CN113987784A - Method and device for obtaining uncertainty by quantifying pressurized water reactor nuclear design software package - Google Patents

Method and device for obtaining uncertainty by quantifying pressurized water reactor nuclear design software package Download PDF

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CN113987784A
CN113987784A CN202111246981.4A CN202111246981A CN113987784A CN 113987784 A CN113987784 A CN 113987784A CN 202111246981 A CN202111246981 A CN 202111246981A CN 113987784 A CN113987784 A CN 113987784A
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uncertainty
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CN113987784B (en
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彭星杰
吴屈
张斌
周冰燕
赵文博
卢宗健
李庆
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Nuclear Power Institute of China
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Abstract

The 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 actually measured data of a critical device and a power plant, and acquiring physical design program calculation data; analyzing whether the data is a direct measurement parameter or not according to the acquired data; if the parameter is a direct parameter, calculating the deviation between a calculated value and a measured value by adopting a direct parameter measurement method, and calculating to obtain the uncertainty of the direct parameter; if the data are indirectly measured parameters, an indirect measurement parameter decomposition method is adopted to decompose the data, the original unreasonable assumption is abandoned, and the calculation uncertainty is obtained through a disturbance method. Compared with the prior art, the invention solves the problem of FQ、FΔHThe unreasonable assumption problem in the process of quantifying the uncertainty of the parameters of the equal indirect measurement is completedIt is good for pressurized water reactor nuclear design software package confirmation system.

Description

Method and device for obtaining uncertainty by quantifying pressurized water reactor nuclear design software package
Technical Field
The invention relates to the technical field of nuclear reactor cores, in particular to a method and a device for obtaining uncertainty by a quantitative pressurized water reactor nuclear design software package.
Background
The Validation (Validation) process of the software is a necessary process before the software is applied to the calculation of the pressurized water reactor refueling design and the safety analysis evaluation. The uncertainty and applicability of the computational procedure are the primary goals of software validation and validation. The pressurized water reactor nuclear design software mainly comprises a component program and a reactor core program. The component calculation program is mainly used for confirming the correctness and the range of the parameters (the small group homogenization constant and the power reconstruction shape factor) used by the reactor core, and is mainly carried out by measuring the parameters such as reactivity parameters, neutron value, burnup, rod power distribution and the like of the reactor core. The confirmation of the component calculation program needs to determine the application range of the fuel component according to the variation ranges of parameters such as different fuel component types, different enrichment degrees, water-uranium ratios, boron concentrations, poison rods and the like. The main purpose of core procedure verification and validation is to give the uncertainty of the calculation of the core parameters and its applicability for the reactor safety analysis calculation. The confirmation of the core procedure mainly depends on the comparison of the calculated values and the measured values, which are usually derived from critical equipment, power plant measured data, and the like. The majority of reactor core parameters in the power plant measured data can be directly measured, and comprise 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 fuel consumption and the like, and the parameters can directly carry out mathematical statistics on a calculated value and a measured deviation sample to obtain calculation uncertainty. Heat flux density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHThe calculation of the uncertainty of the isoparametric is particularly peculiar, since the reactor core cannot directly measure these parameters, but only the reaction of obtaining the position of the detectorAnd the rate is obtained by reconstruction according to the actual measurement result. For such parameters, it can be decomposed into the product of several experimentally measured parameters, and the calculated uncertainty of the indirect quantity is synthesized by the uncertainty transfer formula. However, in the current core nuclear design programs, such as the French SCIENCE system, the uncertainty of the average power P of the assembly is assumed to be equal to the uncertainty of the measurement of the response rate A of the detector in the uncertainty calculation process, and the assumption is unreasonable.
