CN110765618B - Method for calculating response current of self-powered neutron detector in pressurized water reactor - Google Patents
Method for calculating response current of self-powered neutron detector in pressurized water reactor Download PDFInfo
- Publication number
- CN110765618B CN110765618B CN201911028353.1A CN201911028353A CN110765618B CN 110765618 B CN110765618 B CN 110765618B CN 201911028353 A CN201911028353 A CN 201911028353A CN 110765618 B CN110765618 B CN 110765618B
- Authority
- CN
- China
- Prior art keywords
- self
- photon
- neutron detector
- neutron
- powered
- 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
Landscapes
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Measurement Of Radiation (AREA)
Abstract
The invention discloses a method for calculating response current of a self-powered neutron detector in a pressurized water reactor, which is characterized in that under the existing commercial pressurized water reactor core design analysis framework, impulse response functions of different types of particles of the self-powered neutron detector under different burnup and different power levels are obtained through grid calculation, a few-group parameter fitting program is used for fitting the parameters to generate a database of the variation relation of the impulse response functions along with the parameters of the local state of the reactor core, and finally, the current simulation of the self-powered neutron detector under each historical state is completed in the reactor core analysis process; the response current of the self-powered neutron detector at each moment can be accurately obtained without being limited by the change of the reactor core state; the method has good engineering application prospect.
Description
Technical Field
The invention relates to the field of theoretical simulation of an in-reactor monitoring system of a reactor and safe operation of the reactor, in particular to a method for calculating response current of a self-powered neutron detector in a pressurized water reactor.
Background
The self-powered neutron detector is an in-pile fixed detector based on an activation method and mainly comprises an emitter, an insulator and a collector. Neutrons within the nuclear reactor core react with the high neutron sensitivity emitter material to produce electrons, which form an electrical current.
The self-powered neutron detector has the characteristics of small volume, high neutron sensitivity, no need of a bias power supply and the like, is widely applied to heavy water reactors, VVER reactor cores and third-generation large commercial pressurized water reactors at present, can monitor the neutron flux level in the reactor cores in real time, realizes real-time online monitoring on the power distribution of the reactor cores of the reactors, and greatly improves the safety and the economical efficiency of the reactor cores. The simulation of the response current of the self-powered neutron detector is a key link for realizing safety control and protection of the nuclear power plant based on-line monitoring.
The forming process of the response current of the self-powered neutron detector is very complicated, and the key point is to establish the relation between the neutron flux in the reactor core and the response current through an accurate sensitivity coefficient. The prior art is based on an analytical model and Monte Carlo sampling analysis, and uses a neutron source hypothesis with a fixed monoenergetic or typical neutron energy spectrum, neglects the contribution of photons to response current, can roughly give response current under a stable power level, and cannot be applied to the condition when the power state changes. However, in the core in actual operation, the neutron coupling field is constantly changed along with the core state, the correlation between the actual neutron energy spectrum and a plurality of local state parameters such as the actual burnup depth, the boron concentration, the fuel temperature, the moderator temperature and the like is strong, and a dangerous state is more likely to occur in the variable power process. Therefore, it is necessary to invent a response current calculation method suitable for a self-powered neutron detector in a pressurized water reactor.
Disclosure of Invention
In order to overcome the problems in the prior art, the invention aims to provide a method for calculating the response current of a self-powered neutron detector in a pressurized water reactor, which is applied to the response current analysis of the self-powered neutron detector of a reactor core program; the method includes the steps that under the existing commercial pressurized water reactor core design analysis framework, impulse response functions of different types of particles of the self-powered neutron detector under different burnup and different power levels are obtained through grid calculation, a few-group parameter fitting program is used for fitting the parameters to generate a database of the relation of the impulse response functions with the change of the parameters of the local state of the reactor core, and finally, the current simulation of the self-powered neutron detector under each historical state is completed in the reactor core analysis process.
