Knowledge graph-based nuclear fuel assembly damage reason analysis system and method

Abstract

The invention relates to the field of fuel damage diagnosis, in particular to a nuclear fuel assembly damage reason analysis system and method based on a knowledge graph, and the system comprises a data obtaining module which obtains unit information and operation data, the unit information and the operation data are transmitted to a loop radiochemical analysis module, a fuel performance analysis module and a damage reason analysis module; the primary loop radiochemical analysis module judges whether the fuel assembly in the reactor core is damaged or not through screening and calculation analysis, and obtains the damage information of the damaged fuel rod; the fuel performance analysis module performs fuel performance multi-physical coupling calculation to obtain a physical quantity change trend and a failure criterion of the damaged fuel rod; and the damage reason analysis module obtains the damaged fuel rod information and analyzes the fuel rod damage reason. According to the system, the position and the condition of the damaged fuel rod are speculated by arranging the loop radiochemical analysis module, so that the damaged fuel rod is quickly positioned, precious time can be provided for emergency refueling of a reactor core, and meanwhile, a critical path of overhaul is also shortened.

Description

Technical Field

[0001] The present invention relates to the field of fuel damage diagnosis, and in particular to a system and method for analyzing the causes of nuclear fuel assembly damage based on a knowledge graph. Background Art

[0002] Nuclear fuel assemblies are core components of nuclear reactors, and their integrity and reliability are directly related to the safe operation of nuclear power plants. However, during long-term operation, nuclear fuel assemblies may fail due to a variety of reasons. These reasons include debris abrasion, fouling or corrosion, pellet-cladding interaction, and stress corrosion cracking. With the development of nuclear power technology, despite continuous improvements in fuel assembly design and materials, failures still occur from time to time. This not only affects the operating efficiency of nuclear power plants but also poses potential safety risks. As the operating time of domestic nuclear power plants increases, the number of nuclear fuel assembly failures has also increased, and the subsequent problems caused by fuel assembly failure have become more prominent. Rapidly and accurately locating damaged fuel assemblies not only provides valuable time for emergency core refueling, but also shortens the critical path for overhauls.

[0003] In the prior art, there is no complete diagnostic system for the technology of quickly and accurately locating damaged fuel assemblies, and there is currently a lack of a nuclear fuel assembly damage cause analysis system. Summary of the Invention

[0004] The present invention provides a system and method for analyzing the causes of nuclear fuel assembly damage based on a knowledge graph, which is used to solve the problem in the prior art of lacking a system for quickly locating and analyzing nuclear fuel assembly damage.

[0005] The technical solutions of the present invention are as follows:

[0006] The present invention proposes a nuclear fuel assembly damage cause analysis system based on knowledge graph, which includes a data acquisition module, a primary radiochemical analysis module, a fuel performance analysis module and a damage cause analysis module. The data acquisition module acquires unit information and operation data, and transmits the unit information and operation data to the primary radiochemical analysis module, the fuel performance analysis module and the damage cause analysis module; the primary radiochemical analysis module determines whether fuel assembly damage occurs inside the core by screening and calculating and analyzing the unit information and operation data, and obtains damage information of the damaged fuel rod; the fuel performance analysis module performs multi-physics coupling calculation of fuel performance based on the unit information and operation data, and obtains the physical quantity change trend and failure criterion of the damaged fuel rod; the damage cause analysis module acquires the unit information and operation data of the data acquisition module, the calculation results of the fuel performance analysis module and the fuel assembly inspection results provided by the maintenance personnel, and analyzes the cause of the fuel rod damage.

[0007] In some embodiments, the unit information and operating data acquired by the data acquisition module include: core parameter input, core burnup distribution input, primary loop radiochemical parameter input and operating parameter input; the damage information of the damaged fuel rod includes the damaged fuel rod damage index, the number of damaged fuel rods, the damaged fuel burnup, the core position of the damaged fuel and the damaged fuel rod rupture size.

[0008] In some embodiments, the core parameter input specifically includes the unit name, unit thermal power, primary coolant volume, fuel heat release rate, number of fuel rods, core active area height, and fuel pellet diameter; the core burnup distribution input includes the burnup of the fuel assembly and the enrichment of the fuel assembly; the primary radiochemical parameter input includes the specific activity of 12 nuclides, namely Cs-134, Cs-137, I-131, I-132, I-133, I-134, I-135, Xe-133, Xe-135, Kr-85, Kr-87, and Kr-88. The type of nuclides can be adjusted according to needs; the operating parameter input specifically includes the cycle name, input relative power, average moderator temperature, ion adsorption bed purification efficiency, xenon removal rate, number of core fuel enrichment zones, average burnup depth of each enrichment zone, and enrichment of each enrichment zone.

