colliding neutrons angular flux density phi u (r, Ω, E) further calculating the uncontacted neutron flux density φ u (r, E) the calculation formula is:
wherein phi u (r, Ω, E) is the non-impinging neutron angular flux density, r is the spatial position, Ω is the neutron motion direction, E is the neutron energy,
is neutron from r p Where it reaches the direction of motion at r,
is the Dirac function, q is the source strength of the external neutron point source, τ (r) p And r) is neutron slave r p The optical distance from the position r to the position r is calculated by the following formula:
τ(r p ,r)＝∫ s Σ t ds (3)
wherein t Is the total cross section, s is the neutron from r p The length of the path from where the beam travels to where r is reached; calculation of first-time collision source Q from non-collision neutron angular flux density c (r, Ω, E), the calculation formula is:
wherein Q c (r, Ω, E) is the first collision source with neutron energy E in direction Ω at spatial location r, Σ s (r, E ', Ω' → E, Ω) are scattering cross sections of neutrons at spatial location r from energy E 'and direction Ω' to energy E and direction Ω, χ (E) is the fission spectrum, υ is the average number of neutrons per fission, Σ f (r, E ') is the fission cross section of a neutron at spatial position r with energy E', and phi (r, omega ', E') is the neutron angular flux density at spatial position r with energy E 'and direction omega';
step 2: first collision source Q obtained from step 1 by the Monte Carlo method c Starting from (r, omega and E), simulating the neutron transportation process, adopting source bias, simultaneously reducing the weight of a first collision source far away from the target out-of-pile detector, improving the weight of the first collision source near the target out-of-pile detector, counting in a grid where the out-of-pile detector is located, and obtaining the collision neutron flux density phi of the out-of-pile detector c (r,E)；
And step 3: (iii) collisional neutron flux density phi at the out-of-pile detector c (r, E) and the non-collision neutron flux density at the out-of-core detector u (r, E) are added to obtain the neutron flux density phi (r, E) at the target detector, and the neutron flux density phi (r, E) at the target detector is added withResponse function Σ for off-heap probes d The energy and the space are multiplied and integrated to obtain the response RES of the out-of-stack detector.
Compared with the prior art, the invention has the following outstanding advantages:
1. compared with the Monte Carlo method, the non-collision neutron flux is calculated by a semi-analytic ray tracing method, and the calculation speed is high for a determinism method;
2. compared with an external neutron source only distributed in a reactor, the first collision source distributed in the whole reactor area is wider in distribution range, the first collision source is calculated by a Monte Carlo method, the number of neutrons reaching the detector outside the reactor is more, biased sampling operation can be carried out according to the spatial distribution and the direction of the first collision source, the number of neutrons reaching the detector outside the reactor is further improved, the variance of calculation results of the Monte Carlo method is reduced, and calculation efficiency is improved.