Disclosure of Invention
The invention aims to solve the technical problem that the heat flux density heat pipe factor F in the prior artQNuclear enthalpy rising heat pipe factor FΔHEqual parameter uncertainty acquisition method FQ、FΔHThe problem of unreasonable assumption exists in the process of quantifying the uncertainty of the indirect measurement parameters. The invention aims to provide a method and a device for quantifying uncertainty obtained by a pressurized water reactor nuclear design software package, and aims at a heat flux density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHAnd (3) indirectly measuring parameters, adopting a parameter decomposition method, abandoning the original unreasonable assumption, and obtaining the uncertainty of calculation through a disturbance method. Compared with the prior art, the invention can realize direct parameter measurement and solve the problem of FQ、FΔHAnd the problem of unreasonable assumption in the process of quantifying the uncertainty of the indirect measurement parameters is solved, and a software package confirmation system for the nuclear design of the pressurized water reactor 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 from a PWR nuclear design software package, the method comprising:
acquiring actually measured data of a critical device and a power plant, and acquiring physical design program calculation data;
analyzing whether the data is a direct measurement parameter or not according to the acquired data;
if the parameter is a direct parameter, calculating the deviation between a calculated value and a measured value by adopting a direct parameter measurement method, and calculating to obtain the uncertainty of the direct parameter;
if the data are indirect measurement parameters, 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, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition quantity is not a directly measured parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining the uncertainty of the decomposition quantity which is not a directly measured parameter by obtaining the reaction rate, the component average power proportional relation, the correlation coefficient and the like through the disturbance data set;
and carrying out uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty.
The invention relates to a method for obtaining uncertainty by a quantitative pressurized water reactor nuclear design software package, which is characterized in that direct measurement parameters such as critical boron concentration, control rod integral/differential value, isothermal temperature coefficient and the like are subjected to parameter statistics by adopting a direct measurement parameter method to obtain the calculated uncertainty by carrying out parameter statistics on the absolute deviation of a measured value and a calculated value; heat pipe factor F for heat flux densityQNuclear enthalpy rising heat pipe factor FΔHAnd (3) waiting for indirect measurement parameters, adopting an indirect measurement parameter decomposition method, abandoning the original unreasonable assumption, and obtaining the uncertainty of calculation through a disturbance method. Compared with the prior art, the invention can realize direct parameter measurement and simultaneously solve the problem of FQ、FΔHAnd the method solves the problem of unreasonable assumption in the quantitative process of indirect measurement parameter uncertainty, and perfects a validation system of a pressurized water reactor nuclear design software package (component calculation program and reactor core calculation program).
Further, the directly measured parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient, and the like.
Further, the indirect measurement parameter is heat flow density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHAnd the like.
Further, if the measured parameter is an indirect measured parameter, decomposing the data by adopting an indirect measured 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 decomposing the fuel rod power to obtain heat flow density heat pipe factor FQNuclear enthalpy raising heat pipeFactor FΔHEtc. may be translated to a fixed value on the fuel rod power ratio (e.g.,
Figure BDA0003321180260000021
) The decomposition formula is:
Figure BDA0003321180260000022
wherein A is the reaction rate,
Figure BDA0003321180260000023
is the assembly average power, P is the fuel rod power;
s11: the indirect measurement parameter can be decomposed into three decomposition quantities by the above formula, and then the uncertainty of the indirect measurement parameter is decomposed into three parts: uncertainty of rod power distribution
Figure BDA0003321180260000031
Reactivity-component average power conversion factor uncertainty
Figure BDA0003321180260000032
Uncertainty of reaction rate σc(A)；
Respectively calculating the uncertainty of the three decomposed parts; wherein the rod power distribution is not deterministic
Figure BDA0003321180260000033
The reaction rate uncertainty sigma can be obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the actually measured value of the critical device testc(A) And comparing the three-dimensional reactivity distribution of the detector obtained by the calculation of the reactor core program with the actually measured three-dimensional reactivity distribution of the power plant.