[0009] In some embodiments, the primary radiochemical analysis module determines whether fuel assembly damage occurs inside the core through the specific activities of Cs-134 and Cs-137 input by the primary radiochemical parameters. If the specific activities of Cs-134 and Cs-137 are greater than 0, it is determined that fuel assembly damage occurs inside the core.

[0010] In some embodiments, the primary radiochemical analysis module calculates the specific activity ratio of Cs-134 and Cs-137 inputted in the primary radiochemical parameter input, and searches the specific activity ratio in the fuel damage burnup empirical table to obtain the damaged fuel rod burnup; the primary radiochemical analysis module matches the damaged fuel rod burnup with the burnup of the fuel assembly in the core burnup distribution input and the burnup information in the fuel assembly enrichment to determine the damaged fuel rod position and fuel rod enrichment; the primary radiochemical analysis module calculates the damaged fuel rod damage index using formula (1) and formula (2), and the specific formula is as follows:

[0011]

[0012] Among them, FRI p is the fuel damage index, A N,I-131 and A N,I-134 They represent the specific activities of iodine-131 and iodine-134 input into the radiochemical parameters of the first loop; k is the adhesion correction factor; L n : Linear power density, which is the power output per unit length of the fuel rod; L HCRis the linear power density at high burnup rate, that is, the power output per unit length of damaged fuel rods; Pr is the core relative power, which is obtained from unit information and operating data;

[0013]

[0014] Where Y is the fraction of the isotope in the fission product, that is, the proportion of a certain radioactive isotope produced during nuclear fission; λ is the decay constant, which is related to the half-life of the isotope; Bn is the universal purification rate constant; and k is the adhesion correction factor.

[0015] The primary circuit iodine analysis module obtains the specific activity of iodine isotopes and calculates the number of damaged fuel rods using formulas (3) and (4), as follows:

[0016]

[0017] Where R is the fission product release rate, B is the fission product production rate; Cm is the measured radioisotope concentration, which comes from the radioisotope specific activity in the input primary radiochemical parameters; Mc is the coolant mass; λ is the radioisotope decay constant; β p is the coolant purge rate constant; F f is the fuel fission rate, that is, how many fissions occur per second; Y is the cumulative fission fraction of the isotope, that is, how many atoms of the isotope are produced in each fission; x represents the number of damaged fuel rods; H is the geometric factor correction value, v is the interstitial escape rate coefficient; D ′ is the empirical diffusion coefficient; c is a constant; the specific activities of I-131, I-132, I-133, I-134, and I-135 are substituted into the above formula (3) to obtain the corresponding calculated value R / B, and R / B is substituted into formula (4) to form a system of equations, which are solved to obtain the value of x, where x represents the number of damaged fuel rods.

[0018] The primary circuit iodine analysis module obtains the specific activity of iodine isotopes and calculates the rupture size of the damaged fuel rod using the empirical formula (6), as follows:

[0019]

[0020] Where E is the rupture size; vf is the iodine isotope interstitial escape rate coefficient; is the length of the fuel rod; and n is the damage position parameter, where 1 indicates middle damage and 0.5 indicates end damage. The parameter should be selected based on actual conditions.

[0021] In some embodiments, the primary radiochemical analysis module calculates the number of damaged fuel rods using formula (5), as follows:

[0022]

[0023] Among them, D′ I =10 9.857log P-25.1314 D′ NG =10 8.632log P-23.4091 A=x 2 D′, x represents the number of damaged fuel rods, D′ I is the empirical diffusion coefficient of the iodine isotope in a single breakage, D′ NG is the empirical diffusion coefficient of the rare gas in a single damage event; P is the linear power.

[0024] In some embodiments, the fuel damage burnup empirical table is obtained based on a fuel rod damage burnup test, and the fuel damage burnup empirical table includes a relationship between the damaged fuel rod burnup and the specific activity ratio of Cs-134 to Cs-137.

[0025] In some embodiments, the fuel performance analysis module obtains the changing trends of stress, strain and fission gas release through multi-physics coupling calculation of fuel performance, and calculates the two failure criteria of output cumulative damage and total deformation, and displays the changing trends of stress, strain and fission gas release, cumulative damage and total deformation to the user in the form of graphs; the multi-physics coupling calculation of fuel performance specifically includes the fuel performance analysis module establishing a finite element model of mechanical and heat transfer coupling for the fuel rod, dividing the fuel rod into axisymmetric layers and non-axisymmetric layers along the axial direction, and realizing interlayer coupling through air cavity gas parameters; the fuel performance analysis module inputs the unit information and operation data into the finite element model, and obtains the changing trends of stress, strain and fission gas release, output cumulative damage and total deformation through simulation calculation.