In order to achieve the purpose, the invention adopts the following technical scheme,
a method for calculating response current of a self-powered neutron detector in a pressurized water reactor comprises the following steps:
step 1: reading the geometric dimension, material arrangement information and state parameter information of a fuel assembly to be simulated in a pressurized water reactor core, and the geometric structure and material information of a self-powered neutron detector arranged in a measuring pipe of a central instrument of the fuel assembly; wherein the state parameter information comprises a component average burnup depth BU, an in-component fuel temperature TF, an in-component moderator temperature TM, and an in-component boron concentration CB;
step 2: according to the geometric dimension, the material arrangement information and the state parameter information of the fuel assemblies to be simulated in the pressurized water reactor core, which are obtained in the step 1, dividing the fuel grid cells into a plurality of burnup areas along the axial direction, and establishing a two-dimensional assembly calculation model:
firstly, obtaining the boundary of each calculation area in the assembly according to the geometric description of the fuel assembly, then establishing a mathematical equation set of a characteristic line and each calculation area boundary, solving the equation set to calculate the intersection point coordinates of the characteristic line and each calculation area, and obtaining the corresponding characteristic line information;
carrying out transport calculation by adopting a characteristic line method to obtain a neutron-photon flux density shape in a grid where the self-powered neutron detector is positioned; obtaining nuclide composition in each burnup area and the atomic nucleus density of each nuclide under each burnup step by adopting neutron transport-burnup coupling calculation of a pre-estimation correction method;
and step 3: according to the burnup regions divided in the step 2, the obtained nuclide composition and the nucleus density of each nuclide, establishing a mathematical equation of the change of the nucleus density of each nuclide and the nucleus density of other nuclides in each burnup region in unit time, wherein the mathematical equation is called a burnup equation, and solving the burnup equation by adopting a Chebyshev rational approximation method or a linear subchain method to obtain the decay photon source intensity of each burnup region;
and 4, step 4: normalizing the decay photon source intensity of each burnup area obtained by calculating in the step 3 by adopting a linear function normalization method; according to the nuclide composition and the atomic nucleus density of each nuclide in each burnup area obtained in the step 2, the decay photon source intensity after normalization is used as a fixed source item generated by photons at the right end of a photon transport equation, a photon transport equation is established, the simultaneous neutron transport equation and the photon transport equation are solved by adopting a Monte Carlo method to obtain neutron source information, prompt-emission photon source information and slow-emission photon source information near the self-powered neutron detector and power in components, and the sampling probability of the prompt-emission photon source and the slow-emission photon source is determined according to the formulas (3) and (4):
in the formula, PpRepresenting the probability of sampling of a source of prompt photons, PdRepresenting the sampling probability of the slow-emitting photon source, ApRepresenting the amplitude coefficient of the source of prompt photons, obtained by dividing the real power by the power in the assembly obtained statistically, AdRepresenting the amplitude coefficient, w, of the slow-emitting photon sourcepRepresenting the weight of the prompt photon, w, contained in the source informationdRepresenting slow-emitting photon weight, npspRepresenting the number of simulation times, nps, of neutron-photon coupled transport calculation of a prompt photon sourcedRepresenting the simulation times when the neutron-photon coupling transport calculates the slow luminescence source;
and 5: according to the geometric structure and material information of the self-powered neutron detector read in the step 1, the neutron source information, the prompt emission photon source information and the slow emission photon source information near the self-powered neutron detector obtained in the step 4 and the sampling probability of the prompt emission photon source and the slow emission photon source, a cylindrical surface which is 0-0.5cm away from the outer surface of the self-powered neutron detector is taken as an outer boundary, and a two-dimensional self-powered detector calculation model is established for a grid where the self-powered neutron detector is located:
firstly, determining the boundary of each calculation area of the self-powered neutron detector according to the geometric description of the self-powered neutron detector, then determining the angle and the energy of initial neutrons according to the information of a neutron source near the self-powered neutron detector, then determining the type of initial photons according to the sampling probabilities of a prompt photon source and a slow photon source, and finally determining the angle and the energy of the initial photons according to the information of the prompt photon source and the slow photon source near the self-powered neutron detector;
