CN117637217A

CN117637217A - Neutron segment model adjustment method, device, computer equipment and storage medium

Info

Publication number: CN117637217A
Application number: CN202311439286.9A
Authority: CN
Inventors: 厉井钢; 李文淮; 甘平平; 王军令; 卢皓亮; 蔡利; 张香菊; 陈俊; 王超; 彭靖含
Original assignee: China Nuclear Power Technology Research Institute Co Ltd
Current assignee: China Nuclear Power Technology Research Institute Co Ltd
Priority date: 2023-11-01
Filing date: 2023-11-01
Publication date: 2024-03-01

Abstract

The application relates to a neutron science segment model adjustment method, a neutron science segment model adjustment device, computer equipment and a storage medium. The method comprises the following steps: obtaining a fast group neutron flux distribution measured value and a thermal group neutron flux distribution measured value of a target fuel segment; acquiring an initial equivalent section of a target fuel segment, wherein the initial equivalent section comprises an initial fast-group macroscopic absorption section, an initial thermal-group macroscopic absorption section and an initial downward macroscopic scattering section; determining each equivalent cross-section correction amount based on the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value, the first preset relationship and the second preset relationship; adjusting each initial equivalent section according to each equivalent section correction amount to obtain each target equivalent section; and adjusting an initial neutron nugget model of the target fuel nugget based on each target equivalent section to obtain the target neutron nugget model. By adopting the method, the economy of the nuclear power station can be protected, and the accuracy of the neutron science segment model can be improved.

Description

Neutron segment model adjustment method, device, computer equipment and storage medium

Technical Field

The present disclosure relates to the field of nuclear reactor operation, and in particular, to a method and apparatus for adjusting a neutron segment model, a computer device, and a storage medium.

Background

Core on-line monitoring is one of important means for guaranteeing safe operation of a nuclear power station, and particularly, core three-dimensional power distribution on-line monitoring is achieved. The technology can reflect the nuclear reaction condition in the nuclear reactor core in real time, and timely master the states of fuel and thermal fluid, thereby ensuring the safe operation of the nuclear power station. And when the deviation between the theoretical power distribution and the measured power distribution exceeds a certain threshold, further measures are taken.

In the conventional method, when the deviation between the theoretical power distribution and the measured power distribution is large, in order to reduce the deviation, one method is to reduce the power level of the core of the nuclear power plant so that the measured power approaches the theoretical power, which may affect the economy of the nuclear power plant; the other method is a model calibration method based on data assimilation, and model parameter adjustment is carried out by fusing a large amount of experimental data.

Disclosure of Invention

In view of the foregoing, it is desirable to provide a neutron segment model adjustment method, device, computer device, and storage medium, which can protect the economy of a nuclear power plant and improve the accuracy of the neutron segment model.

In a first aspect, the present application provides a neutron nugget model adjustment method applied to a reactor core, the reactor core including a fuel nugget, the fuel nugget including fast group neutrons and thermal group neutrons, the method comprising:

obtaining a fast group neutron flux distribution measured value and a thermal group neutron flux distribution measured value of a target fuel segment;

acquiring an initial equivalent section of the target fuel segment, wherein the initial equivalent section comprises an initial fast group macroscopic absorption section, an initial thermal group macroscopic absorption section and an initial downward macroscopic scattering section;

determining each equivalent cross-section correction amount based on the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value, a first preset relationship and a second preset relationship;

adjusting each initial equivalent section according to each equivalent section correction amount to obtain each target equivalent section;

and adjusting an initial neutron segment model of the target fuel segment based on each target equivalent section to obtain a target neutron segment model.

In one embodiment, the obtaining fast group neutron flux distribution measurements and thermal group neutron flux distribution measurements of the target fuel segments includes:

Obtaining a fast group neutron flux distribution theoretical value and a thermal group neutron flux distribution theoretical value of the target fuel segment according to the initial neutron segment model;

acquiring a power distribution measurement value of the target fuel segment;

and obtaining a fast group neutron flux distribution measurement value and a thermal group neutron flux distribution measurement value of the target fuel section according to a first preset formula, a second preset formula, the fast group neutron flux distribution theoretical value, the thermal group neutron flux distribution theoretical value and the power distribution measurement value of the target fuel section.

In one embodiment, the obtaining the initial equivalent cross section of each segment includes:

obtaining an equivalent cross section information base, wherein the equivalent cross section information base comprises a first mapping relation, a second mapping relation and a third mapping relation, the first mapping relation comprises a mapping relation between a fast group macroscopic absorption cross section of the target fuel section and a first target state parameter corresponding to the fast group macroscopic absorption cross section in all state parameters of the target fuel section, the second mapping relation comprises a mapping relation between a thermal group macroscopic absorption cross section of the target fuel section and a second target state parameter corresponding to the thermal group macroscopic absorption cross section in all state parameters of the target fuel section, and the third mapping relation comprises a mapping relation between a downward macroscopic scattering cross section of the target fuel section and a third target state parameter corresponding to the downward macroscopic scattering cross section in all state parameters of the target fuel section;

Acquiring a first expansion coefficient of the initial fast-group macroscopic absorption section according to the first target state parameter, acquiring a second expansion coefficient of the initial thermal-group macroscopic absorption section according to the second target state parameter, and acquiring a third expansion coefficient of the initial downward macroscopic scattering section according to the third target state parameter;

acquiring the initial fast group macroscopic absorption section according to the first target state parameter and the first expansion coefficient, acquiring the initial thermal group macroscopic absorption section according to the second target state parameter and the second expansion coefficient, and acquiring the initial downward macroscopic scattering section according to the third target state parameter and the third expansion coefficient.

