CN113987784B - Method and device for quantifying uncertainty of acquisition of pressurized water reactor nuclear design software package - Google Patents
Method and device for quantifying uncertainty of acquisition of pressurized water reactor nuclear design software package Download PDFInfo
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Abstract
The inventionThe invention discloses a method and a device for quantifying uncertainty obtained by a pressurized water reactor nuclear design software package, wherein the method comprises the following steps: acquiring actual measurement data of a critical device and a power plant, and simultaneously acquiring calculation data of a physical design program; according to the acquired data, analyzing whether the data is a direct measurement parameter; if the parameter is the direct parameter, calculating the deviation statistics of the calculated value and the measured value by adopting a direct measurement parameter method, and calculating to obtain the uncertainty of the direct measurement parameter; if the data is the indirect measurement parameters, decomposing the data by adopting an indirect measurement parameter decomposition method, discarding the original unreasonable assumption, and obtaining the calculation uncertainty by a disturbance method. Compared with the prior art, the invention solves the problem of F Q 、F ΔH The problem that the assumption is unreasonable in the quantitative process of indirect measurement parameter uncertainty is solved, and a pressurized water reactor nuclear design software package confirmation system is perfected.
Description
Technical Field
The invention relates to the technical field of nuclear reactor cores, in particular to a method and a device for quantifying uncertainty of acquisition of a pressurized water reactor core design software package.
Background
The Validation (Validation) process of the software is a necessary process before the software is applied to the pressurized water reactor fuel design calculation and the safety analysis evaluation. Determining the uncertainty and the application scope of a computing program is a main purpose of software verification and validation. The pressurized water reactor nuclear design software mainly comprises a component program and a reactor core program. The verification of the assembly calculation procedure is mainly aimed at ensuring the correctness and range of the parameters used by the core (few-group homogenization constants and power reconstruction form factors), and is mainly carried out by measuring the parameters such as the reactivity parameter, neutron value, burnup, rod power distribution and the like of the core. Component calculation programIt is believed that the range of applicability will need to be determined for varying ranges of parameters such as different fuel assembly types, different enrichments, uranium ratios, boron concentrations, and poison rods. Verification and validation of core procedures is primarily aimed at giving the computational uncertainty of core parameters and their applicable ranges for reactor safety analysis calculations. The verification of the core procedure is mainly dependent on a comparison of calculated and measured values, which are typically derived from critical devices, plant measured data, etc. Most of the reactor core parameters in the actual measurement data of the power plant can be directly measured, including critical boron concentration, control rod integral/differential value, isothermal temperature coefficient, power distribution, heat pipe factor, enthalpy rise factor, reactor core cycle length, fuel assembly unloading burnup and the like, and the parameters can be directly used for carrying out mathematical statistics on calculated values and actual measurement deviation samples to obtain calculation uncertainty. Heat flux density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH The uncertainty calculation of the parameters is special, because the reactor core can not directly measure the parameters, only the reaction rate of the position of the detector can be measured, and the reaction rate is obtained by reconstruction of the actual measurement result. For such parameters, they can be decomposed into products of several experimentally measured parameters, and then the calculated uncertainty of the indirect quantities is synthesized by an uncertainty transfer formula. However, many core design procedures, such as the french SCIENCE system, currently assume that the uncertainty of the component average power P is equal to the uncertainty of the detector reactivity a measurement in the uncertainty calculation process, and that there is no rationality in this assumption.
Disclosure of Invention
The technical problem to be solved by the invention is that the heat flux density heat pipe factor F in the prior art Q Nuclear enthalpy heat rising pipe factor F ΔH F in equal parameter uncertainty acquisition method Q 、F ΔH The problem of unreasonable assumption exists in the quantitative process of indirect measurement parameter uncertainty. The invention aims to provide a method and a device for quantifying uncertainty obtained by a pressurized water reactor nuclear design software package, which aim at a heat flux density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH Indirect measurement parameters are measured, an original unreasonable assumption is abandoned by adopting a parameter decomposition method, and calculation is obtained by a disturbance methodUncertainty. Compared with the prior art, the invention can realize direct measurement of parameters and solve the problem of F Q 、F ΔH The problem that the assumption is unreasonable in the quantitative process of indirect measurement parameter uncertainty is solved, and a pressurized water reactor nuclear design software package confirmation system is perfected.