simulating a neutron-photon-electron coupling transport process by adopting a Monte Carlo method to obtain an impulse response function of response current of the self-powered neutron detector;
step 6: changing the state parameters: and (3) repeatedly performing the calculation from the step (2) to the step (5) on the average burn-up depth BU of the component, the temperature TF of the fuel in the component, the temperature TM of the moderator in the component and the concentration CB of boron in the component to obtain the impulse response functions of the response currents under different state parameters, wherein the impulse response functions delta of the response currents under different state parameters can be expressed as:
δ＝f(BU,CB,TM,TF) (5)
and 7: and 6, fitting the impulse response function delta of the response current under different state parameters obtained in the step 6 by adopting the following two methods:
a) fitting the discrete function values of the impulse response function delta of the response current under different state parameters at different time respectively, and obtaining a function formed by the discrete function values after fitting, namely an impulse response function fitting relational expression of the response current;
b) firstly, performing polynomial decomposition on impulse response functions of response current under different state parameters, fitting each polynomial, and finally adding the fitted polynomials to obtain an impulse response function fitting relation of the response current;
and 8: performing reactor core diffusion calculation by using a physical calculation program of the pressurized water reactor core to obtain neutron flux and photon flux at a self-powered neutron detector in each component of the reactor core, and performing physical-thermal coupling calculation to obtain state parameters of the components at each burnup point;
and step 9: calculating to obtain an impulse response function of the response current under each fuel consumption point according to the impulse response function fitting relation of the response current obtained in the step 7 and the state parameters of the components under each fuel consumption point obtained in the step 8;
step 10: determining an input function in the self-powered neutron detector system, wherein according to the working principle of the self-powered neutron detector system, the input function of the system can be represented by neutron flux and photon flux of a grid where the self-powered neutron detector is located, so that the real neutron photon flux is adopted as the input function in the self-powered neutron detector system; the input function in a self-powered neutron detector system in a practical working environment can be divided into three parts: neutron input function, prompt photon input function, slow photon input function. The three input functions can be represented by equations (7), (8), (9), respectively:
xneutron[n]＝Anormφn(t)Ri,g (7)
xpp[n]＝Anormφp(t)Ri,g (8)
wherein xneutron[n]Representing the neutron input function, xpp[n]Representing the prompt photon input function, xdp[n]Representing slow-emitting photon input function, AnormDenotes the normalized coefficient, phin(t) represents the neutron flux, φp(t) represents photon flux, Ri,gRepresenting the shape of the flux of the ith group of particles from the grid on which the energized neutron detector is located, SdpDenotes the size, λ, of the slow-emitting photon source per unit volumeiDecay constant of the ith nuclide, t represents a discrete time point;
step 11: performing convolution processing on the three input functions according to the geometric structure of the self-powered neutron detector read in the step 1, the impulse response function of the response current under each burnup point obtained in the step 9 and the three input functions in the self-powered neutron detector system obtained in the step 10 to obtain the real-time response current I of the self-powered neutron detector; the specific calculation formula of the convolution processing is shown as formula (10);
wherein VemitterRepresenting the emitter volume of the self-powered neutron detector, i representing the current generated by the three types of particles, neutron, prompt photon and slow photon, δi(t- τ) represents the impulse response function of the response current of the ith particle at the current burn-up point, xiRepresenting the input functions of three types of particles, neutron, prompt photon and slow photon.
Detailed Description
The present invention will be described in further detail with reference to specific embodiments.
The invention relates to an accurate simulation technology of self-powered neutron detector response current in a pressurized water reactor, which mainly comprises grid calculation, reactor core calculation and self-powered neutron detector current response calculation, and comprises the following specific steps:
step 1: reading the geometric dimension, material arrangement information and state parameter information of a fuel assembly to be simulated in a pressurized water reactor core, and the geometric structure and material information of a self-powered neutron detector arranged in a measuring pipe of a central instrument of the fuel assembly; wherein the state parameter information comprises a component average burnup depth BU, an in-component fuel temperature TF, an in-component moderator temperature TM, and an in-component boron concentration CB;