Detailed Description
According to the invention, the first collision source technology is applied to the calculation of the deep penetration problem, and a semi-analytic ray tracing method and a Monte Carlo method are adopted for calculation. The specific calculation flow of the method comprises the following aspects:
step 1: according to the material and geometrical characteristics of the reactor, the reactor area is divided into an in-core area containing a reactor core, a reflecting layer and a shielding layer and an out-of-core area containing an out-of-core detector, and the neutron flux density is divided into a non-collision neutron flux density phi according to the first collision source technology u (r, E) and collisional neutron flux density phi c (r, E); firstly, dividing the whole reactor area into grids under a three-dimensional rectangular coordinate system, wherein an external neutron source in the actual reactor is generally a source, and the external neutron source is converted into an external neutron source located at the center of the grid of the source area in a conversion mode: integrating the distribution of the source intensity of the external neutron source to the volume of the grid to obtain the source intensity of the neutron source positioned at the center of the grid, namely, the source in the whole grid is equivalent to the point source at the center of the grid; starting from an external neutron point source, calculating by adopting a semi-analytic ray tracing method to obtain the density phi of the non-collision flux in the whole reactor area u (r, E); for an external neutron source, semi-resolved rays are usedTracking method for calculating secondary neutron point source r p Angular flux density of non-collided neutrons phi at a given position r in space where the generated neutrons arrive u (r, E, Ω), the calculation formula is:
can be determined by the non-collision neutron angular flux density phi u (r, E, Ω) further calculating the uncontacted neutron flux density φ u (r, E) according to the formula:
wherein phi u (r, E, Ω) is the non-impinging neutron angular flux density, Ω is the neutron motion direction,
is neutron slave r p Where it reaches the direction of motion at r,
is the Dirac function, q is the source strength of the external neutron point source, τ (r) p And r) is neutron slave r p The optical distance from the position r to the position r is calculated by the following formula:
τ(r p ,r)＝∫ s Σ t ds (3)
wherein t Is the total cross-section, s is the neutron from r p The length of the path from where the beam travels to where r is reached; calculating a first collision source Q from non-collision neutron angular flux density c (r, Ω, E), the calculation formula is:
wherein Q c (r, Ω, E) is the first collision source with neutron energy E in direction Ω at spatial location r, Σ s (r, E ', Ω' → E, Ω) is the neutron flux at spatial position rEnergy E 'and direction Ω' are scattered to scattering cross sections of energy E and direction Ω, χ (E) is the fission spectrum, ν is the average neutron number per fission, Σ f (r, E ') is the fission cross section of a neutron at spatial position r with energy E', and phi (r, omega ', E') is the neutron angular flux density at spatial position r with energy E 'and direction omega';
step 2: starting from the first collision source obtained in the step 1 by a Monte Carlo method, the first collision source Q is c (r, omega, E) is taken as an external neutron source, source bias is carried out according to the spatial distribution of a first collision source, the weight of the first collision source in a grid far away from the out-of-pile detector is improved, the weight of the first collision source in the grid close to the out-of-pile detector is reduced, a neutron transport process (5) is simulated, counting is carried out in the grid at the out-of-pile detector, and the collision neutron flux density phi at the out-of-pile detector is obtained c (r,E)；；
Wherein phi c (r, Ω, E) is the collision neutron angular flux density with neutron energy E at spatial position r and with flight direction Ω, Σ t Is the total cross section ∑ s (r, E ', Ω' → E, Ω) are scattering cross sections for neutrons from energy E 'and direction Ω' to energy E and direction Ω at spatial position r, Σ f (r, E ') is the fission cross section of a neutron at a spatial location r from energy E', χ (E) is the fission spectrum, ν is the average number of neutrons per fission, Q c (r, Ω, E) is the first collision source at spatial position r, direction Ω, energy E;
and step 3: calculating the density phi of non-collision neutron flux at the position of the detector outside the reactor by using a semi-analytic ray tracing method u (r, E) collision neutron flux density phi at the position of the out-of-pile detector calculated by Monte Carlo method c (r, E) adding in the grid at the out-of-pile detector to obtain the neutron flux density phi (r, E) at the out-of-pile detector, and the neutron flux density at the out-of-pile detector and the response function sigma of the detector d Multiplying and integrating the energy and the space to obtain the out-of-pile detectorThe response RES of (a) is calculated by the formula:
RES＝∫∫Σ d φ(r,E)dEdr (6)。
Claims (1)
1. the method for acquiring the response of the detector outside the reactor based on the first collision source-Monte Carlo coupling is characterized in that: the method comprises the following steps:
step 1: according to the material and geometrical characteristics of the reactor, the reactor area is divided into an in-core area containing a reactor core, a reflecting layer and a shielding layer and an out-of-core area containing an out-of-core detector, and the neutron flux density is divided into a non-collision neutron flux density phi according to the first collision source technology u (r, E) and collision neutron flux density phi c (r, E); dividing the whole reactor region into three-dimensional rectangular coordinate system grids, taking an external neutron source in an actual reactor as a source, converting the external neutron source into an external neutron source located at the center of the source region grid, wherein the conversion mode is as follows: integrating the distribution of the source intensity of the external neutron source to the volume of the grid to obtain the source intensity of the neutron source positioned at the center of the grid, namely, the source in the whole grid is equivalent to the point source at the center of the grid; calculating the density phi of non-collision flux in the whole reactor area by using a semi-analytic ray tracing method from an external neutron point source u (r, E); for an external neutron source, calculating the secondary neutron source r by using a semi-analytic ray tracing method p Angular flux density of non-colliding neutrons φ at a given position r in arrival space u (r, Ω, E), the calculation formula is:
from non-colliding neutrons angular flux density phi u (r, Ω, E) further calculating the uncontacted neutron flux density φ u (r, E) according to the formula:
wherein phi u (r, Ω, E) is the non-impinging neutron angular flux density, r is the spatial position, Ω is the neutron direction of motion, E is the neutron energy,
is neutron from r p Where it reaches the direction of motion at r,
is the Dirac function, q is the source strength of the external neutron point source, τ (r) p R) is neutron from r p The optical distance from the position where the light passes to the position r is calculated by the formula:
τ(r,r p )＝∫ s Σ t ds (3)
wherein t Is the total cross section, s is the neutron from r p The length of the path from where the beam travels to where r is reached; calculation of first-time collision source Q from non-collision neutron angular flux density c (r, Ω, E), the calculation formula is:
wherein Q c (r, Ω, E) is the first collision source with neutron energy E in direction Ω at spatial location r, Σ s (r, E ', Ω' → E, Ω) are scattering cross sections of neutrons at spatial location r from energy E 'and direction Ω' to energy E and direction Ω, χ (E) is the fission spectrum, υ is the average number of neutrons per fission, Σ f (r, E ') is the fission cross section of a neutron at spatial position r at energy E', and φ (r, Ω ', E') is the neutron angular flux density at spatial position r at energy E 'and direction Ω';
step 2: first collision source Q obtained from step 1 by the Monte Carlo method c Starting from (r, omega and E), simulating the neutron transportation process, adopting source bias, simultaneously reducing the weight of a first collision source far away from a target out-of-pile detector, improving the weight of the first collision source near the target out-of-pile detector, counting grids where the out-of-pile detectors are located, and obtaining pilesCollision neutron flux density phi at the outer detector c (r,E)；
And 3, step 3: (iii) collisional neutron flux density phi at the out-of-pile detector c (r, E) and the non-collision neutron flux density at the out-of-core detector u (r, E) are added to obtain the neutron flux density phi (r, E) at the target detector, and the neutron flux density phi (r, E) at the target detector and the response function sigma of the out-of-pile detector are compared d The energy and space are multiplied and integrated to obtain the response RES of the out-of-stack detector.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202010382546.3A CN111584019B (en) | 2020-05-08 | 2020-05-08 | Method for obtaining response of detector outside reactor based on first collision source-Monte Carlo coupling |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202010382546.3A CN111584019B (en) | 2020-05-08 | 2020-05-08 | Method for obtaining response of detector outside reactor based on first collision source-Monte Carlo coupling |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| CN111584019A CN111584019A (en) | 2020-08-25 |