Further, if the decomposition quantity is not a direct measurement parameter, performing disturbance calculation on the state of the reactor core to obtain a disturbance data set, and obtaining the reaction rate, the component average power proportional relation, the correlation coefficient and the like through the disturbance data set;
defining the correlation coefficient of the average power of the component and the response rate of the detector as r, and converting the response rate to the average power of the component by the uncertainty of the conversion factor
Figure BDA0003321180260000034
Expressed as:
Figure BDA0003321180260000035
conventional method assumptions
Figure BDA0003321180260000036
Can be directly obtained
Figure BDA0003321180260000037
But there is some irrationality to this assumption. The present invention abandons this assumption and obtains it by the following method
Figure BDA0003321180260000038
S20: in the reactor core procedure, the one-to-one corresponding component power can be calculated by disturbing the reactor core state points (parameters such as boron concentration, temperature and power)
Figure BDA0003321180260000039
And rate of detector reaction { A1，A2，……，AnDisturbing the data set;
s21: derived from the perturbation data set
Figure BDA00033211802600000311
And
Figure BDA00033211802600000312
and the correlation coefficient r;
s22: obtained by using actually measured reaction rate of power plant
Figure BDA00033211802600000313
Through step S21The obtained proportional relation and the related coefficient r solve the uncertainty of the reaction rate-component average power conversion factor
Figure BDA00033211802600000314
Further, performing uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty; the method comprises the following steps:
according to uncertainty of rod power distribution
Figure BDA00033211802600000315
Reactivity-component average power conversion factor uncertainty
Figure BDA00033211802600000316
Uncertainty of reaction rate σc(A) Performing uncertainty synthesis to obtain FQI.e.:
Figure BDA0003321180260000041
this gives FQUncertainty of (D), FΔHCan be obtained by a similar method except that a two-dimensional radial reactivity distribution is used.
In a second aspect, the present invention further provides an apparatus for quantifying the uncertainty obtained by a pressurized water reactor nuclear design software package, the apparatus supporting the method for quantifying the uncertainty obtained by a pressurized water reactor nuclear design software package, the apparatus comprising:
the acquisition unit is used for acquiring actually measured data of the critical device and the power plant and acquiring physical design program calculation data;
the data judging unit is used for analyzing whether the data are direct measurement parameters or not 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 carry out calculation value and measured value deviation statistics according to the fact that the data judged by the data judgment unit is the direct parameter;
the uncertainty calculation unit is indirectly connected with the measurement parameters and is used for decomposing the data by adopting an indirect measurement parameter decomposition method according to the fact that the data judged by the data judgment unit is the indirect measurement parameters to obtain the decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition quantity is not a directly measured parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining the uncertainty of the decomposition quantity which is not a directly measured parameter by obtaining the reaction rate, the component average power proportional relation, the 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 directly measured parameters comprise critical boron concentration, control rod integral/differential value, isothermal temperature coefficient and the like;
the indirect measurement parameter is heat flow density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHAnd the like.
Further, the uncertainty calculation unit for indirectly measuring the parameters performs the following steps:
converting the indirect measurement parameters into fuel rod power, and decomposing the fuel rod power to obtain heat flow density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHEtc. may be translated to a fixed value on the fuel rod power ratio (e.g.,
Figure BDA0003321180260000042
) The decomposition formula is:
Figure BDA0003321180260000043
wherein A is the reaction rate,
Figure BDA0003321180260000044
is the assembly average power, P is the fuel rod power;
the indirect measurement parameter can be decomposed into three decomposition quantities by the above formula, and then the uncertainty of the indirect measurement parameter is decomposed into three parts: uncertainty of rod power distribution
Figure BDA0003321180260000045
Reactivity-component average power conversion factor uncertainty
Figure BDA0003321180260000051
Uncertainty of reaction rate σc(A)；
Respectively calculating the uncertainty of the three decomposed parts; wherein the rod power distribution is not deterministic
Figure BDA0003321180260000052
The reaction rate uncertainty sigma can be obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the actually measured value of the critical device testc(A) And comparing the three-dimensional reactivity distribution of the detector obtained by the calculation of the reactor core program with the actually measured three-dimensional reactivity distribution of the power plant.