In one embodiment, the first preset relationship includes a first correction relationship, a second correction relationship and a third correction relationship, where the first correction relationship is an association relationship between a fast-group macroscopic absorption cross-section correction amount and a fitting coefficient corresponding to the fast-group macroscopic absorption cross-section correction amount and the first target state parameter, the second correction relationship is an association relationship between a thermal-group macroscopic absorption cross-section correction amount and a fitting coefficient corresponding to the thermal-group macroscopic absorption cross-section correction amount and the second target state parameter, and the third correction relationship is an association relationship between a downward macroscopic scattering cross-section correction amount and a fitting coefficient corresponding to the thermal-group macroscopic absorption cross-section correction amount and the third target state parameter.

In one embodiment, the second preset relationship includes: the first formula to be balanced and the second formula to be balanced, wherein the first formula to be balanced is a correlation between a preset fast neutron outflow amount, a preset fast neutron production amount, a fast group neutron flux distribution measurement value, an initial fast group macroscopic absorption section, a fast group macroscopic absorption section correction amount, an initial downward macroscopic scattering section, a downward macroscopic scattering section correction amount and a first flux error, and the second formula to be balanced is a correlation between a preset thermal neutron outflow amount, a thermal group neutron flux distribution measurement value, an initial thermal group macroscopic absorption section, a thermal group macroscopic absorption section correction amount, an initial downward macroscopic scattering section, a downward macroscopic scattering section correction amount and a second flux error.

In one embodiment, the determining each equivalent cross-sectional correction amount based on the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value, the first preset relationship, and the second preset relationship includes:

substituting the first correction relation and the third correction relation into the first formula to be balanced to obtain a first target relation, and substituting the second correction relation and the third correction relation into the second formula to be balanced to obtain a second target relation;

Acquiring each first target fitting coefficient according to the first target relation, acquiring each second target fitting coefficient according to the second target relation, and acquiring each third target fitting coefficient according to a third target relation, wherein the third target relation is one of the first target relation and the second target relation;

and acquiring the fast group macroscopic absorption cross section correction amount according to the first target fitting coefficient and the first correction relation, acquiring the thermal group macroscopic absorption cross section correction amount according to the second target fitting coefficient and the second correction relation, and acquiring the downward macroscopic scattering cross section correction amount according to the third target fitting coefficient and the third correction relation.

In one embodiment, the first preset formula is:

wherein,a theoretical value of neutron flux distribution in said fast group for said target fuel segment n,>a theoretical value of neutron flux distribution in said thermal group for said target fuel segment n,>for the fast group neutron flux distribution measurement value of the target fuel segment n, +.>Neutron flux distribution measurements for the thermal group of the target fuel segment n;

The second preset formula is:

wherein,generating a cross section for a preset fast group macroscopic energy of said target fuel segment n,/for a target fuel segment n>For the preset purposeHeat group macroscopic energy generation section of target fuel segment n +.>Is the power distribution measurement of the target fuel segment n.

In a second aspect, the application also provides a neutron segment model adjustment device. The device comprises:

the first acquisition module is used for acquiring a fast group neutron flux distribution measured value and a thermal group neutron flux distribution measured value of the target fuel segment;

the second acquisition module is used for acquiring an initial equivalent section of the target fuel section, wherein the initial equivalent section comprises an initial fast group macroscopic absorption section, an initial thermal group macroscopic absorption section and an initial downward macroscopic scattering section;

the first determining module is used for determining each equivalent cross-section correction amount based on the fast group neutron flux distribution measured value, the thermal group neutron flux distribution measured value, a first preset relation and a second preset relation;

the third acquisition module is used for adjusting each initial equivalent section according to each equivalent section correction amount to acquire each target equivalent section;

And the fourth acquisition module is used for adjusting the initial neutron segment model of the target segment based on each target equivalent section so as to acquire the target neutron segment model.

In a third aspect, the present application also provides a computer device. The computer device comprises a memory storing a computer program and a processor which when executing the computer program performs the steps of:

and adjusting an initial neutron segment model of the target segment based on each target equivalent section to obtain a target neutron segment model.

In a fourth aspect, the present application also provides a computer-readable storage medium. The computer readable storage medium having stored thereon a computer program which when executed by a processor performs the steps of:

According to the neutron segment model adjustment method, the device, the computer equipment and the storage medium, the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value and the initial equivalent cross section of the target fuel segment are obtained, the equivalent cross section correction amounts are determined based on the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value, the first preset relation and the second preset relation, the initial fast group macroscopic absorption cross section, the initial thermal group macroscopic absorption cross section and the initial downward macroscopic scattering cross section of the target fuel segment are respectively subjected to cross section correction through the equivalent cross section correction amounts, so that each target equivalent cross section with higher accuracy is obtained, and finally the initial neutron segment model of the target segment is adjusted based on each target equivalent cross section, so that the target neutron segment model is obtained. The method does not need to reduce the power level of the reactor core of the nuclear power station, avoids affecting the economy of the nuclear power station, adjusts the initial neutron segment model based on the measured value of the target fuel segment, does not need a large amount of experimental data, and avoids the accuracy reduction of model parameters caused by the inaccuracy of the data when the data are too much.

Drawings

FIG. 1 is a flow chart of a method for adjusting a neutron segment model in one embodiment;

FIG. 2 is a flow chart of step S101 in one embodiment;

FIG. 3 is a flow chart of step S102 in one embodiment;

FIG. 4 is a flow chart of step S103 in one embodiment;

FIG. 5 is a block diagram of a neutron segment model tuning device in one embodiment;

fig. 6 is an internal structural diagram of a computer device in one embodiment.

Detailed Description

In order to make the objects, technical solutions and advantages of the present application more apparent, the present application will be further described in detail with reference to the accompanying drawings and examples. It should be understood that the specific embodiments described herein are for purposes of illustration only and are not intended to limit the present application.

In one embodiment, as shown in fig. 1, a neutron nugget model adjustment method is provided, and is applied to a reactor core, wherein the reactor core comprises a fuel nugget, and the fuel nugget comprises fast group neutrons and thermal group neutrons, and the method comprises the following steps:

s101: and obtaining a fast group neutron flux distribution measurement value and a thermal group neutron flux distribution measurement value of the target fuel segment.