step 2: according to the geometric dimension, material arrangement information and state parameter information of the fuel assemblies to be simulated in the pressurized water reactor core, which are obtained in the step 1, dividing the fuel grid cells into 3 burnup regions along the axial direction, and establishing a two-dimensional assembly calculation model:
firstly, obtaining the boundary of each calculation area in the assembly according to the geometric description of the fuel assembly, then establishing a mathematical equation set of a characteristic line and each calculation area boundary, solving the equation set to calculate the intersection point coordinates of the characteristic line and each calculation area, and obtaining the corresponding characteristic line information;
carrying out transport calculation by adopting a characteristic line method to obtain a neutron-photon flux density shape in a grid where the self-powered neutron detector is positioned; obtaining nuclide composition and the nuclear density of each nuclide in each burnup area under each burnup step through neutron transport-burnup coupling calculation by adopting an estimation correction method;
and step 3: in each burnup region, according to the burnup region divided in the step 2, the obtained nuclide composition and the nucleus density of each nuclide, a mathematical equation of the variation of the nucleus density of each nuclide in unit time and the nucleus densities of other nuclides in each burnup region is established and called as a burnup equation, and the burnup equation is solved by adopting a Chebyshev rational approximation method to obtain the decay photon source intensity of each burnup region;
and 4, step 4: and (4) carrying out normalization processing on the decay photon source intensity of each burnup area obtained by calculation in the step (3) by adopting a linear function normalization method. The specific linear function normalization method has the following calculation formula:
wherein x represents the currently processed sample, min represents the minimum value of the sample data, and max represents the maximum value of the sample data;
according to the nuclide composition in each burnup region and the atomic nucleus density of each nuclide obtained in the step 2, the normalized decay photon source intensity is used as a fixed source item generated by photons at the right end of the transport equation, and a photon transport equation is established, wherein the photon transport equation can be expressed as follows:
where Ω denotes the particle motion direction, ψ denotes the photon angular flux density, ΣtDenotes the total photon cross section, QdecayRepresenting the normalized decay photon source intensity.
The simultaneous neutron transport equation and the photon transport equation are solved by adopting a Monte Carlo method to obtain neutron source information, prompt luminescence source information and slow luminescence source information near the self-powered neutron detector and power in the assembly, and the sampling probability of the prompt luminescence source and the slow luminescence source is determined according to the formulas (3) and (4);
in the formula, PpRepresenting the probability of sampling of a source of prompt photons, PdRepresenting the sampling probability of the slow-emitting photon source, ApComponent representing amplitude coefficient of neutron source, obtained by dividing real power by statisticsInternal power is obtained, AdRepresenting the amplitude coefficient, w, of the slow-emitting photon sourcepRepresenting the weight of the prompt photon, w, contained in the source informationdRepresenting slow-emitting photon weight, npspRepresenting the number of simulation times, nps, of neutron-photon coupled transport calculation of a prompt photon sourcedRepresenting the simulation times when the neutron-photon coupling transport calculates the slow luminescence source;
and 5: according to the geometric structure and material information of the self-powered neutron detector read in the step 1, the neutron source information, the prompt luminescence photon source information and the slow luminescence photon source information near the self-powered neutron detector obtained in the step 4 and the sampling probability of the prompt luminescence photon source and the slow luminescence photon source, in the embodiment, a cylindrical surface which is 0.38634cm away from the outer surface of the self-powered neutron detector is taken as an outer boundary, and a two-dimensional self-powered neutron detector calculation model is established for a grid where the self-powered neutron detector is located:
firstly, determining the boundary of each calculation area of the self-powered neutron detector according to the geometric description of the self-powered neutron detector, then determining the angle and the energy of initial neutrons according to the information of a neutron source near the self-powered neutron detector, then determining the type of initial photons according to the sampling probabilities of a prompt photon source and a slow photon source, and finally determining the angle and the energy of the initial photons according to the information of the prompt photon source and the slow photon source near the self-powered neutron detector;
simulating a neutron-photon-electron coupling transport process by adopting a Monte Carlo method to obtain an impulse response function of response current of the self-powered neutron detector;