Further, in the uncertainty calculation unit of the indirect measurement parameters, if the decomposition quantity is not a direct measurement parameter, performing disturbance calculation on the state of the reactor core to obtain a disturbance data set, and obtaining the reaction rate, the component average power proportional relation, the correlation coefficient and the like through the disturbance data set; the specific implementation process is as follows:
in the reactor core procedure, the one-to-one corresponding component power can be calculated by disturbing the reactor core state points (parameters such as boron concentration, temperature and power)
Figure BDA0003321180260000053
And rate of detector reaction { A1，A2，……，AnDisturbing the data set;
derived from the perturbation data set
Figure BDA0003321180260000055
And
Figure BDA0003321180260000056
and the correlation coefficient r;
obtained by using actually measured reaction rate of power plant
Figure BDA0003321180260000057
Solving the uncertainty of the average power conversion factor of the response rate and the component through the proportional relation and the correlation coefficient r obtained in the step S21
Figure BDA0003321180260000058
Compared with the prior art, the invention has the following advantages and beneficial effects:
the method and the device adopt a direct measurement parameter method to carry out parameter statistics on the 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 the calculation uncertainty; heat pipe factor F for heat flux densityQNuclear enthalpy rising heat pipe factor FΔHAnd (3) waiting for indirect measurement parameters, adopting an indirect measurement parameter decomposition method, abandoning the original unreasonable assumption, and obtaining the uncertainty of calculation through a disturbance method. Compared with the prior art, the invention can realize direct parameter measurement and simultaneously solve the problem of FQ、FΔHAnd the method solves the problem of unreasonable assumption in the quantitative process of indirect measurement parameter uncertainty, and perfects a validation system of a pressurized water reactor nuclear design software package (component calculation program and reactor core calculation program).
Drawings
The accompanying drawings, which are included to provide a further understanding of the embodiments of the invention and are incorporated in and constitute a part of this application, illustrate embodiment(s) of the invention and together with the description serve to explain the principles of the invention. In the drawings:
FIG. 1 is a flow chart of a method for quantifying the uncertainty obtained by a PWR core design software package in accordance with the present invention.
FIG. 2 is a diagram of a device for quantifying uncertainty obtained by a PWR core design software package in accordance with the present invention.
Detailed Description
In order to make the objects, technical solutions and advantages of the present invention more apparent, the present invention is further described in detail below with reference to examples and accompanying drawings, and the exemplary embodiments and descriptions thereof are only used for explaining the present invention and are not meant to limit 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, that is, a logic flowchart of a method for quantifying uncertainty calculated by a pressurized water reactor core design software package;
the invention relates to a method for obtaining uncertainty by quantifying a pressurized water reactor nuclear design software package, which comprises the following steps:
acquiring actually measured data of a critical device and a power plant, and acquiring physical design program calculation data;
analyzing whether the data is a direct measurement parameter or not according to the acquired data;
if the parameter is a direct parameter, calculating the deviation between a calculated value and a measured value by adopting a direct parameter measurement method, and calculating to obtain the uncertainty of the direct parameter; wherein the directly measured parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient, and the like.
If the data are indirect measurement parameters, 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, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition quantity is not a directly measured parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining the uncertainty of the decomposition quantity which is not a directly measured parameter by obtaining the reaction rate, the component average power proportional relation, the correlation coefficient and the like through the disturbance data set; wherein the indirect measurement parameter is heat flow density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHAnd the like.
And carrying out uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty.