[0026] In some embodiments, the damage cause analysis module collects cases that have occurred in the world and have been determined to have caused the damage of fuel assemblies to build a case library. The damage cause analysis module summarizes the characteristics of the damage causes of each fuel rod based on the case library. The damage cause analysis module matches the acquired information with the characteristics of the fuel rod damage causes to determine the cause of fuel assembly damage. If the damage cause analysis module finds that the acquired information is consistent with multiple fuel damage causes, the multiple fuel damage causes are arranged in descending order of frequency of occurrence and displayed to the user.

[0027] In some embodiments, the fuel assembly inspection results provided by the maintenance personnel include damage shape, damage location, damage size, number of cycles experienced by the fuel assembly, fuel pellet structure, and pictures of damaged fuel rods.

[0028] The present invention proposes a method for analyzing the causes of nuclear fuel assembly damage based on a knowledge graph, which includes:

[0029] Step 1: The data acquisition module obtains unit information and operating data, and transmits the unit information and operating data to the primary circuit radiochemical analysis module, fuel performance analysis module and damage cause analysis module;

[0030] Step 2: The primary circuit radiochemical analysis module screens and calculates and analyzes the unit information and operating data to determine whether fuel assembly damage has occurred inside the core and obtain damage information of the damaged fuel rods;

[0031] Step 3: The fuel performance analysis module performs multi-physics coupling calculations on the fuel performance using the unit information and operating data to obtain the physical quantity change trend and failure criteria of the damaged fuel rods.

[0032] Step 4: Based on the damage information of the damaged fuel rod inferred by the primary radiochemical analysis module, conduct an on-site inspection of the damaged fuel assembly and obtain the fuel assembly inspection results;

[0033] Step 5: The damage cause analysis module obtains the unit information and operating data from the data acquisition module, the calculation results of the fuel performance analysis module, and the fuel assembly inspection results provided by the maintenance personnel. The damage cause analysis module collects cases that have occurred in the world and the causes of fuel assembly damage that have been determined to build a case library. The damage cause analysis module summarizes the characteristics of each fuel rod damage cause based on the case library. The damage cause analysis module matches the obtained information with the characteristics of the fuel rod damage cause to determine the cause of the fuel assembly damage.

[0034] The implementation of the present invention has the following beneficial effects:

[0035] This invention proposes a knowledge graph-based nuclear fuel assembly damage cause analysis system and method. This system uses a primary radiochemical analysis module to infer the location and condition of damaged fuel rods, allowing for rapid localization of damaged fuel rods. This not only provides valuable time for emergency core refueling but also shortens the critical path for overhauls. The damage cause analysis module of this system summarizes and organizes the corresponding characteristics of different fuel rod damage causes, analyzing the causes of damaged fuel rods. This effectively helps personnel understand the root cause of damaged fuel rods and provides strong support for preventing subsequent fuel rod damage. BRIEF DESCRIPTION OF THE DRAWINGS

[0036] Figure 1 This is a schematic diagram of a nuclear fuel assembly damage cause analysis system based on a knowledge graph proposed in an embodiment of the present invention;

[0037] Figure 2 This is a flow chart of the primary-loop radiochemical analysis module of a knowledge graph-based nuclear fuel assembly damage cause analysis system proposed in an embodiment of the present invention;

[0038] Figure 3This is a flow chart of the fuel performance analysis module of a knowledge graph-based nuclear fuel assembly damage cause analysis system proposed in an embodiment of the present invention;

[0039] Figure 4 This is a flow chart of a damage cause analysis module of a nuclear fuel assembly damage cause analysis system based on a knowledge graph proposed in an embodiment of the present invention;

[0040] Figure 5 A schematic diagram of core burnup distribution input for a nuclear fuel assembly damage cause analysis system based on a knowledge graph is proposed for an embodiment of the present invention. DETAILED DESCRIPTION

[0041] The technical solution of the present invention is clearly and completely described below in conjunction with the accompanying drawings and specific embodiments. Obviously, the embodiments described are only part of the embodiments of the present invention, not all of them. Based on the embodiments of the present invention, all other embodiments obtained by ordinary technicians in this field without making creative efforts are within the scope of protection of the present invention.

[0042] like Figures 1 to 5 As shown, the present invention proposes a nuclear fuel assembly damage cause analysis system based on knowledge graph, which includes a data acquisition module, a primary circuit radiochemical analysis module, a fuel performance analysis module and a damage cause analysis module. The data acquisition module acquires unit information and operation data, and transmits the unit information and operation data to the primary circuit nitridation analysis module, the fuel performance analysis module and the damage cause analysis module.