step 6: changing the state parameters: and (3) repeatedly performing the calculation from the step (2) to the step (5) on the average burn-up depth BU of the component, the temperature TF of the fuel in the component, the temperature TM of the moderator in the component and the concentration CB of boron in the component to obtain the impulse response functions of the response currents under different state parameters, wherein the impulse response functions delta of the response currents under different state parameters can be expressed as:
δ＝f(BU,CB,TM,TF) (5)
and 7: the present example respectively fits the discrete function values of the impulse response function of the response current under different state parameters at different times, disperses the impulse response function δ of the response current under different state parameters obtained in step 6 into 3000 discrete function values, and functionalizes the discrete function values under different state parameters, and the present example adopts a fitting form:
δ(t)＝f(BU)·f(CB)·f(TM)·f(TF) (6)
in the formula, delta (t) represents an impulse response function expression of the fitted response current at a discrete time point t;
and 8: performing Core diffusion calculation by using a physical calculation program (in the example, a Bamboo-Core program) of the pressurized water reactor Core to obtain neutron flux and photon flux at a self-powered neutron detector in each component of the reactor Core, and performing physical-thermal coupling calculation to obtain state parameters of the components at each burnup point;
and step 9: calculating to obtain an impulse response function of the response current under each fuel consumption point according to the impulse response function fitting relation of the response current obtained in the step 7 and the state parameters of the components under each fuel consumption point obtained in the step 8;
step 10: determining an input function in the self-powered neutron detector system, wherein the input function of the system can be represented by the neutron flux and the photon flux of the grid where the self-powered neutron detector is located according to the working principle of the self-powered neutron detector system, so that the real neutron photon flux is adopted as the input function in the self-powered neutron detector system. The input function in a self-powered neutron detector system in a practical working environment can be divided into three parts: neutron input function, prompt photon input function, slow photon input function. The three input functions can be represented by equations (7), (8), (9), respectively:
xneutron[n]＝Anormφn(t)Ri,g (7)
xpp[n]＝Anormφp(t)Ri,g (8)
wherein xneutron[n]Representing neutron input function，xpp[n]Representing the prompt photon input function, xdp[n]Representing slow-emitting photon input function, AnormDenotes the normalized coefficient, phin(t) represents the neutron flux, φp(t) represents photon flux, Ri,gRepresenting the shape of the flux of the ith group of particles from the grid on which the energized neutron detector is located, SdpDenotes the size, λ, of the slow-emitting photon source per unit volumeiDecay constant of the ith nuclide, t represents a discrete time point;
step 11: performing convolution processing on the three input functions according to the geometric structure of the self-powered neutron detector read in the step 1, the impulse response function of the response current under each burnup point obtained in the step 9 and the three input functions in the self-powered neutron detector system obtained in the step 10 to obtain the real-time response current I of the self-powered neutron detector; the specific calculation formula of the convolution processing is shown in formula (10).
Wherein VemitterRepresenting the emitter volume of the self-powered neutron detector, i representing the current generated by the three types of particles, neutron, prompt photon and slow photon, δi(t- τ) represents the impulse response function of the response current of the ith particle at the current burn-up point, xiRepresenting the input functions of three types of particles, neutron, prompt photon and slow photon.
The method adds photon transport calculation in the calculation of the pressurized water reactor components, adds a photon diffusion calculation and impulse response function few group parameter library in the program of the reactor core of the pressurized water reactor, and can obtain the real-time response current of the in-reactor detector through the calculation of the reactor core without the limitation of specific reactor types, component types and self-powered neutron detectors. The current is divided into neutron generation current, transient photon generation current and neutron generation current according to the generated particles, and the response current of the self-powered neutron detector at each moment can be accurately obtained by performing convolution calculation on the currents generated by different types of particles respectively without limitation of the state change of the reactor core. The method has good engineering application prospect.
As the present invention may be embodied in several forms without departing from the spirit or essential characteristics thereof, it should also be understood that the above-described embodiments are not limited by any of the details of the foregoing description, but rather should be construed broadly within its spirit and scope as defined in the appended claims, and therefore all changes and modifications that fall within the meets and bounds of the claims, or equivalences of such meets and bounds are therefore intended to be embraced by the appended claims.