Specifically, if the measured parameter is an indirect measured parameter, decomposing the data by using an indirect measured 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 decomposing the fuel rod power to obtain heat flow density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHEtc. may be translated to a fixed value on the fuel rod power ratio (e.g.,
Figure BDA0003321180260000061
) The decomposition formula is:
Figure BDA0003321180260000062
wherein A is the reaction rate,
Figure BDA0003321180260000063
is the assembly average power, P is the fuel rod power;
s11: the indirect measurement parameter can be decomposed into three decomposition quantities by the above formula, and then the uncertainty of the indirect measurement parameter is decomposed into three parts: uncertainty of rod power distribution
Figure BDA0003321180260000064
Reactivity-component average power conversion factor uncertainty
Figure BDA0003321180260000065
Uncertainty of reaction rate σc(A)；
Respectively calculating the uncertainty of the three decomposed parts; wherein the rod power distribution is not deterministic
Figure BDA0003321180260000071
The reaction rate uncertainty sigma can be obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the actually measured value of the critical device testc(A) Detector three-dimensional reactivity distribution and electricity obtained through core program calculationAnd comparing the three-dimensional reaction rate distribution actually measured in the factory.
Specifically, if the decomposition quantity is not a direct measurement parameter, performing disturbance calculation on the state of the reactor core to obtain a disturbance data set, and obtaining the reaction rate, the component average power proportional relation, the correlation coefficient and the like through the disturbance data set;
defining the correlation coefficient of the average power of the component and the response rate of the detector as r, and converting the response rate to the average power of the component by the uncertainty of the conversion factor
Figure BDA0003321180260000072
Expressed as:
Figure BDA0003321180260000073
conventional method assumptions
Figure BDA0003321180260000074
Can be directly obtained
Figure BDA0003321180260000075
But there is some irrationality to this assumption. The present invention abandons this assumption and obtains it by the following method
Figure BDA0003321180260000076
S20: in the reactor core procedure, the one-to-one corresponding component power can be calculated by disturbing the reactor core state points (parameters such as boron concentration, temperature and power)
Figure BDA0003321180260000077
And rate of detector reaction { A1，A2，……，AnDisturbing the data set;
s21: derived from the perturbation data set
Figure BDA0003321180260000079
And
Figure BDA00033211802600000710
and the correlation coefficient r;
s22: obtained by using actually measured reaction rate of power plant
Figure BDA00033211802600000711
Solving the uncertainty of the average power conversion factor of the response rate and the component through the proportional relation and the correlation coefficient r obtained in the step S21
Figure BDA00033211802600000712
Specifically, the uncertainty synthesis is carried out to obtain indirect measurement parameters and calculate uncertainty; the method comprises the following steps:
according to uncertainty of rod power distribution
Figure BDA00033211802600000713
Reactivity-component average power conversion factor uncertainty
Figure BDA00033211802600000714
Uncertainty of reaction rate σc(A) Performing uncertainty synthesis to obtain FQI.e.:
Figure BDA00033211802600000715
this gives FQUncertainty of (D), FΔHCan be obtained by a similar method except that a two-dimensional radial reactivity distribution is used.
The invention relates to a method for obtaining uncertainty by a quantitative pressurized water reactor nuclear design software package, which is characterized in that direct measurement parameters such as critical boron concentration, control rod integral/differential value, isothermal temperature coefficient and the like are subjected to parameter statistics by adopting a direct measurement parameter method to obtain the calculated uncertainty by carrying out parameter statistics on the absolute deviation of a measured value and a calculated value; heat pipe factor F for heat flux densityQNuclear enthalpy rising heat pipe factor FΔHMeasuring parameters by equal intervals, usingAnd (4) a measurement parameter decomposition method is adopted, the original unreasonable assumption is abandoned, and the calculation uncertainty is obtained through a disturbance method. Compared with the prior art, the invention can realize direct parameter measurement and simultaneously solve the problem of FQ、FΔHAnd the method solves the problem of unreasonable assumption in the quantitative process of indirect measurement parameter uncertainty, and perfects a validation system of a pressurized water reactor nuclear design software package (component calculation program and reactor core calculation program).