The nuclear fission reaction can occur in the nuclear reactor core, one nuclear can be fissionally generated to generate a plurality of fast neutrons, the fast neutrons collide with medium nuclear again, the energy of the fast neutrons is reduced to be slowed down into slow neutrons (namely thermal neutrons), the fast neutrons or the thermal neutrons can be absorbed by the nuclear again, further the next reaction is initiated, the fuel in the core is equivalent to a plurality of fuel segments by dividing the fuel in the core, the power distribution of each fuel segment is analyzed, and the power distribution of the whole core can be analyzed more accurately. Therefore, the fast group neutron flux distribution measured value and the thermal group neutron flux distribution measured value of the target fuel segment can be obtained, and the fast group neutron flux distribution measured value and the thermal group neutron flux distribution measured value of the target fuel segment are used for analyzing the operation condition of the reactor.

S102: obtaining an initial equivalent cross section of the target fuel segment, wherein the initial equivalent cross section comprises an initial fast-group macroscopic absorption cross section, an initial thermal-group macroscopic absorption cross section and an initial downward macroscopic scattering cross section.

As nuclear reactions occur in the nuclear reactor at any time, all nuclear reactions are classified, nine reaction types can be obtained, each reaction type can be equivalently a macroscopic cross section, the probability of a certain specific nuclear reaction between an incident particle and a target is represented, and nine equivalent cross sections are respectively: fast group diffusion coefficient, thermal group diffusion coefficient, downward macroscopic scattering section, fast group macroscopic absorption section, thermal group macroscopic absorption section, fast group macroscopic neutron generation section, thermal group macroscopic neutron generation section, fast group macroscopic energy generation section, and thermal group macroscopic energy generation section. Since the downward macroscopic scattering section, the fast-group macroscopic absorption section and the thermal-group macroscopic absorption section are more modified in actual operation, only the acquisition and modification of the initial sections of the downward macroscopic scattering section, the fast-group macroscopic absorption section and the thermal-group macroscopic absorption section are described in the application, and in actual application, technicians can acquire other equivalent sections and modify other equivalent sections according to similar methods in the application as required to adjust the neutron segment model.

S103: and determining each equivalent cross-section correction amount based on the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value, the first preset relation and the second preset relation.

And determining the correction quantity of the initial fast group macroscopic absorption section, the correction quantity of the initial thermal group macroscopic absorption section and the correction quantity of the initial downward macroscopic scattering section respectively according to the acquired fast group neutron flux distribution measured value, the thermal group neutron flux distribution measured value, the first preset relation and the second preset relation.

S104: and adjusting each initial equivalent section according to each equivalent section correction amount to obtain each target equivalent section.

After the correction of the initial fast-group macroscopic absorption section, the correction of the initial thermal-group macroscopic absorption section and the correction of the initial downward macroscopic scattering section are obtained, the correction of the initial fast-group macroscopic absorption section can be used for correcting the initial fast-group macroscopic absorption section to obtain a target fast-group macroscopic absorption section, the correction of the initial thermal-group macroscopic absorption section is used for correcting the initial thermal-group macroscopic absorption section to obtain a target thermal-group macroscopic absorption section, and the correction of the initial downward macroscopic scattering section is used for correcting the initial downward macroscopic scattering section to obtain the target downward macroscopic scattering section.

S105: and adjusting an initial neutron nugget model of the target fuel nugget based on each target equivalent section to obtain the target neutron nugget model.

The initial neutron segment model of the target fuel segment can be adjusted according to the target fast-group macroscopic absorption section, the target thermal-group macroscopic absorption section and the target downward macroscopic scattering section to obtain the target neutron segment model of the target fuel segment, so that the gap between the theoretical value and the measured value of the fast-group neutron flux and the thermal-group neutron flux of the target fuel segment is reduced, and the gap between the theoretical value and the measured value of the power distribution of the target fuel segment is reduced. By traversing all fuel segments in the reactor core or setting all fuel segments in the reactor core as target fuel segments, target neutron segment models of all fuel segments can be obtained, and based on the obtained target neutron segment models, the difference between the theoretical value and the measured value of fast group neutron flux and thermal group neutron flux in the reactor core can be reduced, and the difference between the theoretical value and the measured value of power distribution of the reactor core can be reduced.

According to the neutron segment model adjustment method, the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value and the initial equivalent cross section of the target fuel segment are obtained, the equivalent cross section correction amounts are determined based on the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value, the first preset relation and the second preset relation, the initial fast group macroscopic absorption cross section, the initial thermal group macroscopic absorption cross section and the initial downward macroscopic scattering cross section of the target fuel segment are subjected to cross section correction respectively through the equivalent cross section correction amounts, so that each target equivalent cross section with higher accuracy is obtained, and finally the initial neutron segment model of the target segment is adjusted based on each target equivalent cross section, so that the target neutron segment model is obtained. The method does not need to reduce the power level of the reactor core of the nuclear power station, avoids affecting the economy of the nuclear power station, adjusts the initial neutron segment model based on the measured value of the target fuel segment, does not need a large amount of experimental data, and avoids the accuracy reduction of model parameters caused by the inaccuracy of the data when the data are too much.

In one embodiment, as shown in fig. 2, step S101 includes:

S201: and obtaining a fast group neutron flux distribution theoretical value and a thermal group neutron flux distribution theoretical value of the target fuel segment according to the initial neutron segment model.

Theoretical calculation is carried out through an initial neutron segment model, so that the theoretical value of three-dimensional power distribution of the target fuel segment, the theoretical value of fast group neutron flux distribution and the theoretical value of thermal group neutron flux distribution can be obtained.

S202: a power distribution measurement of the target fuel segment is obtained.

The nuclear reactor is also provided with a reactor core three-dimensional on-line monitoring system which can acquire the power distribution measured value of the target fuel segment.

