Method, system, equipment and medium for identifying number and size of damaged reactor fuel cladding

Abstract

The invention provides a method, system, equipment and medium for identifying the number and size of reactor fuel cladding damage, and the method comprises the steps: S1, obtaining the unit crevasse release amount reference data of each nuclide based on transient release modeling experiments of different groove sizes; s2, obtaining the total release amount of nuclides based on the obtained unit crevasse release amount reference data of each nuclide; and S3, constructing a matrix reversibility constraint, and realizing the unique solution inversion of the damage parameter. The method can improve the accuracy and efficiency of fuel cladding damage detection, provides reliable method support for safe operation of a reactor, and can be widely applied to monitoring and evaluation of the safety of the fuel assembly of the pressurized water reactor nuclear power station.

Description

TECHNICAL FIELD

[0001] The present application relates to the technical field of nuclear reactor safety monitoring, in particular to a detection method and system for identifying the number and size of reactor fuel cladding breakage. BACKGROUND

[0002] Nuclear energy plays an important role in China's energy security strategy. As a clean and efficient energy choice, nuclear power plants bear a great responsibility during operation. However, the fuel rods inside the reactor continuously accumulate radioactive products during operation, and the fuel cladding may be damaged due to various physical and chemical factors. Once the cladding is damaged, the number and size of the damage must be evaluated quickly through detection data to determine whether the reactor needs to be shut down and what emergency measures to take. In order to accurately evaluate the damage, efficient algorithms need to be developed to process and analyze relevant data and establish a mapping relationship between damage characteristics and detection signals.

[0003] Currently, most fuel cladding breakage detection methods internationally rely on experimental reactors or simulation experiments under fixed conditions. These methods usually require complex experimental equipment and high costs, and the scalability and applicability of experimental results are limited. Existing detection technologies mainly use different nuclide specific activity ratios and neural network algorithms to identify the extent of damage. Although these methods can determine whether the cladding is damaged to some extent, they cannot accurately identify the number and size of the breakage, making it difficult to provide more detailed damage information. In addition, traditional algorithms have problems such as insufficient precision and low efficiency when dealing with complex data, making it difficult to meet actual needs. In order to overcome these limitations, it is urgent to develop more flexible and efficient algorithms that can accurately identify and quantify the number and size of fuel cladding breakage under a wider range of conditions, thereby improving the safety and reliability of nuclear power plant operation.

[0004] Patent document CN115965254A (application number: 202211083298.8) discloses a sodium-cooled fast reactor core damage evaluation method based on radioactive nuclide sampling. The method calculates the release amount of nuclides at the sampling time of the sodium-cooled fast reactor and the decay of the parent nucleus at the sampling time, takes the activity of the third generation of decay daughter bodies to obtain the release fraction of the nuclides at the sampling time of the sodium-cooled fast reactor under unexpected accident conditions. The calculated nuclide release fraction is compared with the actual nuclide release fraction under unexpected accidents to determine the type of sodium-cooled fast reactor core damage, which includes three types of core damage: no damage, fuel cladding gas tightness damage, and fuel cladding damage. The determined core type is used to quickly guide safety protection and rescue measures in emergency situations.

[0005] Therefore, the purpose of the present application is to provide a reactor fuel cladding breakage quantity and size detection method and system, which improves the detection accuracy and efficiency by combining the experimental law with the linear equation set, and provides more reliable decision support for the safe operation of nuclear power plants. SUMMARY

[0006] In view of the defects in the prior art, the purpose of the present application is to provide a method, system, device and medium for identifying the number and size of reactor fuel cladding breakage.

[0007] According to the method for identifying the number and size of reactor fuel cladding breakage provided by the present application, the method comprises the following steps:

[0008] Step S1: Based on the transient release modeling experiment of different bevel sizes, the unit break release amount reference data of each nuclide is obtained;

[0009] Step S2: Based on the unit break release amount reference data of each nuclide obtained, the total release amount of the nuclide is obtained.

[0010] Step S3: The matrix invertibility constraint is constructed to realize the unique solution inversion of the breakage parameters.

[0011] Preferably, the step S1 comprises: for each break size, the unit release amount of the nuclide corresponding to the sampling time is measured by the modeling experiment table; and the bevel size is consistent with the selected nuclide quantity.