Example 2
As shown in fig. 2, this embodiment is different from embodiment 1 in that this embodiment provides an apparatus for quantifying the uncertainty of the acquisition of a pressurized water reactor nuclear design software package, which supports the method for quantifying the uncertainty of the acquisition of a pressurized water reactor nuclear design software package described in embodiment 1, and the apparatus includes:
the acquisition unit is used for acquiring actually measured data of the critical device and the power plant and acquiring physical design program calculation data;
the data judging unit is used for analyzing whether the data are direct measurement parameters or not 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 carry out calculation value and measured value deviation statistics according to the fact that the data judged by the data judgment unit is the direct parameter;
the uncertainty calculation unit is indirectly connected with the measurement parameters and is used for decomposing the data by adopting an indirect measurement parameter decomposition method according to the fact that the data judged by the data judgment unit is the indirect measurement parameters to obtain the decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition quantity is not a directly measured parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining the uncertainty of the decomposition quantity which is not a directly measured parameter by obtaining the reaction rate, the component average power proportional relation, the 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.
To further illustrate the present embodiment, the directly measured parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient, etc.;
the indirect measurement parameter is heat flow density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHAnd the like.
To further illustrate the present embodiment, the uncertainty calculation unit indirectly measuring the parameter performs the following process:
converting the indirect measurement parameters into fuel rod power, and decomposing the fuel rod power to obtain heat flow density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔHEtc. may be translated to a fixed value on the fuel rod power ratio (e.g.,
Figure BDA0003321180260000081
) The decomposition formula is:
Figure BDA0003321180260000091
wherein A is the reaction rate,
Figure BDA0003321180260000092
is the assembly average power, P is the fuel rod power;
the indirect measurement parameter can be decomposed into three decomposition quantities by the above formula, and then the uncertainty of the indirect measurement parameter is decomposed into three parts: uncertainty of rod power distribution
Figure BDA0003321180260000093
Reactivity-component average power conversion factor uncertainty
Figure BDA0003321180260000094
Uncertainty of reaction rate σc(A)；
Respectively calculating the uncertainty of the three decomposed parts; wherein the rod power distribution is uncertainDegree of rotation
Figure BDA0003321180260000095
The reaction rate uncertainty sigma can be obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the actually measured value of the critical device testc(A) And comparing the three-dimensional reactivity distribution of the detector obtained by the calculation of the reactor core program with the actually measured three-dimensional reactivity distribution of the power plant.
For further explanation of this embodiment, in the uncertainty calculation unit of the indirect measurement parameter, if the decomposition amount is not a direct measurement parameter, performing disturbance calculation on the state of the reactor core to obtain a disturbance data set, and obtaining a reaction rate, a component average power proportional relation, a correlation coefficient, and the like through the disturbance data set;
defining the correlation coefficient of the average power of the component and the response rate of the detector as r, and converting the response rate to the average power of the component by the uncertainty of the conversion factor
Figure BDA0003321180260000096
Expressed as:
Figure BDA0003321180260000097
conventional method assumptions
Figure BDA0003321180260000098
Can be directly obtained
Figure BDA0003321180260000099
But there is some irrationality to this assumption. The present invention gets rid of this assumption and obtains it by the following procedure
Figure BDA00033211802600000910
In the reactor core procedure, the one-to-one corresponding component power can be calculated by disturbing the reactor core state points (parameters such as boron concentration, temperature and power)
Figure BDA00033211802600000911
And rate of detector reaction { A1,A2,……,AnDisturbing the data set;
derived from the perturbation data set
Figure BDA00033211802600000912
And
Figure BDA00033211802600000913
and the correlation coefficient r;
obtained by using actually measured reaction rate of power plant
Figure BDA00033211802600000914
Solving the uncertainty of the average power conversion factor of the response rate and the component through the proportional relation and the correlation coefficient r obtained in the step S21
Figure BDA00033211802600000915
The device provided by the invention is used for carrying out parameter statistics on the absolute deviation of a measured value and a calculated value by adopting a direct measurement parameter method 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 densityQNuclear enthalpy rising heat pipe factor FΔHAnd (3) waiting for indirect measurement parameters, adopting an indirect measurement parameter decomposition method, abandoning the original unreasonable assumption, and obtaining the uncertainty of calculation through a disturbance method. Compared with the prior art, the invention can realize direct parameter measurement and simultaneously solve the problem of FQ、FΔHAnd the method solves the problem of unreasonable assumption in the quantitative process of indirect measurement parameter uncertainty, and perfects a validation system of a pressurized water reactor nuclear design software package (component calculation program and reactor core calculation program).