[0043] like Figure 2 As shown in the figure, the primary radiochemical analysis module determines whether fuel assembly damage has occurred inside the core by screening and calculating the unit information and operation data, and obtains the damage information of the damaged fuel rod. The unit information and operation data include core parameter input, core burnup distribution input, primary radiochemical parameter input and operation parameter input. The core parameter input specifically includes unit name, unit thermal power, core relative power, primary coolant volume, fuel heat release rate, number of fuel rods, core active area height, and fuel pellet diameter; Figure 5As shown in the figure, the core burnup distribution input includes the burnup of each fuel assembly and the fuel assembly enrichment according to the fuel assembly core loading diagram; the primary loop radiochemical parameter input includes the specific activity of 12 nuclides, namely Cs-134, Cs-137, I-131, I-132, I-133, I-134, I-135, Xe-133, Xe-135, Kr-85, Kr-87 and Kr-88. The type of nuclides can be adjusted according to needs; the operating parameter input specifically includes the cycle name, input relative power, average moderator temperature, ion adsorption bed purification efficiency, xenon removal rate, number of core fuel enrichment zones, average burnup depth of each enrichment zone, and enrichment of each enrichment zone.

[0044] The primary radiochemical analysis module determines whether a fuel assembly has been damaged within the core by measuring the specific activities of Cs-134 and Cs-137. These long-half-life, high-fission-yield radioactive fission products are the primary nuclides used to monitor and locate damaged fuel assemblies. Detection of Cs-134 and Cs-137 in the reactor's primary loop indicates a potential fuel assembly failure. For the same fuel assembly enrichment, the higher the burnup of the damaged fuel assembly, the greater the ratio of the specific activities of Cs-134 and Cs-137. Furthermore, for the same burnup, the higher the enrichment of the damaged fuel assembly, the smaller the activity ratio of 134Cs and 137Cs. Therefore, the enrichment and burnup level of the damaged fuel rods can be estimated by the ratio of the radioactive specific activities of Cs-134 and Cs-137 in the coolant, and the approximate location of the damaged fuel can be determined based on the core fuel loading plan.

[0045] The primary radiochemical analysis module calculates and analyzes damaged fuel rod damage information, including the damaged fuel rod damage index, the number of damaged fuel rods, the damaged fuel burnup range, the damaged fuel's core location, and the damaged fuel rod's rupture dimensions. The module uses the following calculation and analysis method: To determine the damaged fuel burnup range and core location, the module uses an empirical table of damaged fuel rods derived from fuel rod burnup testing. The damaged fuel rod burnup empirical table records the relationship between the damaged fuel rod burnup extent and the specific activity ratio of Cs-134 to Cs-137. The primary radiochemical analysis module determines the damaged fuel rod burnup extent by calculating the specific activity ratio of Cs-134 to Cs-137 from the primary radiochemical parameter input and querying the fuel damage burnup empirical table. The primary radiochemical analysis module matches the damaged fuel rod burnup with the fuel assembly burnup and fuel assembly enrichment information from the core burnup distribution input to determine the damaged fuel rod location and fuel rod enrichment.

[0046] The primary circuit radiochemical analysis module calculates the damaged fuel rod damage index using formula (1) and formula (2), as follows:

[0047] FRI p =[A N,I-131 -kA N,I-134 ]×[(L n / LHGR)×(100 / P r )] 1.5 (1)

[0048] Among them, FRI p is the fuel damage index, A N,I-131 and A N,I-134 They represent the specific activities of iodine-131 and iodine-134 input into the radiochemical parameters of the first loop; k is the adhesion correction factor; L n : Linear power density, which is the power output per unit length of the fuel rod, specifically 18.0kW / m; L HCR is the linear power density at high burnup rate, that is, the power output per unit length of the damaged fuel rod, specifically the linear power density at full power, in kW / m, which is calculated through the core parameter input; Pr is the core relative power, which is obtained from the unit information and operation data.