The invention is realized by the following technical scheme:
in a first aspect, the present invention provides a method for quantifying an uncertainty obtained by a pressurized water reactor core design software package, the method comprising:
acquiring actual measurement data of a critical device and a power plant, and simultaneously acquiring calculation data of a physical design program;
according to the acquired data, analyzing whether the data is a direct measurement parameter;
if the parameter is the direct parameter, calculating the deviation statistics of the calculated value and the measured value by adopting a direct measurement parameter method, and calculating to obtain the uncertainty of the direct measurement parameter;
if the measured parameter is the indirect measurement parameter, decomposing the data by adopting an indirect measurement parameter decomposition method to obtain a decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not again, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition amount is not a direct measurement parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining uncertainty that the decomposition amount is not the direct measurement parameter through obtaining a reaction rate, an average power proportion relation of a component, a correlation coefficient and the like through the disturbance data set;
and performing uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty.
The invention relates to a method for obtaining uncertainty by quantifying a pressurized water reactor nuclear design software package, which aims at direct measurement parameters such as critical boron concentration, control rod integral/differential value, isothermal temperature coefficient and the like, and carries out parameter statistics on absolute deviation of a measured value and a calculated value by adopting a direct measurement parameter method to obtain the calculated uncertainty; heat pipe factor F for heat flux density Q Nuclear enthalpy heat rising pipe factor F ΔH Indirect measurement parameters are equal, an indirect measurement parameter decomposition method is adopted, the original unreasonable assumption is abandoned,the computational uncertainty is obtained by a perturbation method. Compared with the prior art, the invention can realize direct measurement of parameters and solve the problem of F Q 、F ΔH The problem that the assumption is unreasonable in the quantitative process of indirect measurement parameters is solved, and a pressurized water reactor nuclear design software package (component calculation program and reactor core calculation program) confirmation system is perfected.
Further, the directly measured parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient, etc.
Further, the indirect measurement parameter is a heat flux density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH Etc.
Further, if the measured parameter is an indirect measurement parameter, decomposing the data by adopting an indirect measurement parameter decomposition method to obtain a decomposition amount; the method specifically comprises the following steps:
s10: converting the indirect measurement parameters into fuel rod power, decomposing, and heating the heat flow density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH The equivalent may be translated into a constant fuel rod power ratio (e.g.,
) The decomposition formula is:
wherein A is the reaction rate,
the average power of the components is that of the fuel rod, and P is that of the fuel rod;
s11: the indirect measurement parameters can be decomposed into three decomposition amounts through the above, and the uncertainty of the indirect measurement parameters is further decomposed into three parts: rod power distribution uncertainty
Response rate-component average Power conversion factor uncertainty +.>
Degree of uncertainty of reaction rate sigma c (A)；
Respectively solving uncertainty of the three decomposed parts; wherein the stick power distribution uncertainty
Can be obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the experimental actual measurement value of the critical device, and the uncertainty sigma of the reaction rate c (A) The three-dimensional reaction rate distribution of the detector, which can be calculated through the reactor core program, is obtained by comparing the three-dimensional reaction rate distribution of the detector obtained through calculation with the actual measurement of the three-dimensional reaction rate distribution of the power plant.