Claims (1)
1. A method for calculating response current of a self-powered neutron detector in a pressurized water reactor is characterized by comprising the following steps:
step 1: reading the geometric dimension, material arrangement information and state parameter information of a fuel assembly to be simulated in a pressurized water reactor core, and the geometric structure and material information of a self-powered neutron detector arranged in a measuring pipe of a central instrument of the fuel assembly; wherein the state parameter information comprises a component average burnup depth BU, an in-component fuel temperature TF, an in-component moderator temperature TM, and an in-component boron concentration CB;
step 2: according to the geometric dimension, the material arrangement information and the state parameter information of the fuel assemblies to be simulated in the pressurized water reactor core, which are obtained in the step 1, dividing the fuel grid cells into a plurality of burnup areas along the axial direction, and establishing a two-dimensional assembly calculation model:
firstly, obtaining the boundary of each calculation area in the assembly according to the geometric description of the fuel assembly, then establishing a mathematical equation set of a characteristic line and each calculation area boundary, solving the equation set to calculate the intersection point coordinates of the characteristic line and each calculation area, and obtaining the corresponding characteristic line information;
carrying out transport calculation by adopting a characteristic line method to obtain a neutron-photon flux density shape in a grid where the self-powered neutron detector is positioned; obtaining nuclide composition in each burnup area and the atomic nucleus density of each nuclide under each burnup step by adopting neutron transport-burnup coupling calculation of a pre-estimation correction method;
and step 3: according to the burnup regions divided in the step 2, the obtained nuclide composition and the nucleus density of each nuclide, establishing a mathematical equation of the change of the nucleus density of each nuclide and the nucleus density of other nuclides in each burnup region in unit time, wherein the mathematical equation is called a burnup equation, and solving the burnup equation by adopting a Chebyshev rational approximation method or a linear subchain method to obtain the decay photon source intensity of each burnup region;
and 4, step 4: normalizing the decay photon source intensity of each burnup area obtained by calculating in the step 3 by adopting a linear function normalization method; according to the nuclide composition and the atomic nucleus density of each nuclide in each burnup area obtained in the step 2, the decay photon source intensity after normalization is used as a fixed source item generated by photons at the right end of a photon transport equation, a photon transport equation is established, the simultaneous neutron transport equation and the photon transport equation are solved by adopting a Monte Carlo method to obtain neutron source information, prompt-emission photon source information and slow-emission photon source information near the self-powered neutron detector and power in components, and the sampling probability of the prompt-emission photon source and the slow-emission photon source is determined according to the formulas (3) and (4):
in the formula, PpRepresenting the probability of sampling of a source of prompt photons, PdRepresenting the sampling probability of the slow-emitting photon source, ApRepresenting the amplitude coefficient of the source of prompt photons, obtained by dividing the real power by the power in the assembly obtained statistically, AdRepresenting the amplitude coefficient, w, of the slow-emitting photon sourcepRepresenting the weight of the prompt photon, w, contained in the source informationdRepresenting slow-emitting photon weight, npspRepresenting the number of simulation times, nps, of neutron-photon coupled transport calculation of a prompt photon sourcedRepresenting the simulation times when the neutron-photon coupling transport calculates the slow luminescence source;
and 5: according to the geometric structure and material information of the self-powered neutron detector read in the step 1, the neutron source information, the prompt emission photon source information and the slow emission photon source information near the self-powered neutron detector obtained in the step 4 and the sampling probability of the prompt emission photon source and the slow emission photon source, a cylindrical surface which is 0-0.5cm away from the outer surface of the self-powered neutron detector is taken as an outer boundary, and a two-dimensional self-powered detector calculation model is established for a grid where the self-powered neutron detector is located:
firstly, determining the boundary of each calculation area of the self-powered neutron detector according to the geometric description of the self-powered neutron detector, then determining the angle and the energy of initial neutrons according to the information of a neutron source near the self-powered neutron detector, then determining the type of initial photons according to the sampling probabilities of a prompt photon source and a slow photon source, and finally determining the angle and the energy of the initial photons according to the information of the prompt photon source and the slow photon source near the self-powered neutron detector;