[0012] Preferably, the step S2 comprises: the total release amount of the nuclide is the linear superposition of the release amount of each break size at the corresponding time and the corresponding break size quantity;

[0013]

[0014] Wherein, a, b, c are the breakage quantities to be solved, n ii And the unit release amount measured by the modeling experiment.

[0015] Preferably, the step S3 comprises: the modeling experiment is performed by reasonably designing the break size (d1, d2, d3), so as to ensure that the determinant of the coefficient matrix composed of the unit release amount obtained from the experiment is non-zero;

[0016]

[0017] So as to ensure that the equation set has a unique solution, and avoid the problems of multiple solutions or no solution caused by matrix.

[0018] According to the system for identifying the number and size of reactor fuel cladding breakage provided by the present application, the system comprises:

[0019] Module M1: Based on the transient release modeling experiment of different bevel sizes, the unit break release amount reference data of each nuclide is obtained;

[0020] Module M2: based on the obtained nuclide unit break release amount reference data to obtain the total release amount of nuclides;

[0021] Module M3: construct matrix invertibility constraints to realize unique solution inversion of damage parameters.

[0022] Preferably, the module M1 comprises: for each break size, the unit release amount of the corresponding sampling time is determined by the modeling experiment platform; the groove size is consistent with the selected number of nuclides.

[0023] Preferably, the module M2 comprises: the total release amount of nuclides is the linear superposition of the release amount at the corresponding time under each break size and the corresponding break size number;

[0024]

[0025] Wherein, a, b, c are the break sizes to be solved, n ii And the unit release amount is measured by the modeling experiment.

[0026] Preferably, the module M3 comprises: by reasonably designing the break size (d1, d2, d3) to carry out the modeling experiment, it is ensured that the determinant of the coefficient matrix composed of the unit release amount obtained from the experiment is non-zero;

[0027]

[0028] So as to ensure that the equation set has a unique solution, and avoid the problem of multiple solutions or no solution caused by matrix.

[0029] According to the method provided by the application, the computer readable storage medium storing the computer program is provided, and the computer program is executed by the processor to realize the steps of the method for identifying the number and size of the fuel cladding damage of the reactor.

[0030] According to the method provided by the application, the electronic device is provided, which comprises a memory, a processor and a computer program stored in the memory and capable of running on the processor, and the computer program is executed by the processor to realize the steps of the method for identifying the number and size of the fuel cladding damage of the reactor.

[0031] Compared with the prior art, the application has the beneficial effects as follows:

[0032] 1. The application can improve the precision and efficiency of fuel cladding damage detection, provide reliable method support for safe operation of the reactor, and can be widely applied to the monitoring and evaluation of the safety of fuel assemblies of pressurized water reactor nuclear power plants;

[0033] 2. Compared with the method of relying on the nuclide activity ratio (such as 131I / 133Compared with the prior art of Xe) or neural network model, the present application realizes transparent calculation and physical interpretability of the size and number of damage by a linear inversion model driven by experimental data, while avoiding the uncertainty of nonlinear fitting, thereby providing a more efficient and robust solution for nuclear reactor damage diagnosis.

[0034] 3、 The present application directly solves the size and number of the breach by combining the detection data, thereby providing quantitative and high-precision damage parameters for reactor safety evaluation. BRIEF DESCRIPTION OF DRAWINGS

[0035] Other features, objects, and advantages of the present application will become more apparent from the following detailed description of non-limiting embodiments with reference to the following drawings:

[0036] Figure 1 Flowchart of a method for identifying the number and size of reactor fuel cladding damage. DETAILED DESCRIPTION

[0037] The present application will be described in detail below with reference to specific embodiments. The following examples will help those skilled in the art to further understand the present application, but do not limit the present application in any form. It should be noted that those skilled in the art can make several changes and improvements without departing from the concept of the present application. These are within the scope of the present application.

[0038] Example 1

[0039] According to the method for identifying the number and size of reactor fuel cladding damage provided by the present application, as shown in Figure 1 , it comprises: by constructing the linear mapping relationship between the nuclide release amount and the breach parameter, combining the matrix invertibility constraint, realizing the unique solution inversion of the damage parameter.

[0040] Step 1: Based on the transient release modeling experiment of different breach sizes (d1, d2, d3), the unit breach release amount reference data of each nuclide (such as Cs, Kr, etc.) is obtained, and the following mapping relationship is formed:

[0041] For each breach size such as d1, the unit release amount of the nuclide corresponding to the sampling time is measured by the modeling experiment table, denoted as n 11 (I), n 12 (Cs), n 13 (Kr), etc., the breach size d i should be consistent with the number of selected nuclides;

[0042] The total release amount of the nuclide detected in the actual project (∑n i1 , ∑n i2 , ∑n i3) is the linear superposition of the release amount corresponding to the time under each break size and the number of its corresponding break size (a, b, c).