Further, if the decomposition amount is not a direct measurement parameter, performing disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining a reaction rate, a component average power proportion relation, a correlation coefficient and the like through the disturbance data set;
defining the correlation coefficient of the component average power and the detector reaction rate as r, and determining the uncertainty of the reaction rate-component average power conversion factor
Expressed as:
conventional methods assume that
Can directly obtain->
But this assumption is somewhat unreasonable. The present invention abandons this hypothesis and gets +.>
S20: in the reactor core program, by carrying out disturbance on reactor core status points (parameters such as boron concentration, temperature, power and the like), the module power corresponding to one can be calculated
And detector reaction rate { A } 1 ，A 2 ，……，A n Disturbance data set;
s21: from the disturbance data set
And->
Is a proportional relation and a correlation coefficient r;
s22: obtained by actually measuring reaction rate of power plant
Solving uncertainty of reaction rate-component average power conversion factor by proportional relation and correlation coefficient r obtained in step S21 +.>
Further, uncertainty synthesis is carried out, so that indirect measurement parameters are obtained, and uncertainty is calculated; comprising the following steps:
from bar power distribution uncertainty
Response rate-component average power conversion factor uncertainty
Degree of uncertainty of reaction rate sigma c (A) Synthesizing uncertainty to obtain F Q Uncertainty of (2), namely:
the above-mentioned F can be obtained Q Uncertainty of F ΔH Can be obtained by a similar method, except that a two-dimensional radial reaction rate distribution is used.
In a second aspect, the present invention further provides an apparatus for quantifying an uncertainty obtained by a pressurized water reactor core design software package, where the apparatus supports a method for quantifying an uncertainty obtained by a pressurized water reactor core design software package, where the apparatus includes:
the acquisition unit is used for acquiring actual measurement data of the critical device and the power plant and simultaneously acquiring calculation data of a physical design program;
the data judging unit is used for analyzing whether the data is a direct measurement parameter according to the acquired data;
the uncertainty calculation unit of the direct measurement parameter is used for calculating the uncertainty of the direct measurement parameter by adopting a direct measurement parameter method to calculate the deviation between a calculated value and a measured value according to the fact that the data are the direct parameter according to the judgment of the data judgment unit;
an uncertainty calculating unit indirectly connected with the measurement parameters, which is used for decomposing the data by adopting an indirect measurement parameter decomposition method according to the judgment of the data judging unit that the data is the indirect measurement parameters to obtain the decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not again, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition amount is not a direct measurement parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining uncertainty that the decomposition amount is not the direct measurement parameter through obtaining a reaction rate, an average power proportion relation of a component, a correlation coefficient and the like through the disturbance data set;
and the uncertainty synthesis unit is used for performing uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty.
Further, the direct measurement parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient and the like;
the saidThe indirect measurement parameter is the heat flux density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH Etc.
Further, the execution process of the uncertainty calculation unit of the indirect measurement parameter is as follows:
converting the indirect measurement parameters into fuel rod power, decomposing, and heating the heat flow density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH The equivalent may be translated into a constant fuel rod power ratio (e.g.,
) The decomposition formula is:
wherein A is the reaction rate,
the average power of the components is that of the fuel rod, and P is that of the fuel rod;
the indirect measurement parameters can be decomposed into three decomposition amounts through the above, and the uncertainty of the indirect measurement parameters is further decomposed into three parts: rod power distribution uncertainty
Response rate-component average Power conversion factor uncertainty +.>
Degree of uncertainty of reaction rate sigma c (A)；
Respectively solving uncertainty of the three decomposed parts; wherein the stick power distribution uncertainty
Can be obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the experimental actual measurement value of the critical device, and the uncertainty sigma of the reaction rate c (A) Can pass through the reactor core programAnd comparing the calculated three-dimensional reaction rate distribution of the detector with the actually measured three-dimensional reaction rate distribution of the power plant.