simulating a neutron-photon-electron coupling transport process by adopting a Monte Carlo method to obtain an impulse response function of response current of the self-powered neutron detector;
step 6: changing the state parameters: and (3) repeatedly performing the calculation from the step (2) to the step (5) on the average burn-up depth BU of the component, the temperature TF of the fuel in the component, the temperature TM of the moderator in the component and the concentration CB of boron in the component to obtain the impulse response functions of the response currents under different state parameters, wherein the impulse response functions delta of the response currents under the different state parameters are represented as:
δ＝f(BU,CB,TM,TF) (5)
and 7: and 6, fitting the impulse response function delta of the response current under different state parameters obtained in the step 6 by adopting the following two methods:
a) fitting the discrete function values of the impulse response function delta of the response current under different state parameters at different time respectively, and obtaining a function formed by the discrete function values after fitting, namely an impulse response function fitting relational expression of the response current;
b) firstly, performing polynomial decomposition on impulse response functions of response current under different state parameters, fitting each polynomial, and finally adding the fitted polynomials to obtain an impulse response function fitting relation of the response current;
and 8: performing reactor core diffusion calculation by using a physical calculation program of the pressurized water reactor core to obtain neutron flux and photon flux at a self-powered neutron detector in each component of the reactor core, and performing physical-thermal coupling calculation to obtain state parameters of the components at each burnup point;
and step 9: calculating to obtain an impulse response function of the response current under each fuel consumption point according to an impulse response function fitting relation of the response current obtained after fitting by any one of the two methods in the step 7 and the state parameters of the component under each fuel consumption point obtained in the step 8;
step 10: determining an input function in the self-powered neutron detector system, wherein the input function of the system is represented by neutron flux and photon flux of a grid where the self-powered neutron detector is located according to the working principle of the self-powered neutron detector system, so that the real neutron photon flux is adopted as the input function in the self-powered neutron detector system; the input function in the self-powered neutron detector system under the actual working environment is divided into three parts: a neutron input function, a prompt photon input function and a slow photon input function; the three input functions are represented by equations (7), (8), (9), respectively:
xneutron[n]＝Anormφn(t)Ri,g (7)
xpp[n]＝Anormφp(t)Ri,g (8)
wherein xneutron[n]Representing the neutron input function, xpp[n]Representing the prompt photon input function, xdp[n]Representing slow-emitting photon input function, AnormDenotes the normalized coefficient, phin(t) represents the neutron flux, φp(t) represents photon flux, Ri,gRepresenting the shape of the flux of the ith group of particles from the grid on which the energized neutron detector is located, SdpDenotes the size, λ, of the slow-emitting photon source per unit volumeiDecay constant of the ith nuclide, t represents a discrete time point;
step 11: performing convolution processing on the three input functions according to the geometric structure of the self-powered neutron detector read in the step 1, the impulse response function of the response current under each burnup point obtained in the step 9 and the three input functions in the self-powered neutron detector system obtained in the step 10 to obtain the real-time response current I of the self-powered neutron detector; the specific calculation formula of the convolution processing is shown as formula (10);
wherein VemitterRepresenting the emitter volume of the self-powered neutron detector, i representing the current generated by the three types of particles, neutron, prompt photon and slow photon, δi(t- τ) represents the impulse response function of the response current of the ith particle at the current burn-up point, xiRepresenting the input functions of three types of particles, neutron, prompt photon and slow photon.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201911028353.1A CN110765618B (en) | 2019-10-28 | 2019-10-28 | Method for calculating response current of self-powered neutron detector in pressurized water reactor |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN201911028353.1A CN110765618B (en) | 2019-10-28 | 2019-10-28 | Method for calculating response current of self-powered neutron detector in pressurized water reactor |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| CN110765618A CN110765618A (en) | 2020-02-07 |
| CN110765618B true CN110765618B (en) | 2021-04-13 |
Family
ID=69334092
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN201911028353.1A Active CN110765618B (en) | 2019-10-28 | 2019-10-28 | Method for calculating response current of self-powered neutron detector in pressurized water reactor |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN110765618B (en) |