S203: and obtaining a fast group neutron flux distribution measured value and a thermal group neutron flux distribution measured value of the target fuel section according to the first preset formula, the second preset formula, the fast group neutron flux distribution theoretical value, the thermal group neutron flux distribution theoretical value and the power distribution measured value of the target fuel section.

Assuming that the initial neutron segment model can effectively predict the local neutron spectrum, a first preset formula is provided:

wherein,theoretical value of neutron flux distribution in fast group for target fuel segment n +.>Theoretical value of neutron flux distribution in thermal mass for target fuel segment n +.>Measurement of the neutron flux distribution in the fast group for the target fuel segment n, >A measure of neutron flux distribution in the thermal mass for the target fuel segment n.

And combining the definition of the power measurement value of the target fuel segment, namely a second preset formula:

wherein,generating a cross section for the fast group macroscopic energy of a preset target fuel segment n, +.>Macroscopic energy of heat group for preset target fuel segment nProducing a cross section->Is a power distribution measurement of the target fuel segment n.

And for the target fuel segment n, obtaining a fast group neutron flux distribution measurement value and a thermal group neutron flux distribution measurement value of the target fuel segment based on the first preset formula, the second preset formula, the fast group neutron flux distribution theoretical value, the thermal group neutron flux distribution theoretical value and the power distribution measurement value of the target fuel segment.

In one embodiment, as shown in fig. 3, step S102 includes:

s301: obtaining an equivalent section information base, wherein the equivalent section information base comprises a first mapping relation, a second mapping relation and a third mapping relation, the first mapping relation comprises a mapping relation between a fast group macroscopic absorption section of a target fuel section and a first target state parameter corresponding to the fast group macroscopic absorption section in all state parameters of the target fuel section, the second mapping relation comprises a mapping relation between a thermal group macroscopic absorption section of the target fuel section and a second target state parameter corresponding to the thermal group macroscopic absorption section in all state parameters of the target fuel section, and the third mapping relation comprises a mapping relation between a downward macroscopic scattering section of the target fuel section and a third target state parameter corresponding to the downward macroscopic scattering section in all state parameters of the target fuel section.

It will be appreciated that the equivalent cross-section information library may include mapping relationships between all nine equivalent cross-sections and state parameters of the fuel segment, including burnup, boron concentration, moderator density, effective fuel temperature, xenon concentration, pu/U ratio, etc., and that the number of state parameters of the fuel segment affecting the equivalent cross-sections may be defined in the present invention, and that, illustratively, each equivalent cross-section may be defined to have a mapping relationship with 3 state parameters, assuming that the fast-group macroscopic absorption cross-section of the target fuel segment is related to the burnup, boron concentration, xenon concentration of the target fuel segment, and the hot-group macroscopic absorption cross-section of the target fuel segment is related to the boron concentration, moderator density, effective fuel temperature of the target fuel segmentThe downward macroscopic scattering cross section of the fuel segment is related to the burnup, moderator density, and fuel effective temperature of the target fuel segment, i.e., the first target state parameter isWherein->For the burnup of the target fuel segment n, +.>Boron concentration for target fuel segment n, +.>For the xenon concentration of the target fuel segment n, the second target state parameter isWherein->Boron concentration for target fuel segment n, +.>Moderator density for target fuel segment n,/- >The fuel effective temperature of the target fuel segment n is the third target state parameter ofWherein->For the burnup of the target fuel segment n, +.>Moderator density for target fuel segment n,/->Is the fuel effective temperature of the target fuel segment n.

S302: acquiring a first expansion coefficient of an initial fast group macroscopic absorption section according to a first target state parameter, acquiring a second expansion coefficient of an initial thermal group macroscopic absorption section according to a second target state parameter, and acquiring a third expansion coefficient of an initial downward macroscopic scattering section according to a third target state parameter.

Before each initial equivalent section is obtained according to the target state parameter fitting, a preset first expansion coefficient, a preset second expansion coefficient and a preset third expansion coefficient which are needed by calculating each initial equivalent section are also needed to be obtained from an equivalent section information base.

S303: acquiring an initial fast group macroscopic absorption section according to a first target state parameter and a first expansion coefficient, acquiring an initial thermal group macroscopic absorption section according to a second target state parameter and a second expansion coefficient, and acquiring an initial downward macroscopic scattering section according to a third target state parameter and a third expansion coefficient.

After the first target state parameter and the first expansion coefficient are obtained, an initial fast-group macroscopic absorption section can be obtained through fitting, and similarly, an initial thermal group macroscopic absorption section can be obtained based on the second target state parameter and the second expansion coefficient, and an initial downward macroscopic scattering section can be obtained based on the third target state parameter and the third expansion coefficient:

Wherein,for the initial fast-group macroscopic absorption cross section, +.>For the burnup of the target fuel segment n, +.>Boron concentration for target fuel segment n, +.>Xenon concentration for target fuel segment n, +.>For each first expansion coefficient, < >>For the initial heat group macroscopic absorption cross section, +.>Boron concentration for target fuel segment n, +.>Moderator density for target fuel segment n,/->Fuel effective temperature for target fuel segment n, +.>For each second expansion coefficient, < >>For an initial downward macroscopic scattering cross section, +.>For each third expansion coefficient, < >>For the burnup of the target fuel segment n, +.>Moderator density for target fuel segment n,/->Is the fuel effective temperature of the target fuel segment n.

In one embodiment, the first preset relation includes a first correction relation, a second correction relation and a third correction relation, where the first correction relation is an association relation between a fast group macroscopic absorption cross section correction amount, a fitting coefficient corresponding to the fast group macroscopic absorption cross section correction amount, and a first target state parameter, the second correction relation is an association relation between a heat group macroscopic absorption cross section correction amount, a fitting coefficient corresponding to the heat group macroscopic absorption cross section correction amount, and a second target state parameter, and the third correction relation is an association relation between a downward macroscopic scattering cross section correction amount, a fitting coefficient corresponding to the heat group macroscopic absorption cross section correction amount, and a third target state parameter.