Further, in the uncertainty calculation unit of the indirect connection measurement parameter, if the decomposition amount is not the direct measurement parameter, performing disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining a reaction rate, a component average power proportion relation, a correlation coefficient and the like through the disturbance data set; the specific implementation process is as follows:
in the reactor core program, by carrying out disturbance on reactor core status points (parameters such as boron concentration, temperature, power and the like), the module power corresponding to one can be calculated
And detector reaction rate { A } 1 ，A 2 ，……，A n Disturbance data set;
from the disturbance data set
And->
Is a proportional relation and a correlation coefficient r;
obtained by actually measuring reaction rate of power plant
Solving uncertainty of reaction rate-component average power conversion factor by proportional relation and correlation coefficient r obtained in step S21 +.>
Compared with the prior art, the invention has the following advantages and beneficial effects:
the method and the device of the invention adopt a direct measurement parameter method to carry out parameter statistics on the absolute deviation of the measured value and the calculated value aiming at the direct measurement parameters such as critical boron concentration, control rod integral/differential value, isothermal temperature coefficient and the like to obtain calculation uncertaintyDegree of certainty; heat pipe factor F for heat flux density Q Nuclear enthalpy heat rising pipe factor F ΔH And (3) indirectly measuring parameters, adopting an indirect measurement parameter decomposition method, discarding the original unreasonable assumption, and obtaining the calculation uncertainty through a disturbance method. Compared with the prior art, the invention can realize direct measurement of parameters and solve the problem of F Q 、F ΔH The problem that the assumption is unreasonable in the quantitative process of indirect measurement parameters is solved, and a pressurized water reactor nuclear design software package (component calculation program and reactor core calculation program) confirmation system is perfected.
Drawings
The accompanying drawings, which are included to provide a further understanding of embodiments of the invention and are incorporated in and constitute a part of this application, illustrate embodiments of the invention. In the drawings:
FIG. 1 is a flow chart of a method for quantifying the uncertainty obtained by a pressurized water reactor core design software package according to the present invention.
FIG. 2 is a block diagram of an apparatus for quantifying uncertainty in acquisition of a pressurized water reactor core design software package in accordance with the present invention.
Detailed Description
For the purpose of making apparent the objects, technical solutions and advantages of the present invention, the present invention will be further described in detail with reference to the following examples and the accompanying drawings, wherein the exemplary embodiments of the present invention and the descriptions thereof are for illustrating the present invention only and are not to be construed as limiting the present invention.
Example 1
As shown in fig. 1, fig. 1 is a flowchart of a method for quantifying uncertainty obtained by a pressurized water reactor core design software package according to the present invention, namely, a logic flowchart of a method for quantifying uncertainty calculated by the pressurized water reactor core design software package;
the invention discloses a method for quantifying uncertainty obtained by a pressurized water reactor nuclear design software package, which comprises the following steps:
acquiring actual measurement data of a critical device and a power plant, and simultaneously acquiring calculation data of a physical design program;
according to the acquired data, analyzing whether the data is a direct measurement parameter;
if the parameter is the direct parameter, calculating the deviation statistics of the calculated value and the measured value by adopting a direct measurement parameter method, and calculating to obtain the uncertainty of the direct measurement parameter; wherein the directly measured parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient, etc.
If the measured parameter is the indirect measurement parameter, decomposing the data by adopting an indirect measurement parameter decomposition method to obtain a decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not again, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition amount is not a direct measurement parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining uncertainty that the decomposition amount is not the direct measurement parameter through obtaining a reaction rate, an average power proportion relation of a component, a correlation coefficient and the like through the disturbance data set; wherein the indirect measurement parameter is a heat flux density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH Etc.
And performing uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty.
Specifically, if the measured parameter is an indirect measurement parameter, decomposing the data by adopting an indirect measurement parameter decomposition method to obtain a decomposition amount; the method specifically comprises the following steps:
s10: converting the indirect measurement parameters into fuel rod power, decomposing, and heating the heat flow density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH The equivalent may be translated into a constant fuel rod power ratio (e.g.,
) The decomposition formula is:
wherein A is the reaction rate,
the average power of the components is that of the fuel rod, and P is that of the fuel rod;
s11: the indirect measurement parameters can be decomposed into three decomposition amounts through the above, and the uncertainty of the indirect measurement parameters is further decomposed into three parts: rod power distribution uncertainty
Response rate-component average Power conversion factor uncertainty +.>
Degree of uncertainty of reaction rate sigma c (A)；
Respectively solving uncertainty of the three decomposed parts; wherein the stick power distribution uncertainty
Can be obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the experimental actual measurement value of the critical device, and the uncertainty sigma of the reaction rate c (A) The three-dimensional reaction rate distribution of the detector, which can be calculated through the reactor core program, is obtained by comparing the three-dimensional reaction rate distribution of the detector obtained through calculation with the actual measurement of the three-dimensional reaction rate distribution of the power plant.