[0049]

[0050] Where Y is the fraction of the isotope in the fission product, that is, the proportion of a certain radioactive isotope produced during nuclear fission; λ is the decay constant, which is related to the half-life of the isotope and can be obtained by searching the literature; B n is the universal purification rate constant, which is 2x10 -5 s -1 ; k is the adhesion correction coefficient, k is a constant, and the calculated result is 0.0318. The primary iodine analysis module obtains the specific activity of iodine isotopes and calculates the number of damaged fuel rods using formulas (3) and (4), as follows:

[0051]

[0052] Where R is the fission product release rate, B is the fission product production rate; Cm is the measured radioisotope concentration, which is derived from the radioisotope specific activity in the input primary radiochemical parameters; Mc is the coolant mass, which can be calculated based on the core parameter input; λ is the radioisotope decay constant, which can be obtained by searching the literature; β p is the coolant purification rate constant, which can be obtained by searching relevant literature; F fis the fuel fission rate, that is, how many fissions occur per second, which can be obtained by searching relevant literature; Y is the cumulative fission fraction of the isotope, that is, how many atoms of the isotope are produced per fission, which can be obtained by searching relevant literature. x represents the number of damages when multiple damages occur; H is the geometric factor correction value, which can be obtained by searching relevant literature; v is the interstitial escape rate coefficient, which can be obtained by searching relevant literature; D ′ is the empirical diffusion coefficient, which can be obtained by searching relevant literature; c is a constant. Substituting the specific activities of I-131, I-132, I-133, I-134, and I-135 into the above formula (3) yields the corresponding calculated value R / B. Substituting R / B into formula (4) forms a system of equations, which can be solved to obtain the values of x, v, and c, where x represents the number of damages that occur in multiple damages.

[0053] In some embodiments, the primary radiochemical analysis module uses formula (5) to directly calculate the number of damaged fuel rods.

[0054]

[0055] Where: D′ I =10 9.857log P-25.1314 D′ NG =10 8.632log P-23.4091 A=x 2 D′, x represents the number of damages that occur in multiple cases, D′ I is the empirical diffusion coefficient of the iodine isotope in a single breakage, D′ NG is the empirical diffusion coefficient of the rare gas in a single damage situation; P is the line power, ranging from 25-60kW / m, obtained based on the actual situation of the unit.

[0056] The primary circuit iodine analysis module obtains the specific activity of iodine isotopes and calculates the rupture size of the damaged fuel rod using the empirical formula (6), as follows:

[0057]

[0058] Where E is the rupture size, vf is the iodine isotope interstitial escape rate coefficient, which is the experimental fitting value; l is the length of the fuel rod, which can be obtained by searching relevant literature; n is the damage location parameter, 1 indicates middle damage, and 0.5 indicates end damage. Both can be selected according to actual conditions.

[0059] In the above calculations, the staff can input the known parameters that need to be found in the literature or obtained through experiments into the primary radiochemical analysis module in advance.

[0060] The fuel performance analysis module inputs and stores unit information and operating data. Based on this input data, the system performs multi-physics coupled fuel performance calculations, determining changes in physical quantities such as stress, strain, and fission gas release. It then outputs two failure criteria: cumulative damage and total deformation. The results are then presented graphically. These two failure criteria effectively distinguish between mechanical and human factors causing fuel rod failure. The module constructs a coupled finite element model based on mechanical and heat transfer, dividing the fuel rods axially into axisymmetric and non-axisymmetric layers. Interlayer coupling is achieved using gas cavity parameters. Axisymmetric layers are modeled using one-dimensional finite elements, while non-axisymmetric layers are modeled using two-dimensional RZ finite elements. Interlayer coupling between axisymmetric and non-axisymmetric layers is achieved by Picard iterations based on gas cavity pressure, temperature, and fission gas release. Ultimately, changes in fuel pellet stress, strain, and fission gas release are calculated, and the two failure criteria, cumulative damage and total deformation, are output. For thermodynamic modeling, the total heat transfer coefficient in the interstitial heat transfer model is equal to the sum of the heat transfer coefficients of the gas layer, the solid contact portion, and the radiation heat transfer coefficient. The Ross & Stoute model is used to calculate the equivalent heat transfer coefficient for the gas heat transfer model. The solid contact heat transfer model uses an elastic contact model based on surface roughness. For radiation heat transfer, the Stefan-Boltzmann law is used to correct the model. For the fuel material model, the Fink-Lucuta model is used to calculate the thermal conductivity of uranium dioxide in the fuel rods, and this model is corrected for burnup, porosity, and radiation damage. The ESCORE model is used for the densification model, the dual-mechanism model of solid and gaseous fission products is used for the swelling model, and the MATPRO model, which includes thermal creep and irradiation creep, is used for the creep model. For the cladding material model, the Hoppe irradiation creep model is used for the Zr alloy model, and the linear hardening elastoplastic model is used for the SS316 austenitic stainless steel model, with a yield stress of 200 MPa and a hardening constant of 1600 MPa.