The first correction relation, the second correction relation and the third correction relation are respectively:

wherein, the formula (6) is a first correction relation, the formula (7) is a second correction relation, the formula (8) is a third correction relation, each equivalent section can be defined to have a mapping relation with 3 state parameters, namely n=3,for rapid group macroscopic absorption cross-section correction, +.>And->For each fitting coefficient corresponding to the correction quantity of the macro absorption cross section of the fast group,for the first target state parameter,/for the first target state parameter>Macroscopic absorption cross-sectional correction for heat group, +.>And->Fitting coefficients corresponding to the correction amounts of macroscopic absorption cross sections of the thermal group, < >>For the second target state parameter,/->For the downward macroscopic scatter cross-section correction, +.>And->For each fitting coefficient corresponding to the downward macroscopic scatter cross-section correction,is a third target state parameter.

In practical application, the equivalent cross-section correction amount can be defined and obtained based on other state parameters of the non-target state parameters of the target fuel segment n as required to improve the accuracy of the equivalent cross-section correction amount, and the quick-group macroscopic absorption cross-section correction amount can be defined according to five state parameters of fuel consumption, boron concentration, moderator density, fuel effective temperature and xenon concentration, for example, due to the following reasons For the burnup of the target fuel segment n, +.>For the boron concentration of the target fuel segment n,for the xenon concentration of the target fuel segment n, it is possible to define +.>For slowing agent density,/->The effective temperature of the fuel is further defined as the correction amount of the macro absorption cross section of the fast group. And more complex correction relations such as spline function fitting formulas and the like can be designed according to the needs to obtain the correction quantity of each equivalent section.

In one embodiment, the second preset relationship comprises: the first formula to be balanced and the second formula to be balanced, wherein the first formula to be balanced is a correlation relationship between a preset fast neutron outflow amount, a preset fast neutron production amount, a fast group neutron flux distribution measurement value, an initial fast group macroscopic absorption section, a fast group macroscopic absorption section correction amount, an initial downward macroscopic scattering section, a downward macroscopic scattering section correction amount and a first flux error, and the second formula to be balanced is a correlation relationship between a preset thermal neutron outflow amount, a thermal group neutron flux distribution measurement value, an initial thermal group macroscopic absorption section, a thermal group macroscopic absorption section correction amount, an initial downward macroscopic scattering section, a downward macroscopic scattering section correction amount and a second flux error.

In the uncorrected initial neutron block model, neutron balance is realized in the target fuel block, and after a fast group neutron flux distribution theoretical value and a thermal group neutron flux distribution theoretical value in a neutron balance equation are replaced by a fast group neutron flux distribution measured value and a thermal group neutron flux distribution measured value, the original balanced neutron balance equation is not balanced any more, so that a first formula to be balanced and a second formula to be balanced can be obtained:

Wherein, the formula (9) is a first formula to be balanced, the formula (10) is a second formula to be balanced,for the first flux error, +.>For a predetermined fast neutron outflow, i.e. a predetermined fast group neutron flux which flows out of the six surfaces of the segment,for a preset fast neutron production, i.e. a preset fast neutron flux due to nuclear fission,for the second flux error, +.>The method is characterized in that the preset thermal neutron outflow quantity is preset, namely the thermal group neutron flux which flows out of six surfaces of the segment in a net mode.

In one embodiment, as shown in fig. 4, step S103 includes:

s401: substituting the first correction relation and the third correction relation into a first formula to be balanced to obtain a first target relation, and substituting the second correction relation and the third correction relation into a second formula to be balanced to obtain a second target relation.

Substituting the first correction relation and the third correction relation into a first formula to be balanced and substituting the second correction relation and the third correction relation into a second formula to be balanced to obtain a first target relation and a second target relation, wherein the formula (11) is the first target relation, and the formula (12) is the second target relation:

S402: and obtaining each first target fitting coefficient according to the first target relation, obtaining each second target fitting coefficient according to the second target relation, and obtaining each third target fitting coefficient according to the third target relation, wherein the third target relation is one of the first target relation and the second target relation.

In order to minimize the deviation of the measured neutron flux distribution from the theoretical neutron flux distribution, it is desirable to haveAnd->Since the polynomial equation (11) and equation (12) are continuously derivable, fitting coefficients +_for each of the first target equations>And->Respectively deriving, such as: />To obtain the fitting coefficients of the first object, and similarly, fitting coefficients of the second object relation>And->Respectively deriving, such as: />To obtain the second target fitting coefficients, fitting coefficients in the first target relation or the second target relation>And->Respectively deriving, such as: />To obtain respective third target fitting coefficients.

S403: and acquiring a fast macro absorption cross section correction amount according to each first target fitting coefficient and the first correction relation, acquiring a thermal group macro absorption cross section correction amount according to each second target fitting coefficient and the second correction relation, and acquiring a downward macro scattering cross section correction amount according to each third target fitting coefficient and the third correction relation.

After the first target fitting coefficients, the second target fitting coefficients and the third target fitting coefficients are obtained, the first target fitting coefficients are substituted into a first correction relation, namely a formula (6), so that a fast-group macroscopic absorption cross section correction amount is obtained, the second target fitting coefficients are substituted into a second correction relation, namely a formula (7), so that a thermal-group macroscopic absorption cross section correction amount is obtained, and the third target fitting coefficients are substituted into a third correction relation, namely a formula (8), so that a downward macroscopic scattering cross section correction amount is obtained. In application, different types of fuel can be combined to obtain each fitting coefficient and each equivalent cross-section correction amount, so that the neutron segment model can be adjusted more accurately.