Specifically, if the decomposition amount is not a direct measurement parameter, performing disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining a reaction rate, a component average power proportion relation, a correlation coefficient and the like through the disturbance data set;
defining the correlation coefficient of the component average power and the detector reaction rate as r, and determining the uncertainty of the reaction rate-component average power conversion factor
Expressed as:
conventional methods assume that
Can directly obtain->
But this assumption is somewhat unreasonable. The present invention abandons this hypothesis and gets +.>
S20: in the reactor core program, by carrying out disturbance on reactor core status points (parameters such as boron concentration, temperature, power and the like), the module power corresponding to one can be calculated
And detector reaction rate { A } 1 ，A 2 ，……，A n Disturbance data set;
s21: from the disturbance data set
And->
Is a proportional relation and a correlation coefficient r;
s22: obtained by actually measuring reaction rate of power plant
Solving uncertainty of reaction rate-component average power conversion factor by proportional relation and correlation coefficient r obtained in step S21 +.>
Specifically, uncertainty synthesis is carried out, so that indirect measurement parameters are obtained, and uncertainty is calculated; comprising the following steps:
from bar power distribution uncertainty
Response rate-component average power conversion factor uncertainty
Degree of uncertainty of reaction rate sigma c (A) Synthesizing uncertainty to obtain F Q Uncertainty of (2), namely:
the above-mentioned F can be obtained Q Uncertainty of F ΔH Can be obtained by a similar method, except that a two-dimensional radial reaction rate distribution is used.
The invention relates to a method for obtaining uncertainty by quantifying a pressurized water reactor nuclear design software package, which aims at direct measurement parameters such as critical boron concentration, control rod integral/differential value, isothermal temperature coefficient and the like, and carries out parameter statistics on absolute deviation of a measured value and a calculated value by adopting a direct measurement parameter method to obtain the calculated uncertainty; heat pipe factor F for heat flux density Q Nuclear enthalpy heat rising pipe factor F ΔH And (3) indirectly measuring parameters, adopting an indirect measurement parameter decomposition method, discarding the original unreasonable assumption, and obtaining the calculation uncertainty through a disturbance method. Compared with the prior art, the invention can realize direct measurement of parameters and solve the problem of F Q 、F ΔH The problem that the assumption is unreasonable in the quantitative process of indirect measurement parameters is solved, and a pressurized water reactor nuclear design software package (component calculation program and reactor core calculation program) confirmation system is perfected.
Example 2
As shown in fig. 2, the difference between the present embodiment and embodiment 1 is that the present embodiment provides a device for quantifying the uncertainty of obtaining a pressurized water reactor core design software package, which supports the method for quantifying the uncertainty of obtaining a pressurized water reactor core design software package described in embodiment 1, and the device includes:
the acquisition unit is used for acquiring actual measurement data of the critical device and the power plant and simultaneously acquiring calculation data of a physical design program;
the data judging unit is used for analyzing whether the data is a direct measurement parameter according to the acquired data;
the uncertainty calculation unit of the direct measurement parameter is used for calculating the uncertainty of the direct measurement parameter by adopting a direct measurement parameter method to calculate the deviation between a calculated value and a measured value according to the fact that the data are the direct parameter according to the judgment of the data judgment unit;
an uncertainty calculating unit indirectly connected with the measurement parameters, which is used for decomposing the data by adopting an indirect measurement parameter decomposition method according to the judgment of the data judging unit that the data is the indirect measurement parameters to obtain the decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not again, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition amount is not a direct measurement parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining uncertainty that the decomposition amount is not the direct measurement parameter through obtaining a reaction rate, an average power proportion relation of a component, a correlation coefficient and the like through the disturbance data set;
and the uncertainty synthesis unit is used for performing uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty.