| CN111584019B true CN111584019B (en) | 2022-12-09 |
Family
ID=72124789
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN202010382546.3A Active CN111584019B (en) | 2020-05-08 | 2020-05-08 | Method for obtaining response of detector outside reactor based on first collision source-Monte Carlo coupling |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN111584019B (en) |
Families Citing this family (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN113011077B (en) * | 2021-04-02 | 2023-05-05 | 西安交通大学 | Method for obtaining response of external detector of nuclear reactor by coupling discrete longitudinal standard-Monte Carlo method |
| CN113553550B (en) * | 2021-07-27 | 2022-10-04 | 清华大学 | Monte Carlo particle transport calculation method and related calculation method |
| CN115274160B (en) * | 2022-06-14 | 2024-08-16 | 西北核技术研究所 | Reactor thermal neutron flux detection method and device based on silicon carbide semiconductor |
Family Cites Families (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN101399091B (en) * | 2008-11-07 | 2012-02-01 | 西安交通大学 | Method for on-line monitoring neutron flux distribution of nuclear reactor core |
| CN108447574B (en) * | 2018-03-28 | 2020-08-11 | 中广核工程有限公司 | Method and system for calculating neutron counting rate of out-of-reactor detector of nuclear power station |
| CN109767854B (en) * | 2019-01-21 | 2022-07-08 | 中国科学院合肥物质科学研究院 | Reactor internal neutron three-dimensional distribution measuring system based on reactor external measurement data |
-
2020
- 2020-05-08 CN CN202010382546.3A patent/CN111584019B/en active Active
Also Published As
| Publication number | Publication date |
|---|---|
| CN111584019A (en) | 2020-08-25 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| CN111584019B (en) | Method for obtaining response of detector outside reactor based on first collision source-Monte Carlo coupling | |
| CN108549753B (en) | Radiation shielding calculation method for coupling point kernel integration method and Monte Carlo method | |
| US20180060463A1 (en) | Hybrid Monte Carlo and Deterministic Particle Transport Method Based on Transition Area | |
| US6795801B1 (en) | Apparatus and method for analyzing anisotropic particle scattering in three-dimensional geometries | |
| CN113609099B (en) | Method for manufacturing fusion reactor multi-group shielding database based on Monte Carlo method | |
| Dedrick | Deuteron-model calculation of the high-energy nuclear photoeffect | |
| JP4373876B2 (en) | Nuclear fuel nuclear constant preparation method, core design method using the same, nuclear fuel nuclear constant preparation apparatus, and core design apparatus using the same | |
| CN109655473B (en) | Simulation method and system of flash photography image receiving device for counting of point detector | |
| CN114997035B (en) | Global variance reduction method based on volume and non-counting area correction | |
| Long et al. | MONK10: A Monte Carlo code for criticality analysis | |
| CN111584104B (en) | Method for obtaining response of detector outside reactor based on multiple collision source-Monte Carlo coupling | |
| Jafarikia et al. | Built in importance estimation in forward Monte Carlo calculations | |
| Gollin et al. | Charm production by muons and its role in scale noninvariance | |
| CN114781230A (en) | Method for obtaining gamma dose rate distribution based on finite element-Monte Carlo-point kernel integral coupling | |
| Kaltiaisenaho | Statistical Tests and the Underestimation of Variance in Serpent 2 | |
| CN111553110B (en) | Monte Carlo simulation-based nuclear electromagnetic pulse current source term calculation method | |
| CN113553550B (en) | Monte Carlo particle transport calculation method and related calculation method | |
| Bremer | Measurements and simulations on position dependencies in the response of single PWO crystals and a prototype for the PANDA EMC | |
| Zavorka et al. | Hybrid unstructured mesh geometry model of the next-generation spallation neutron source at LANSCE | |
| Karriem | Development of the method of characteristics for embedded lattice physics calculations | |
| Shultis et al. | Extensions to the integral line-beam method for gamma-ray skyshine analyses | |
| CN108595796A (en) | The computational methods and medium of large space deep penetration radiation field Dose Rate | |
| Boyington | Neutron flux characterization of the kansas state university triga mark ii’s northeast beam port | |
| Matijević et al. | Spent fuel pool dose rate calculations using point kernel and hybrid deterministic-stochastic shielding methods | |
| Walters et al. | A response function methodology to calculate induced fission neutron flux distribution in an active interrogation system |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| PB01 | Publication | ||
| PB01 | Publication | ||
| SE01 | Entry into force of request for substantive examination | ||
| SE01 | Entry into force of request for substantive examination | ||
| GR01 | Patent grant | ||
| GR01 | Patent grant |