After the fast-group macroscopic absorption section correction amount, the thermal-group macroscopic absorption section correction amount and the downward macroscopic scattering section correction amount are obtained, the fast-group macroscopic absorption section correction amount and the initial fast-group macroscopic absorption section are added to obtain a target fast-group macroscopic absorption section, the thermal-group macroscopic absorption section correction amount and the initial thermal-group macroscopic absorption section are added to obtain a target thermal-group macroscopic absorption section, and the downward macroscopic scattering section correction amount and the initial downward macroscopic scattering section are added to obtain a target downward macroscopic scattering section.

And replacing the initial fast group macroscopic absorption section, the initial thermal group macroscopic absorption section and the initial downward macroscopic scattering section in the initial neutron segment model of the target fuel segment by the target fast group macroscopic absorption section, the target thermal group macroscopic absorption section and the target downward macroscopic scattering section to obtain the target neutron segment model.

In application, the neutron science segment model can be adjusted in real time by acquiring the theoretical value of fast group neutron flux distribution, the theoretical value of thermal group neutron flux distribution, the power distribution measured value, the fast group neutron flux distribution measured value and the thermal group neutron flux distribution measured value of each fuel segment in the reactor core in real time and acquiring and calculating the equivalent cross-section correction of each fuel segment in real time.

Before fitting coefficients of the equivalent cross-section correction amounts are actually applied, the correctness of the fitting coefficients can be verified by constructing a theoretical model of a reference table-disturbance dynamic: first, establishing a neutron reactor core segment model of a referenceForm the set of relevant equivalent cross sections->Theoretical neutron flux distribution->And->And power distribution->Re-establishing a perturbed neutron block model in the core>Regarding the disturbance state model as an actual reactor state model to form a related equivalent section set +. >Form a disturbed flux distribution->And->And power distribution->. Then, each fitting coefficient and each equivalent cross-section correction amount are obtained according to the method of obtaining the fitting coefficient and the equivalent cross-section correction amount of each equivalent cross-section correction amount in the above embodiment>And according to the corrected equivalent section +.>Establishing a correction model->And calculating the corrected neutron flux distribution based on the correction model>And->And modified power profile +.>. Then, the correction model can be compared from a plurality of angles>And disturbance state model->If the difference between the corrected equivalent section and the equivalent section of the disturbance state model is compared, the following steps are performed: />Comparing the difference between the neutron flux distribution of the correction model and the neutron flux distribution of the disturbance state model: />Comparing the difference between the power distribution of the correction model and the neutron flux distribution of the disturbance state model: />Only when all threshold comparisons are respectively smaller than the preset thresholdAnd->And when the fitting coefficient is correct, the fitting coefficient correction method can be verified, and the neutron segment model adjustment method is applied to actual engineering.

It should be understood that, although the steps in the flowcharts related to the above embodiments are sequentially shown as indicated by arrows, these steps are not necessarily sequentially performed in the order indicated by the arrows. The steps are not strictly limited to the order of execution unless explicitly recited herein, and the steps may be executed in other orders. Moreover, at least some of the steps in the flowcharts described in the above embodiments may include a plurality of steps or a plurality of stages, which are not necessarily performed at the same time, but may be performed at different times, and the order of the steps or stages is not necessarily performed sequentially, but may be performed alternately or alternately with at least some of the other steps or stages.

Based on the same inventive concept, the embodiment of the application also provides a neutron segment model adjusting device for realizing the neutron segment model adjusting method. The implementation of the solution provided by the device is similar to the implementation described in the above method, so the specific limitation in the embodiments of the device for adjusting a neutron segment model provided below may be referred to the limitation of the method for adjusting a neutron segment model hereinabove, and will not be repeated herein.

In one embodiment, as shown in fig. 5, there is provided a neutron nugget model adjustment device, including: a first acquisition module 501, a second acquisition module 502, a first determination module 503, a third acquisition module 504, and a fourth acquisition module 505, wherein:

the first acquisition module 501 is configured to acquire a fast group neutron flux distribution measurement and a thermal group neutron flux distribution measurement of a target fuel segment.

The second acquisition module 502 is configured to acquire an initial equivalent cross-section of the target fuel segment, where the initial equivalent cross-section includes an initial fast-group macroscopic absorption cross-section, an initial thermal-group macroscopic absorption cross-section, and an initial downward macroscopic scattering cross-section.

As nuclear reactions occur in the nuclear reactor at any time, all nuclear reactions are classified, nine reaction types can be obtained, each reaction type can be equivalently a macroscopic cross section, the probability of a certain specific nuclear reaction between an incident particle and a target is indicated, and nine equivalent cross sections are respectively: fast group diffusion coefficient, thermal group diffusion coefficient, downward macroscopic scattering section, fast group macroscopic absorption section, thermal group macroscopic absorption section, fast group macroscopic neutron generation section, thermal group macroscopic neutron generation section, fast group macroscopic energy generation section, and thermal group macroscopic energy generation section. Since the downward macroscopic scattering section, the fast-group macroscopic absorption section and the thermal-group macroscopic absorption section are more modified in actual operation, only the acquisition and modification of the initial sections of the downward macroscopic scattering section, the fast-group macroscopic absorption section and the thermal-group macroscopic absorption section are described in the application, and in actual application, technicians can acquire other equivalent sections and modify other equivalent sections according to similar methods in the application as required to adjust the neutron segment model.

The first determining module 503 is configured to determine each equivalent cross-sectional correction amount based on the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value, the first preset relationship, and the second preset relationship.

The third obtaining module 504 is configured to adjust each initial equivalent cross section according to each equivalent cross section correction amount, and obtain each target equivalent cross section.

The fourth obtaining module 505 is configured to adjust an initial neutron segment model of the target segment based on each target equivalent section to obtain the target neutron segment model.

The initial neutron segment model of the target fuel segment can be adjusted according to the target fast-group macroscopic absorption section, the target thermal-group macroscopic absorption section and the target downward macroscopic scattering section to obtain the target neutron segment model, so that the gap between the theoretical value and the measured value of the fast-group neutron flux and the thermal-group neutron flux is reduced, and the gap between the theoretical value and the measured value of the power distribution is reduced.