For further explanation of this embodiment, the directly measured parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient, etc.;
the indirect measurement parameter is heat flux density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH Etc.
For further explanation of the present embodiment, the procedure of the uncertainty calculation unit of the indirect connection measurement parameter is as follows:
converting the indirect measurement parameters into fuel rod power, decomposing, and heating the heat flow density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH The equivalent may be translated into a constant fuel rod power ratio (e.g.,
) The decomposition formula is:
wherein A is the reaction rate,
the average power of the components is that of the fuel rod, and P is that of the fuel rod;
the indirect measurement parameters can be decomposed into three decomposition amounts through the above, and the uncertainty of the indirect measurement parameters is further decomposed into three parts: rod power distribution uncertainty
Response rate-component average Power conversion factor uncertainty +.>
Degree of uncertainty of reaction rate sigma c (A)；
Respectively solving uncertainty of the three decomposed parts; wherein the stick power distribution uncertainty
Can be obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the experimental actual measurement value of the critical device, and the uncertainty sigma of the reaction rate c (A) The three-dimensional reaction rate distribution of the detector, which can be calculated through the reactor core program, is obtained by comparing the three-dimensional reaction rate distribution of the detector obtained through calculation with the actual measurement of the three-dimensional reaction rate distribution of the power plant.
In order to further describe the embodiment, in the uncertainty calculation unit of the indirect connection measurement parameter, if the decomposition amount is not the direct measurement parameter, performing disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining a reaction rate, a component average power proportional relationship, a correlation coefficient and the like through the disturbance data set;
defining the correlation coefficient of the component average power and the detector reaction rate as r, and determining the uncertainty of the reaction rate-component average power conversion factor
Expressed as:
conventional methods assume that
Can directly obtain->
But this assumption is somewhat unreasonable. The present invention abandons this hypothesis and gets +.>
In the reactor core program, by carrying out disturbance on reactor core status points (parameters such as boron concentration, temperature, power and the like), the module power corresponding to one can be calculated
And detector reaction rate { A } 1 ,A 2 ,……,A n Disturbance data set;
from the disturbance data set
And->
Is a proportional relation and a correlation coefficient r;
obtained by actually measuring reaction rate of power plant
Solving uncertainty of reaction rate-component average power conversion factor by proportional relation and correlation coefficient r obtained in step S21 +.>
The device adopts a direct measurement parameter method to carry out parameter statistics on absolute deviation between a measured value and a calculated value aiming at direct measurement parameters such as critical boron concentration, control rod integral/differential value, isothermal temperature coefficient and the like to obtain calculation uncertainty; heat pipe factor F for heat flux density Q Nuclear enthalpy heat rising pipe factor F ΔH And (3) indirectly measuring parameters, adopting an indirect measurement parameter decomposition method, discarding the original unreasonable assumption, and obtaining the calculation uncertainty through a disturbance method. Compared with the prior art, the invention can realize direct measurement of parameters and solve the problem of F Q 、F ΔH The problem that the assumption is unreasonable in the quantitative process of indirect measurement parameters is solved, and a pressurized water reactor nuclear design software package (component calculation program and reactor core calculation program) confirmation system is perfected.
It will be appreciated by those skilled in the art that embodiments of the present application may be provided as a method, system, or computer program product. Accordingly, the present application may take the form of an entirely hardware embodiment, an entirely software embodiment, or an embodiment combining software and hardware aspects. Furthermore, the present application may take the form of a computer program product embodied on one or more computer-usable storage media (including, but not limited to, disk storage, CD-ROM, optical storage, and the like) having computer-usable program code embodied therein.