[0061] The damage cause analysis module obtains the unit information and operating data of the data acquisition module, the calculation results of the fuel performance analysis module and the fuel assembly inspection results provided by the maintenance personnel. The fuel assembly inspection results provided by the maintenance personnel include the damage shape, damage location, damage size, number of cycles experienced by the fuel assembly, fuel pellet structure, and pictures of damaged fuel rods.

[0062] The damage cause analysis module analyzes the causes of fuel rod failure. Possible causes include foreign matter, grid-to-fuel-rod vibration wear (GTRF), pellet-to-cladding interaction (PCI), hydrogenation, pellet surface loss, fouling corrosion, and end-plug weld defects. The module compiles a case library of confirmed fuel assembly failures worldwide. Based on this case library, the module summarizes the characteristics corresponding to each fuel rod failure cause and develops a method for determining the cause of fuel rod failure. The specific method is shown in Table 1. The module's case library is continuously updated with fuel rod failure incidents, and the content of the fuel rod failure cause determination method is also continuously updated with the case library. This allows the module to provide increasingly accurate analysis of fuel assembly failure causes.

[0063] Table 1 Methods for determining the causes of fuel rod damage

[0064]

[0065]

[0066] As shown in Table 1, if the damaged fuel rod meets the specific characteristics of the checked fuel assembly failure causes in the table above, the fuel assembly failure is classified as that cause. If multiple fuel failure causes are met, the failure causes are sorted from highest to lowest according to the current frequency in Table 1 and displayed to the user. EFPD in the table above is a unit of measurement for burnup depth; 1 EFPD represents one day of continuous reactor operation at 100% full power.

[0067] The present invention proposes a method for analyzing the causes of nuclear fuel assembly damage based on a knowledge graph, which includes:

[0068] Step 1: The data acquisition module acquires unit information and operating data, and transmits the unit information and operating data to the primary radiochemical analysis module, the fuel performance analysis module, and the damage cause analysis module. The specific information includes: the unit information and operating data input by the primary radiochemical analysis module include core parameter input, core burnup distribution input, primary radiochemical parameter input, and operating parameter input. The core parameter input specifically includes unit name, unit thermal power, primary coolant volume, fuel heat release rate, number of fuel rods, core active area height, and fuel pellet diameter; Figure 5As shown in the figure, the core burnup distribution input includes the burnup of each fuel assembly and the fuel assembly enrichment according to the fuel assembly core loading diagram; the primary loop radiochemical parameter input includes the specific activity of 12 nuclides, namely Cs-134, Cs-137, I-131, I-132, I-133, I-134, I-135, Xe-133, Xe-135, Kr-85, Kr-87 and Kr-88. The type of nuclides can be adjusted according to needs; the operating parameter input specifically includes the cycle name, input relative power, average moderator temperature, ion adsorption bed purification efficiency, xenon removal rate, number of core fuel enrichment zones, average burnup depth of each enrichment zone, and enrichment of each enrichment zone.

[0069] Step 2: The primary radiochemical analysis module screens and calculates and analyzes the unit information and operating data to determine whether fuel assembly damage has occurred inside the core and obtain damage information of the damaged fuel rods. The damage information includes the damaged fuel rod damage index, the number of damaged fuel rods, the damaged fuel burnup value range, the core location of the damaged fuel, and the size of the damaged fuel rod rupture.

[0070] Step 3: The fuel performance analysis module uses unit information and operating data to perform multi-physics coupling calculations on fuel performance, constructs a mechanical and heat transfer coupled finite element model, divides the fuel rods into axisymmetric layers and non-axisymmetric layers along the axial direction, and realizes interlayer coupling through the gas parameters of the air cavity. The changes in physical quantities such as stress, strain, and fission gas release of the damaged fuel rods are obtained, and the two failure criteria of cumulative damage and total deformation are output.

[0071] Step 4: Based on the damage information of the damaged fuel rod inferred by the primary radiochemical analysis module, conduct an on-site inspection of the damaged fuel assembly and collect the fuel assembly inspection results.

[0072] Step 5: The damage cause analysis module obtains the unit information and operating data from the data acquisition module, the calculation results of the fuel performance analysis module, and the fuel assembly inspection results provided by the maintenance personnel. The damage cause analysis module collects cases that have occurred in the world and the causes of fuel assembly damage that have been determined to build a case library. The damage cause analysis module summarizes the characteristics of each fuel rod damage cause based on the case library. The damage cause analysis module matches the obtained information with the characteristics of the fuel rod damage cause to determine the cause of the fuel assembly damage. If the information obtained by the damage cause analysis module matches multiple fuel damage causes, the multiple fuel damage causes are arranged in descending order of frequency and displayed to the user.