According to the neutron segment model adjusting device, the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value and the initial equivalent cross section of the target fuel segment are obtained, the equivalent cross section correction amounts are determined based on the fast group neutron flux distribution measurement value, the thermal group neutron flux distribution measurement value, the first preset relation and the second preset relation, the initial fast group macroscopic absorption cross section, the initial thermal group macroscopic absorption cross section and the initial downward macroscopic scattering cross section of the target fuel segment are respectively subjected to cross section correction through the equivalent cross section correction amounts, so that each target equivalent cross section with higher accuracy is obtained, and finally the initial neutron segment model of the target segment is adjusted based on each target equivalent cross section, so that the target neutron segment model is obtained. The method does not need to reduce the power level of the reactor core of the nuclear power station, avoids affecting the economy of the nuclear power station, adjusts the initial neutron segment model based on the measured value of the target fuel segment, does not need a large amount of experimental data, and avoids the accuracy reduction of model parameters caused by the inaccuracy of the data when the data are too much.

In one embodiment, the first acquisition module includes: the device comprises a first acquisition sub-module, a second acquisition sub-module and a third acquisition sub-module.

The first acquisition submodule is used for acquiring a fast group neutron flux distribution theoretical value and a thermal group neutron flux distribution theoretical value of the target fuel segment according to the initial neutron segment model.

The second acquisition sub-module is used for acquiring a power distribution measured value of the target fuel segment.

The third obtaining submodule is used for obtaining the fast group neutron flux distribution measured value and the thermal group neutron flux distribution measured value of the target fuel segment according to the first preset formula, the second preset formula, the fast group neutron flux distribution theoretical value, the thermal group neutron flux distribution theoretical value and the power distribution measured value of the target fuel segment

In one embodiment, the second acquisition module includes: the system comprises a fourth acquisition sub-module, a fifth acquisition sub-module and a sixth acquisition sub-module.

The fourth obtaining sub-module is configured to obtain an equivalent cross section information base, where the equivalent cross section information base includes a first mapping relationship, a second mapping relationship, and a third mapping relationship, the first mapping relationship includes a mapping relationship between a fast-group macroscopic absorption cross section of the target fuel segment and a first target state parameter corresponding to the fast-group macroscopic absorption cross section in each state parameter of the target fuel segment, the second mapping relationship includes a mapping relationship between a thermal-group macroscopic absorption cross section of the target fuel segment and a second target state parameter corresponding to the thermal-group macroscopic absorption cross section in each state parameter of the target fuel segment, and the third mapping relationship includes a mapping relationship between a downward macroscopic scattering cross section of the target fuel segment and a third target state parameter corresponding to the downward macroscopic scattering cross section in each state parameter of the target fuel segment.

The fifth acquisition submodule is used for acquiring a first expansion coefficient of the initial fast group macroscopic absorption section according to the first target state parameter, acquiring a second expansion coefficient of the initial thermal group macroscopic absorption section according to the second target state parameter, and acquiring a third expansion coefficient of the initial downward macroscopic scattering section according to the third target state parameter.

The sixth acquisition sub-module is used for acquiring an initial fast group macroscopic absorption section according to the first target state parameter and the first expansion coefficient, acquiring an initial thermal group macroscopic absorption section according to the second target state parameter and the second expansion coefficient, and acquiring an initial downward macroscopic scattering section according to the third target state parameter and the third expansion coefficient

In one embodiment, the first determination module includes: the seventh acquisition sub-module, the eighth acquisition sub-module, and the ninth acquisition sub-module.

The seventh obtaining submodule is used for substituting the first correction relation and the third correction relation into the first formula to be balanced to obtain a first target relation, substituting the second correction relation and the third correction relation into the second formula to be balanced to obtain a second target relation.

The eighth obtaining sub-module is configured to obtain each first target fitting coefficient according to a first target relational expression, obtain each second target fitting coefficient according to a second target relational expression, and obtain each third target fitting coefficient according to a third target relational expression, where the third target relational expression is one of the first target relational expression and the second target relational expression.

The ninth obtaining submodule is used for obtaining the macroscopic absorption cross section correction quantity of the fast group according to the first target fitting coefficient and the first correction relation, obtaining the macroscopic absorption cross section correction quantity of the heat group according to the second target fitting coefficient and the second correction relation, and obtaining the macroscopic scattering cross section correction quantity downwards according to the third target fitting coefficient and the third correction relation.

The above-mentioned each module in the neutron segment model adjustment device may be implemented in whole or in part by software, hardware, and a combination thereof. The above modules may be embedded in hardware or may be independent of a processor in the computer device, or may be stored in software in a memory in the computer device, so that the processor may call and execute operations corresponding to the above modules.

In one embodiment, a computer device is provided, comprising a memory and a processor, the memory having stored therein a computer program, the processor implementing the steps of the method embodiments described above when the computer program is executed.

The computer device may be a terminal, and its internal structure may be as shown in fig. 6. The computer device includes a processor, a memory, an input/output interface, a communication interface, a display unit, and an input means. The processor, the memory and the input/output interface are connected through a system bus, and the communication interface, the display unit and the input device are connected to the system bus through the input/output interface. Wherein the processor of the computer device is configured to provide computing and control capabilities. The memory of the computer device includes a non-volatile storage medium and an internal memory. The non-volatile storage medium stores an operating system and a computer program. The internal memory provides an environment for the operation of the operating system and computer programs in the non-volatile storage media. The input/output interface of the computer device is used to exchange information between the processor and the external device. The communication interface of the computer device is used for carrying out wired or wireless communication with an external terminal, and the wireless mode can be realized through WIFI, a mobile cellular network, NFC (near field communication) or other technologies. The computer program, when executed by a processor, implements a neutron segment model tuning method. The display unit of the computer device is used for forming a visual picture, and can be a display screen, a projection device or a virtual reality imaging device. The display screen can be a liquid crystal display screen or an electronic ink display screen, and the input device of the computer equipment can be a touch layer covered on the display screen, can also be a key, a track ball or a touch pad arranged on the shell of the computer equipment, and can also be an external keyboard, a touch pad or a mouse and the like.