[0043] The total release amount is expressed as a linear equation of the number of breaks, and the following equation group is established:

[0044]

[0045] Where a, b, c are the number of breaks to be solved, n ii and are the unit release amounts measured by the modeling experiment.

[0046] Step 2: Perform a modeling experiment by reasonably designing the break size (d1, d2, d3) to ensure that the determinant of the coefficient matrix composed of the unit release amounts obtained from the experiment is non-zero, i.e.

[0047]

[0048] Thus, it is ensured that the equation group has a unique solution, avoiding the problem of multiple or no solutions due to matrix problems.

[0049] This embodiment can realize linear inversion, abandon the non-linear assumption of the traditional ratio method, and directly correlate the total release amount with the number and size of breaks through a linear model, significantly reducing error accumulation;

[0050] Reversibility design can be realized, the optimal selection of break size in the experimental stage is ensured to ensure the reversibility of the coefficient matrix, and the reliability of the solution is fundamentally improved; Multi-parameter adaptability: supporting complex scenarios of multiple break sizes (d1, d2, d3…) and multiple nuclides (Cs, Kr, Xe, etc.), which can be extended to higher dimensions.

[0051] The application also provides a system for identifying the number and size of reactor fuel cladding breaks, which can be realized by executing the flow steps of the method for identifying the number and size of reactor fuel cladding breaks, i.e. those skilled in the art can understand the method for identifying the number and size of reactor fuel cladding breaks as the preferred embodiment of the system for identifying the number and size of reactor fuel cladding breaks.

[0052] Example 2

[0053] Example 2 is a preferred example of example 1

[0054] According to the method for identifying the number and size of reactor fuel cladding breaks provided by the application, the method comprises the following steps:

[0055] First, the experimental data acquisition and benchmark database establishment. Select the size of the breach parameters to be analyzed (for example, d1 = 50 μm, d2 = 100 μm, d3 = 200 μm), to ensure the number of breach size and target nuclide species (such as Cs, Kr, I) is consistent. In the simulation of reactor operating conditions of the modeling experiment table, each breach size is transient release experiment, respectively, to determine the amount of each nuclide at a specific time point per breach release. For example, the breach size d1 corresponding to the release of nuclide I, Cs, Kr is n 11 , n 12 , n 13 , the breach size d2 corresponding to the release amount is n 21 , n 22 , n 23 , the breach size d3 corresponding to the release amount is n 31 , n 32 , n 33 , forming a benchmark database.

[0056] Second, the construction of linear inversion equation set. In the actual engineering detection, the total release amount of each nuclide (∑n i1 , ∑n i2 , ∑n i3 ) is the linear superposition of the release amount of each breach size. Let the breach size d1, d2, d3 corresponding to the breach number is a, b, c, the equation set is established:

[0057]

[0058] The equation set is converted to matrix form N·x = S, where the coefficient matrix N is composed of experimental benchmark data, x is the breach number vector to be solved, and S is the total release amount vector obtained by engineering measurement.

[0059] Subsequently, the matrix invertibility verification and unique solution inversion. Calculate the determinant det(N) of the coefficient matrix N, if the determinant is not zero (det(N)≠0), it indicates that the matrix is invertible, and the equation set has a unique solution. The breach number is calculated by matrix inversion or Gaussian elimination method, that is, x = N -1 ·S, the results a, b, c are the actual number of each breach size.

[0060] Finally, the results are verified and optimized. Check the physical rationality of the breach number a, b, c (such as non-negative integer characteristics), if negative or non-integer solution appears, the experimental data needs to be rechecked or the breach size design needs to be optimized. By adjusting the breach size group or increasing the nuclide species, the experiment and inversion process are repeated, which can further improve the inversion accuracy.

[0061] Implementation example:

[0062] For example, the total release amount ∑ni1 = 50,∑n i2 = 120,∑n i3 = 80, the experimental benchmark matrix is:

[0063]

[0064] By x = N -1 S, a = 10, b = 5, c = 5, indicating that there are 10 d1 size cracks, 5 d2 size cracks, and 5 d3 size cracks.