Families Citing this family (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN111931356B (en) * | 2020-07-22 | 2023-08-04 | 上海核工程研究设计院股份有限公司 | Combined analysis method for response current of reactor core self-powered detector |
| CN112632771A (en) * | 2020-12-21 | 2021-04-09 | 中国核动力研究设计院 | Method and equipment for solving response sensitivity of fuel consumption calculation |
| CN113435096B (en) * | 2021-06-23 | 2022-04-19 | 兰州大学 | High-precision numerical model calculation method and system for steady-state response of self-powered detector |
| CN114783637B (en) * | 2022-04-20 | 2024-04-26 | 西安交通大学 | Method for calculating correction factors of pressurized water reactor multi-fuel cycle dynamic rod etching space |
| CN114722646B (en) * | 2022-06-10 | 2022-08-26 | 西安交通大学 | Method for optimizing three-dimensional measuring point arrangement of self-powered detector based on Kriging model |
Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| EP0496998A1 (en) * | 1991-01-30 | 1992-08-05 | Combustion Engineering, Inc. | Extending in-core instrument life |
| EP2338157B1 (en) * | 2008-10-13 | 2012-10-31 | Commissariat à l'Énergie Atomique et aux Énergies Alternatives | Device for the on line measurement of rapid and epithermal neutrons |
| CN103617816A (en) * | 2013-10-29 | 2014-03-05 | 中国广核集团有限公司 | Reactor core power distribution measuring method |
Family Cites Families (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN108899102A (en) * | 2018-07-06 | 2018-11-27 | 中国核动力研究设计院 | The method for arranging of the fixed self-power neutron detector general for presurized water reactor |
-
2019
- 2019-10-28 CN CN201911028353.1A patent/CN110765618B/en active Active
Patent Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| EP0496998A1 (en) * | 1991-01-30 | 1992-08-05 | Combustion Engineering, Inc. | Extending in-core instrument life |
| EP2338157B1 (en) * | 2008-10-13 | 2012-10-31 | Commissariat à l'Énergie Atomique et aux Énergies Alternatives | Device for the on line measurement of rapid and epithermal neutrons |
| CN103617816A (en) * | 2013-10-29 | 2014-03-05 | 中国广核集团有限公司 | Reactor core power distribution measuring method |
Non-Patent Citations (2)
| Title |
|---|
| 压水堆堆芯功率分布在线监测计算中的探测器失效诊断与处理方法;李茁，等;《强激光与粒子束》;20170131;第29卷(第1期);全文 * |
| 铑自给能探测器电流信号计算方法及影响因素;李树成，等;《核电子学与探测技术》;20110531;第31卷(第5期);全文 * |
Also Published As
| Publication number | Publication date |
|---|---|
| CN110765618A (en) | 2020-02-07 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| CN110765618B (en) | Method for calculating response current of self-powered neutron detector in pressurized water reactor | |
| Williams et al. | The embedded self-shielding method | |
| Wagner et al. | Review of hybrid (deterministic/Monte Carlo) radiation transport methods, codes, and applications at Oak Ridge National Laboratory | |
| CN108549753B (en) | Radiation shielding calculation method for coupling point kernel integration method and Monte Carlo method | |
| Jessee et al. | Polaris: a new two-dimensional lattice physics analysis capability for the SCALE code system | |
| CN104376217A (en) | Radiation shielding calculation method based on monte carlo self-adaptive variance reduction | |
| CN113139325B (en) | Implementation method for uniform distribution of global variance of reactor based on Monte Carlo critical calculation | |
| CN107038294A (en) | For the Resonance self-shielding computational methods based on equivalent one-dimensional rod model of light water reactor | |
| CN112069670B (en) | Monte Carlo simulation method for active critical device in critical approaching process | |
| Cao et al. | A two-step neutron spectrum unfolding method for fission reactors based on artificial neural network | |
| Kotnik et al. | Validation and evaluation of the ADVANTG code on the ICSBEP skyshine benchmark experiment | |
| CN117593474A (en) | Three-dimensional power distribution reconstruction method for pressurized water reactor core | |
| Ziani et al. | Eigenvalue sensitivity and nuclear data uncertainty analysis for the Moroccan TRIGA Mark II research reactor using SCALE6. 2 and MCNP6. 2 | |
| Yamamoto et al. | AEGIS: an advanced lattice physics code for light water reactor analyses | |
| CN116561987A (en) | Method for quantitatively evaluating instantaneous gamma response current of vanadium self-energized neutron detector | |
| Zhang et al. | Rasterized coarse mesh finite difference acceleration on method of characteristics for a reactor core with a generalized boundary | |
| Yang et al. | Comparison of neutron energy spectrum unfolding methods and evaluation of rationality criteria | |
| Smith et al. | An analysis of various solution strategies and perturbations on inputs of the reactor shielding problem | |
| Rabir et al. | Measurement and simulation of thermal neutron flux distribution in the RTP core | |
| Matijević et al. | PCA Benchmark Analysis with ADVANTG3. 0.1. and MCNP6. 1.1 b Codes | |
| Kang et al. | The advanced multilevel predictor-corrector quasi-static method for pin-resolved neutron kinetics simulation | |
| Roy et al. | Pulsed neutron source measurements in the BRAHMMA accelerator-driven subcritical system | |
| Zheng et al. | A Hybrid Monte-Carlo-Deterministic Method for AP1000 Ex-core Detector Response Simulation | |
| Gentry et al. | Source Range Detector Response Modeling Using VERA | |
| Larsen | An overview of neutron transport problems and simulation techniques |
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 |