It will be appreciated by those skilled in the art that the structure shown in fig. 6 is merely a block diagram of some of the structures associated with the present application and is not limiting of the computer device to which the present application may be applied, and that a particular computer device may include more or fewer components than shown, or may combine certain components, or have a different arrangement of components.

In one embodiment, a computer-readable storage medium is provided, on which a computer program is stored which, when executed by a processor, implements the steps of the method embodiments described above.

Those skilled in the art will appreciate that implementing all or part of the above described methods may be accomplished by way of a computer program stored on a non-transitory computer readable storage medium, which when executed, may comprise the steps of the embodiments of the methods described above. Any reference to memory, database, or other medium used in the various embodiments provided herein may include at least one of non-volatile and volatile memory. The nonvolatile Memory may include Read-Only Memory (ROM), magnetic tape, floppy disk, flash Memory, optical Memory, high density embedded nonvolatile Memory, resistive random access Memory (ReRAM), magnetic random access Memory (Magnetoresistive Random Access Memory, MRAM), ferroelectric Memory (Ferroelectric Random Access Memory, FRAM), phase change Memory (Phase Change Memory, PCM), graphene Memory, and the like. Volatile memory can include random access memory (Random Access Memory, RAM) or external cache memory, and the like. By way of illustration, and not limitation, RAM can be in the form of a variety of forms, such as static random access memory (Static Random Access Memory, SRAM) or dynamic random access memory (Dynamic Random Access Memory, DRAM), and the like. The databases referred to in the various embodiments provided herein may include at least one of relational databases and non-relational databases. The non-relational database may include, but is not limited to, a blockchain-based distributed database, and the like. The processors referred to in the embodiments provided herein may be general purpose processors, central processing units, graphics processors, digital signal processors, programmable logic units, quantum computing-based data processing logic units, etc., without being limited thereto.

The technical features of the above embodiments may be arbitrarily combined, and all possible combinations of the technical features in the above embodiments are not described for brevity of description, however, as long as there is no contradiction between the combinations of the technical features, they should be considered as the scope of the description.

The above examples only represent a few embodiments of the present application, which are described in more detail and are not to be construed as limiting the scope of the present application. It should be noted that it would be apparent to those skilled in the art that various modifications and improvements could be made without departing from the spirit of the present application, which would be within the scope of the present application. Accordingly, the scope of protection of the present application shall be subject to the appended claims.

Claims

1. A neutron nugget model adjustment method applied to a reactor core, the reactor core comprising a fuel nugget, the fuel nugget comprising fast neutrons and thermal neutrons, the method comprising:

2. The method of claim 1, wherein the obtaining fast group neutron flux distribution measurement and thermal group neutron flux distribution measurement of the target fuel segment comprises:

acquiring a power distribution measurement value of the target fuel segment;

3. The method of claim 1, wherein the obtaining the initial equivalent cross-section of the target fuel segment comprises:

4. The method according to claim 3, wherein the first preset relation includes a first correction relation, a second correction relation and a third correction relation, wherein the first correction relation is a correlation between a fast-group macroscopic absorption cross-section correction amount, a fitting coefficient corresponding to the fast-group macroscopic absorption cross-section correction amount and the first target state parameter, the second correction relation is a correlation between a thermal-group macroscopic absorption cross-section correction amount, a fitting coefficient corresponding to the thermal-group macroscopic absorption cross-section correction amount and the second target state parameter, and the third correction relation is a correlation between a downward macroscopic scattering cross-section correction amount, a fitting coefficient corresponding to the thermal-group macroscopic absorption cross-section correction amount and the third target state parameter.

5. The method of claim 4, wherein the second predetermined relationship comprises: the first formula to be balanced and the second formula to be balanced, wherein the first formula to be balanced is a correlation between a preset fast neutron outflow amount, a preset fast neutron production amount, a fast group neutron flux distribution measurement value, an initial fast group macroscopic absorption section, a fast group macroscopic absorption section correction amount, an initial downward macroscopic scattering section, a downward macroscopic scattering section correction amount and a first flux error, and the second formula to be balanced is a correlation between a preset thermal neutron outflow amount, a thermal group neutron flux distribution measurement value, an initial thermal group macroscopic absorption section, a thermal group macroscopic absorption section correction amount, an initial downward macroscopic scattering section, a downward macroscopic scattering section correction amount and a second flux error.

6. The method of claim 5, wherein determining the equivalent cross-sectional corrections based on the fast group neutron flux distribution measurement, the thermal group neutron flux distribution measurement, the first predetermined relationship, and the second predetermined relationship comprises:

7. The method for adjusting a neutron segment model according to claim 2, wherein the first preset formula is:

wherein,a theoretical value of neutron flux distribution in said fast group for said target fuel segment n,>for the target fuel sectionTheoretical value of neutron flux distribution in said thermal group for block n,>for the fast group neutron flux distribution measurement value of the target fuel segment n, +.>Neutron flux distribution measurements for the thermal group of the target fuel segment n;

The second preset formula is:

wherein,generating a cross section for a preset fast group macroscopic energy of said target fuel segment n,/for a target fuel segment n>Generating a cross section for the preset heat mass macroscopic energy of the target fuel segment n, < ->Is the power distribution measurement of the target fuel segment n.

8. A neutron block model adjustment device, the device comprising: the first acquisition module is used for acquiring a fast group neutron flux distribution measured value and a thermal group neutron flux distribution measured value of the target fuel segment;

9. A computer device comprising a memory and a processor, the memory storing a computer program, characterized in that the processor implements the steps of the method of any of claims 1 to 7 when the computer program is executed.

10. A computer readable storage medium, on which a computer program is stored, characterized in that the computer program, when being executed by a processor, implements the steps of the method of any of claims 1 to 7.