[0065] The method realizes efficient and accurate inversion of the damage parameters by linear inversion model and matrix invertibility constraints, avoids the multi-solution problem of traditional methods, and provides a reliable basis for reactor safety evaluation.

[0066] Those skilled in the art know that, in addition to implementing the system provided by the present application and each device, module, unit thereof in the form of pure computer readable program code, the system provided by the present application and each device, module, unit thereof can also be realized in the form of logic gates, switches, application specific integrated circuits, programmable logic controllers and embedded microcontrollers by logically programming the method steps to achieve the same function. Therefore, the system provided by the present application and each device, module, unit thereof can be considered as a hardware component, and the devices, modules, units included therein for realizing various functions can also be considered as structures within the hardware component; the devices, modules, units for realizing various functions can also be considered as both software modules realizing the method and structures within the hardware component.

[0067] The specific embodiments of the present application are described above. It should be understood that the present application is not limited to the above specific embodiments, and those skilled in the art can make various changes or modifications within the scope of the claims, which does not affect the essential content of the present application. In the case of no conflict, the embodiments of the present application and the features in the embodiments can be combined with each other arbitrarily.

Claims

1. A method of identifying the number and size of reactor fuel cladding breaches, comprising: The method comprises the following steps: Step S1: obtaining the unit breach release amount of each nuclide based on the transient release modeling experiment of different groove sizes; Step S2: obtaining the total release amount of the nuclides based on the obtained unit breach release amount of each nuclide; Step S3: constructing a matrix invertibility constraint to realize the unique solution inversion of the damage parameters.

2. The method of claim 1, wherein, The step S1 comprises: for each breach size, determining the unit release amount of the nuclides at the corresponding sampling time through the modeling experiment table; and the groove size is consistent with the number of selected nuclides.

3. The method of claim 1, wherein: The step S2 comprises: the total release amount of the nuclides is the linear superposition of the release amounts at the corresponding times under each breach size and the number of the corresponding breach sizes; Wherein, a, b, c are the number of breaks to be solved, n ii And so on are the unit release measured by the simulation experiment.

4. The method of claim 1, wherein, The step S3 comprises: performing the modeling experiment by reasonably designing the breach sizes (d1, d2, d3) to ensure that the determinant of the coefficient matrix composed of the unit release amounts obtained from the experiment is non-zero; Thus, the existence of a unique solution of the equation set is ensured, and the problems of multiple solutions or no solution caused by matrix problems are avoided.

5. A system for identifying the number and size of reactor fuel cladding breaches, comprising: The method comprises the following steps: Module M1: obtaining the unit breach release amount of each nuclide based on the transient release modeling experiment of different groove sizes; Module M2: obtaining the total release amount of the nuclides based on the obtained unit breach release amount of each nuclide; Module M3: constructing a matrix invertibility constraint to realize the unique solution inversion of the damage parameters.

6. The system for identifying the number and size of breaches in a reactor fuel cladding of claim 5, wherein, The module M1 comprises: for each breach size, determining the unit release amount of the nuclides at the corresponding sampling time through the modeling experiment table; and the groove size is consistent with the number of selected nuclides.

7. The system for identifying the number and size of breaches in a reactor fuel cladding of claim 5, wherein, The module M2 comprises: the total release amount of the nuclides is the linear superposition of the release amounts at the corresponding times under each breach size and the number of the corresponding breach sizes; Wherein, a, b, c are the number of breaks to be solved, n ii And so on are the unit release measured by the simulation experiment.

8. The system for identifying the number and size of breaches in a reactor fuel cladding of claim 7, wherein, The module M3 comprises: performing the modeling experiment by reasonably designing the breach sizes (d1, d2, d3) to ensure that the determinant of the coefficient matrix composed of the unit release amounts obtained from the experiment is non-zero; Thus, the existence of a unique solution of the equation set is ensured, and the problems of multiple solutions or no solution caused by matrix problems are avoided.

9. A computer readable storage medium storing a computer program, characterized in that, The computer program is executed by the processor to realize the steps of the method for identifying the number and size of the fuel cladding damage of the reactor according to any one of claims 1 to 4.

10. An electronic device comprising a memory, a processor, and a computer program stored on the memory and executable on the processor, characterized in that, The computer program is executed by the processor to realize the steps of the method for identifying the number and size of the fuel cladding damage of the reactor according to any one of claims 1 to 4.

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