As will be appreciated by one skilled in the art, 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 flow diagrams and/or block diagrams, and combinations of flows and/or blocks in the flow diagrams 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 above-mentioned embodiments are intended to illustrate the objects, technical solutions and advantages of the present invention in further detail, and it should be understood that the above-mentioned embodiments are merely exemplary embodiments of the present invention, and are not intended to limit the scope of the present invention, and any modifications, equivalent substitutions, improvements and the like made within the spirit and principle of the present invention should be included in the scope of the present invention.

Claims (10)

1. A method for quantifying uncertainty obtained by a pressurized water reactor nuclear design software package is characterized by comprising the following steps:
acquiring actually measured data of a critical device and a power plant, and acquiring physical design program calculation data;
analyzing whether the data is a direct measurement parameter or not according to the acquired data;
if the parameter is a direct parameter, calculating the deviation between a calculated value and a measured value by adopting a direct parameter measurement method, and calculating to obtain the uncertainty of the direct parameter;
if the data are indirect measurement parameters, 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, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition quantity is not a directly measured parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining the reaction rate, the component average power proportional relation and the correlation coefficient through the disturbance data set to obtain the uncertainty of the decomposition quantity which is not a directly measured parameter;
and carrying out uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty.
2. The method of claim 1, wherein the directly measured parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient.
3. The method for quantifying the uncertainty obtained by a PWR nuclear design software package according to claim 1, wherein the indirect measurement parameter is heat flux density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔH。
4. The method for obtaining the uncertainty of the quantitative pressurized water reactor nuclear design software package according to claim 1, wherein if the measured parameter is an indirect measured parameter, the data is decomposed by an indirect measured parameter decomposition method to obtain a decomposition amount; the method specifically comprises the following steps:
s10: and decomposing the indirect measurement parameters after converting the indirect measurement parameters into the power of the fuel rod, wherein the decomposition formula is as follows:
Figure FDA0003321180250000011
wherein A is the reaction rate,
Figure FDA0003321180250000012
is the assembly average power, P is the fuel rod power;
s11: the indirect measurement parameters are decomposed into three decomposition quantities by the above formula, and then the uncertainty of the indirect measurement parameters is decomposed into three parts: uncertainty of rod power distribution
Figure FDA0003321180250000013
Reactivity-component average power conversion factor uncertainty
Figure FDA0003321180250000014
Uncertainty of reaction rate σc(A)。
5. The method for obtaining the uncertainty of the quantified PWR nuclear design software package according to claim 4, wherein if the decomposition quantity is not a direct measurement parameter, performing disturbance calculation on the state of the reactor core 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 procedure, the one-to-one corresponding component power is calculated by disturbing the reactor core state points
Figure FDA0003321180250000021
And rate of detector reaction { A1,A2,……,AnDisturbing the data set;
s21: derived from the perturbation data set
Figure FDA0003321180250000022
And
Figure FDA0003321180250000023
and the correlation coefficient r;
s22: obtained by using actually measured reaction rate of power plant
Figure FDA0003321180250000024
Solving the uncertainty of the average power conversion factor of the response rate and the component through the proportional relation and the correlation coefficient r obtained in the step S21
Figure FDA0003321180250000025
6. The method for obtaining the uncertainty of the quantitative pressurized water reactor nuclear design software package according to claim 5, wherein the uncertainty synthesis is carried out to obtain indirect measurement parameters and calculate the uncertainty; the method comprises the following steps:
according to uncertainty of rod power distribution
Figure FDA0003321180250000026
Reactivity-component average power conversion factor uncertainty
Figure FDA0003321180250000027