[0073] The above embodiments merely illustrate several implementations of the present invention, and while their descriptions are relatively specific and detailed, they should not be construed as limiting the scope of the present invention. It should be noted that a person skilled in the art would be able to make numerous modifications and improvements without departing from the spirit of the present invention, all of which fall within the scope of protection of the present invention. Therefore, the scope of protection of the present invention shall be determined by the appended claims.

Claims

1. A nuclear fuel assembly damage cause analysis system based on knowledge graph, characterized in that: The system includes a data acquisition module, a primary radiochemical analysis module, a fuel performance analysis module, and a damage cause analysis module. The data acquisition module acquires unit information and operating data and transmits the unit information and operating data to the primary radiochemical analysis module, the fuel performance analysis module, and the damage cause analysis module. The primary radiochemical analysis module screens and calculates and analyzes the unit information and operating data to determine whether a fuel assembly is damaged in the core and obtain damage information of the damaged fuel rod. The fuel performance analysis module performs multi-physics coupling calculations on fuel performance using unit information and operating data to obtain the physical quantity change trend and failure criteria of the damaged fuel rod. The damage cause analysis module obtains the unit information and operating data from the data acquisition module, the calculation results of the fuel performance analysis module, and the fuel assembly inspection results provided by maintenance personnel to analyze the cause of fuel rod damage.

2. A nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 1, characterized in that: The unit information and operating data acquired by the data acquisition module include: core parameter input, core burnup distribution input, primary circuit radiochemical parameter input and operating parameter input; the damage information of the damaged fuel rod includes the damaged fuel rod damage index, the number of damaged fuel rods, the damaged fuel burnup, the core position of the damaged fuel and the damaged fuel rod rupture size.

3. A nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 2, characterized in that: The core parameter input specifically includes the unit name, unit thermal power, primary coolant volume, fuel heat release rate, number of fuel rods, core active area height, and fuel pellet diameter; the core burnup distribution input includes the fuel assembly burnup and fuel assembly enrichment; the primary radiochemical parameter input includes the specific activities of 12 nuclides: Cs-134, Cs-137, I-131, I-132, I-133, I-134, I-135, Xe-133, Xe-135, Kr-85, Kr-87, and Kr-88. The types of nuclides can be adjusted as needed; the operating parameter input specifically includes the cycle name, input relative power, average moderator temperature, ion adsorption bed purification efficiency, xenon removal rate, number of core fuel enrichment zones, average burnup depth of each enrichment zone, and enrichment of each enrichment zone.

4. A nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 3, characterized in that: The primary radiochemical analysis module determines whether fuel assembly damage occurs inside the core based on the specific activities of Cs-134 and Cs-137 input as primary radiochemical parameters. If the specific activities of Cs-134 and Cs-137 are greater than 0, it is determined that fuel assembly damage occurs inside the core.

5. A nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 4, characterized in that: The primary radiochemical analysis module calculates the specific activity ratio of Cs-134 and Cs-137 input by the primary radiochemical parameter input, and searches the specific activity ratio in the fuel damage burnup empirical table to obtain the damaged fuel rod burnup; the primary radiochemical analysis module matches the damaged fuel rod burnup with the burnup of the fuel assembly and the fuel assembly enrichment in the core burnup distribution input to determine the damaged fuel rod position and fuel rod enrichment; the primary radiochemical analysis module calculates the damaged fuel rod damage index by formula (1) and formula (2), and the specific formula is as follows: Among them, FRI p is the fuel damage index, A N,I-131 and A N,I-134 They represent the specific activities of iodine-131 and iodine-134 input into the radiochemical parameters of the first loop; k is the adhesion correction factor; L n : Linear power density, which is the power output per unit length of the fuel rod; L HCR is the linear power density at high burnup rate, that is, the power output per unit length of damaged fuel rods; Pr is the core relative power, which is obtained from unit information and operating data; Where Y is the fraction of the isotope in the fission product, that is, the proportion of a certain radioactive isotope produced during nuclear fission; λ is the decay constant, which is related to the half-life of the isotope; Bn is the universal purification rate constant; and k is the adhesion correction factor. The primary circuit iodine analysis module obtains the specific activity of iodine isotopes and calculates the number of damaged fuel rods using formulas (3) and (4), as follows: Where R is the fission product release rate, B is the fission product production rate; Cm is the measured radioisotope concentration, which comes from the radioisotope specific activity in the input primary radiochemical parameters; Mc is the coolant mass; λ is the radioisotope decay constant; β p is the coolant purge rate constant; F f is the fuel fission rate, that is, how many fissions occur per second; Y is the cumulative fission fraction of the isotope, that is, how many atoms of this isotope are produced in each fission; x represents the number of damaged fuel rods; H is the geometric factor correction value, v is the interstitial escape rate coefficient; D′ is the empirical diffusion coefficient; c is a constant; substitute the specific activities of I-131, I-132, I-133, I-134, and I-135 into the above formula (3) respectively to obtain the corresponding calculated value R / B, substitute R / B into formula (4) to form a system of equations, and solve to obtain the value of x, where x represents the number of damaged fuel rods. The primary iodine analysis module obtains the specific activity of iodine isotopes and calculates the rupture size of the damaged fuel rod using the empirical formula (6), as follows: Where E is the rupture size; vf is the iodine isotope interstitial escape rate coefficient; l is the length of the fuel rod; and n is the damage position parameter, where 1 indicates damage in the middle and 0.5 indicates damage at the end. The value is selected based on the actual situation.