The present application is described with reference to flowchart illustrations and/or block diagrams of methods, apparatus (systems) and computer program products according to embodiments of the application. It will be understood that each flow and/or block of the flowchart illustrations and/or block diagrams, and combinations of flows and/or blocks in the flowchart illustrations and/or block diagrams, can be implemented by computer program instructions. These computer program instructions may be provided to a processor of a general purpose computer, special purpose computer, embedded processor, or other programmable data processing apparatus to produce a machine, such that the instructions, which execute via the processor of the computer or other programmable data processing apparatus, create means for implementing the functions specified in the flowchart flow or flows and/or block diagram block or blocks.
These computer program instructions may also be stored in a computer-readable memory that can direct a computer or other programmable data processing apparatus to function in a particular manner, such that the instructions stored in the computer-readable memory produce an article of manufacture including instruction means which implement the function specified in the flowchart flow or flows and/or block diagram block or blocks.
These computer program instructions may also be loaded onto a computer or other programmable data processing apparatus to cause a series of operational steps to be performed on the computer or other programmable apparatus to produce a computer implemented process such that the instructions which execute on the computer or other programmable apparatus provide steps for implementing the functions specified in the flowchart flow or flows and/or block diagram block or blocks.
The foregoing description of the embodiments has been provided for the purpose of illustrating the general principles of the invention, and is not meant to limit the scope of the invention, but to limit the invention to the particular embodiments, and any modifications, equivalents, improvements, etc. that fall within the spirit and principles of the invention are intended to be included within the scope of the invention.
Claims (4)
1. A method for quantifying the uncertainty of a pressurized water reactor core design software package acquisition, the method comprising:
acquiring actual measurement data of a critical device and a power plant, and simultaneously acquiring calculation data of a physical design program;
according to the acquired data, analyzing whether the data is a direct measurement parameter;
if the measured value is the direct measurement parameter, calculating the deviation statistics of the calculated value and the measured value by adopting a direct measurement parameter method, and calculating to obtain the uncertainty of the direct measurement parameter;
if the measured parameter is the indirect measurement parameter, decomposing the data by adopting an indirect measurement parameter decomposition method to obtain a decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not again, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition amount is not a direct measurement parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining the uncertainty that the decomposition amount is not the direct measurement parameter through the disturbance data set to obtain the reaction rate, the component average power proportional relation and the correlation coefficient;
performing uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty;
the direct measurement parameters comprise critical boron concentration, control rod integral/differential value and isothermal temperature coefficient;
the indirect measurement parameter is heat flux density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH ；
If the data is the indirect measurement parameter, decomposing the data by adopting an indirect measurement parameter decomposition method to obtain a decomposition amount; the method specifically comprises the following steps:
s10: converting the indirect measurement parameters into fuel rod power and then decomposing the fuel rod power, wherein a decomposition formula is as follows:
wherein A is the reaction rate,
the average power of the components is that of the fuel rod, and P is that of the fuel rod;
s11: decomposing the indirect measurement parameter into three decomposition amounts through the above, and then decomposing the uncertainty of the indirect measurement parameter into three parts: rod power distribution uncertainty
Response rate-component average Power conversion factor uncertainty +.>
Degree of uncertainty of reaction rate sigma c (A)；
The uncertainty synthesis is carried out, so that indirect measurement parameters are obtained, and uncertainty is calculated; comprising the following steps:
from bar power distribution uncertainty
Response rate-component average Power conversion factor uncertainty +.>
Degree of uncertainty of reaction rate sigma c (A) Synthesizing uncertainty to obtain F Q Uncertainty of (2), namely:
2. the method for obtaining uncertainty by using a quantitative pressurized water reactor nuclear design software package according to claim 1, wherein if the decomposition amount is not a direct measurement parameter, performing disturbance calculation on a reactor core state to obtain a disturbance data set, and obtaining a reaction rate, a component average power proportional relation and a correlation coefficient through the disturbance data set; the method specifically comprises the following substeps:
s20: in the reactor core program, the disturbance is carried out on reactor core status points, so that the one-to-one corresponding assembly power is calculated
And detector reaction rate { A } 1 ,A 2 ,……,A n Disturbance data set;
s21: from the disturbance data set
And->
Ratio of (3)Relationship and correlation coefficient r;
s22: obtained by actually measuring reaction rate of power plant
Solving uncertainty of reaction rate-component average power conversion factor by proportional relation and correlation coefficient r obtained in step S21 +.>
3. An apparatus for quantifying pressurized water reactor core design package acquisition uncertainty, the apparatus supporting a method for quantifying pressurized water reactor core design package acquisition uncertainty as recited in any one of claims 1-2, the apparatus comprising:
the acquisition unit is used for acquiring actual measurement data of the critical device and the power plant and simultaneously acquiring calculation data of a physical design program;
the data judging unit is used for analyzing whether the data is a direct measurement parameter according to the acquired data;
the uncertainty calculation unit of the direct measurement parameter is used for calculating the uncertainty of the direct measurement parameter by adopting a direct measurement parameter method to calculate the deviation between a calculated value and a measured value according to the judgment that the data is the direct measurement parameter by the data judgment unit;
an uncertainty calculation unit of the indirect measurement parameters, configured to decompose the data by using an indirect measurement parameter decomposition method according to the judgment of the data judgment unit that the data is the indirect measurement parameters, so as to obtain a decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not again, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition amount is not a direct measurement parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining the uncertainty that the decomposition amount is not the direct measurement parameter through the disturbance data set to obtain the reaction rate, the component average power proportional relation and the correlation coefficient;
uncertainty synthesis unit, which synthesizes uncertainty to obtain indirect measurement parameters and calculate uncertainty;
the direct measurement parameters comprise critical boron concentration, control rod integral/differential value and isothermal temperature coefficient;
the indirect measurement parameter is heat flux density heat pipe factor F Q Nuclear enthalpy heat rising pipe factor F ΔH ；
The uncertainty calculation unit of the indirect measurement parameter performs the following steps:
converting the indirect measurement parameters into fuel rod power and then decomposing the fuel rod power, wherein a decomposition formula is as follows:
wherein A is the reaction rate,
the average power of the components is that of the fuel rod, and P is that of the fuel rod;
decomposing the indirect measurement parameter into three decomposition amounts through the above, and then decomposing the uncertainty of the indirect measurement parameter into three parts: rod power distribution uncertainty
Response rate-component average power conversion factor uncertainty
Degree of uncertainty of reaction rate sigma c (A)；
Respectively solving uncertainty of the three decomposed parts; wherein the stick power distribution uncertainty
The reactor core program power distribution is obtained by comparing the rod power after reconstruction with the experimental actual measurement value of the critical device, and the reaction rate is not highDegree of certainty sigma c (A) The three-dimensional reaction rate distribution of the detector obtained through calculation of the reactor core program is obtained through comparison with the actual measurement three-dimensional reaction rate distribution of the power plant;
the uncertainty synthesis unit performs the following steps:
from bar power distribution uncertainty
Response rate-component average Power conversion factor uncertainty +.>
Degree of uncertainty of reaction rate sigma c (A) Synthesizing uncertainty to obtain F Q Uncertainty of (2), namely:
4. the device for obtaining uncertainty by quantifying a pressurized water reactor nuclear design software package according to claim 3, wherein in the uncertainty calculation unit of the indirect measurement parameter, if the decomposition amount is not a direct measurement parameter, performing disturbance calculation on a reactor core state to obtain a disturbance data set, and obtaining a reaction rate, a component average power proportional relation and a correlation coefficient through the disturbance data set; the specific implementation process is as follows:
in the reactor core program, the disturbance is carried out on reactor core status points, so that the one-to-one corresponding assembly power is calculated
And detector reaction rate { A } 1 ,A 2 ,……,A n Disturbance data set;
from the disturbance data set
And->
Is a proportional relation and a correlation coefficient r;
obtained by actually measuring reaction rate of power plant
Solving uncertainty of reaction rate-component average power conversion factor by proportional relation and correlation coefficient r obtained in step S21 +.>
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