Uncertainty of reaction rate σc(A) Carrying out synthesis with uncertainty to obtain FQI.e.:
Figure FDA0003321180250000028
7. an apparatus for quantifying the uncertainty obtained by a PWR nuclear design package, the apparatus supporting a method of quantifying the uncertainty obtained by a PWR nuclear design package according to any of claims 1 to 6, the apparatus comprising:
the acquisition unit is used for acquiring actually measured data of the critical device and the power plant and acquiring physical design program calculation data;
the data judging unit is used for analyzing whether the data are direct measurement parameters or not 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 carry out calculation value and measured value deviation statistics according to the fact that the data judged by the data judgment unit is the direct parameter;
the uncertainty calculation unit is indirectly connected with the measurement parameters and is used for decomposing the data by adopting an indirect measurement parameter decomposition method according to the fact that the data judged by the data judgment unit is the indirect measurement parameters to obtain the decomposition amount; judging whether the decomposition amount is a direct measurement parameter or not, and if the decomposition amount is the direct measurement parameter, calculating by adopting a direct measurement parameter method; if the decomposition quantity is not a directly measured parameter, carrying out disturbance calculation on the reactor core state to obtain a disturbance data set, and obtaining the reaction rate, the component average power proportional relation and the correlation coefficient through the disturbance data set to obtain the uncertainty of the decomposition quantity which is not a directly measured parameter;
and the uncertainty synthesis unit is used for performing uncertainty synthesis to obtain indirect measurement parameters and calculating uncertainty.
8. The apparatus of claim 7, wherein the direct measurement parameters include critical boron concentration, control rod integral/differential value, isothermal temperature coefficient;
the indirect measurement parameter is heat flow density heat pipe factor FQNuclear enthalpy rising heat pipe factor FΔH。
9. The apparatus of claim 7, wherein the uncertainty calculation unit indirectly connected to the measured parameter is configured to:
and decomposing the indirect measurement parameters after converting the indirect measurement parameters into the power of the fuel rod, wherein the decomposition formula is as follows:
Figure FDA0003321180250000031
wherein A is the reaction rate,
Figure FDA0003321180250000032
is the assembly average power, P is the fuel rod power;
the indirect measurement parameters are decomposed into three decomposition quantities by the above formula, and then the uncertainty of the indirect measurement parameters is decomposed into three parts: uncertainty of rod power distribution
Figure FDA0003321180250000033
Reactivity-component average power conversion factor uncertainty
Figure FDA0003321180250000034
Uncertainty of reaction rate σc(A)；
Respectively calculating the uncertainty of the three decomposed parts; wherein the rod power distribution is not deterministic
Figure FDA0003321180250000035
The reaction rate uncertainty sigma is obtained by comparing the rod power after the reactor core program power distribution is reconstructed with the actually measured value of the critical device testc(A) Detector three-dimensional inverse obtained through core program calculationAnd comparing the response rate distribution with the actually measured three-dimensional reaction rate distribution of the power plant.
10. The apparatus of claim 9, wherein in the uncertainty calculation unit indirectly connected to the measured parameter, if the decomposition is not a directly measured parameter, the disturbance calculation is performed on the core state to obtain a disturbance data set, and the reaction rate, the component average power proportional relationship, and the correlation coefficient are obtained from the disturbance data set; the specific implementation process is as follows:
in the reactor core procedure, the one-to-one corresponding component power is calculated by disturbing the reactor core state points
Figure FDA0003321180250000036
And rate of detector reaction { A1,A2,……,AnDisturbing the data set;
derived from the perturbation data set
Figure FDA0003321180250000037
And
Figure FDA0003321180250000038
and the correlation coefficient r;
obtained by using actually measured reaction rate of power plant
Figure FDA0003321180250000039
Solving the uncertainty of the average power conversion factor of the response rate and the component through the proportional relation and the correlation coefficient r obtained in the step S21
Figure FDA00033211802500000310
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