6. A nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 3, characterized in that: The primary circuit radiochemical analysis module calculates the number of damaged fuel rods using formula (5), which is as follows: Among them, D′ I =10 9.857log P-25.1314 D′ NG =10 8.632log P-23.4091 A=x 2 D′, x represents the number of damaged fuel rods, D′ I is the empirical diffusion coefficient of the iodine isotope in a single breakage, D′ NG is the empirical diffusion coefficient of the rare gas in a single damage; P is the linear power.

7. The nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 5 is characterized in that: The fuel damage burnup empirical table is obtained based on a fuel rod damage burnup test, and includes a relationship between the damaged fuel rod burnup and the specific activity ratio of Cs-134 to Cs-137.

8. The nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 3 is characterized in that: The fuel performance analysis module obtains the changing trends of stress, strain and fission gas release through multi-physics coupling calculation of fuel performance, and calculates two failure criteria of output cumulative damage and total deformation, and presents the changing trends of stress, strain and fission gas release, cumulative damage and total deformation to the user in the form of charts; the multi-physics coupling calculation of fuel performance specifically includes the fuel performance analysis module establishing a finite element model coupled with mechanics and heat transfer for fuel rods, dividing the fuel rods into axisymmetric layers and non-axisymmetric layers along the axial direction, and realizing interlayer coupling through air cavity gas parameters; the fuel performance analysis module inputs unit information and operation data into the finite element model, and obtains the changing trends of stress, strain and fission gas release, output cumulative damage and total deformation through simulation calculation.

9. The nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 3 is characterized in that: The damage cause analysis module collects cases that have occurred in the world and have confirmed the causes of fuel assembly damage to build a case library. The damage cause analysis module summarizes the characteristics of each fuel rod damage cause based on the case library. The damage cause analysis module matches the acquired information with the characteristics of the fuel rod damage cause to determine the cause of the fuel assembly damage. If the acquired information matches multiple fuel damage causes, the multiple fuel damage causes are arranged in descending order of frequency and displayed to the user.

10. A nuclear fuel assembly damage cause analysis system based on knowledge graph according to claim 9, characterized in that: The fuel assembly inspection results provided by the maintenance personnel include the damage shape, damage location, damage size, number of cycles experienced by the fuel assembly, fuel pellet structure and pictures of the damaged fuel rods.

11. A method for analyzing the cause of nuclear fuel assembly damage based on a knowledge graph according to any one of claims 1 to 10, characterized in that: The method comprises: Step 1: The data acquisition module obtains unit information and operating data, and transmits the unit information and operating data to the primary circuit radiochemical analysis module, fuel performance analysis module and damage cause analysis module; Step 2: The primary circuit radiochemical analysis module screens and performs calculations and analysis on the unit information and operation data to determine whether fuel assembly damage occurs inside the core and obtain damage information of the damaged fuel rods; Step 3: The fuel performance analysis module performs multi-physics coupling calculation of fuel performance based on the unit information and operation data to obtain the physical quantity change trend and failure criterion of the damaged fuel rod; Step 4: Based on the damage information of the damaged fuel rod inferred by the primary radiochemical analysis module, conduct an on-site inspection of the damaged fuel assembly and obtain the fuel assembly inspection results; Step 5: The damage cause analysis module obtains the unit information and operating data from the data acquisition module, the calculation results of the fuel performance analysis module, and the fuel assembly inspection results provided by the maintenance personnel. The damage cause analysis module collects cases that have occurred in the world and the causes of fuel assembly damage that have been determined to build a case library. The damage cause analysis module summarizes the characteristics of each fuel rod damage cause based on the case library. The damage cause analysis module matches the obtained information with the characteristics of the fuel rod damage cause to determine the cause of the fuel assembly damage.

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