Pressurized water reactor full-core fuel rod core block-cladding contact damage allowance online calculation method

Abstract

The invention discloses a pressurized water reactor full-core fuel rod core block-cladding contact damage margin online calculation method, which comprises the following steps of: 1, establishing a geometric and physical model of a reactor core based on a pressurized water reactor core overall parameter, and obtaining three-dimensional distribution of linear power density of a reactor core fuel rod online by using reactor core physical calculation software; 2, analyzing the full-core fuel rod by using fuel performance analysis software in combination with geometric and material parameters of the fuel rod, and calculating to obtain full-core cladding stress and strain three-dimensional distribution; and thirdly, introducing strain energy density as a pellet-cladding contact damage criterion, and comparing the calculated strain energy density with critical strain energy to obtain a pellet-cladding contact damage margin so as to realize online evaluation of the safety state of the cladding. According to the method disclosed by the invention, the strain energy density-based on-line refined evaluation of the full-core fuel rod pellet-cladding contact damage allowance is realized, and a more comprehensive, on-line and reliable technical means is provided for the operation safety of a pressurized water reactor.

Description

Technical Field

[0001] This invention relates to the field of commercial pressurized water reactor core safety analysis technology, specifically to an online calculation method for contact damage margin of fuel rod pellets and cladding in a pressurized water reactor core. Background Technology

[0002] As the first barrier containing radioactive materials in a nuclear reactor, the structural integrity of the fuel rod cladding is crucial for the safe and stable operation of the reactor. Pellet cladding interaction (PCI) is one of the important factors leading to cladding failure. The occurrence of PCI is closely related to factors such as the operating history of the fuel rods, burnup depth, and power change rate.

[0003] In existing technologies, to mitigate the risk of cladding failure due to PCI (Potential Processing Injection), operational measures such as limiting reactor power levels, controlling fuel rod burn-up depth, and limiting power ramp-up rates are commonly employed. However, these control measures, to some extent, limit the flexibility of reactor operation and reduce the unit's economic efficiency. As nuclear power generation continues to account for an increasing proportion of my country's power structure, nuclear power units need to participate more frequently and deeply in grid peak shaving. Under such operating conditions, reactor power fluctuates frequently, and cladding stress and strain conditions become complex, significantly increasing the risk of cladding failure due to PCI.

[0004] Furthermore, some nuclear power units, after prolonged low-power operation, need to quickly return to full-power operation to meet grid demands. During the rapid power increase phase, the cladding experiences a combination of thermal stress and chemical effects, significantly increasing its failure probability. Therefore, in-depth research into the PCI mechanism and scientific assessment methods for cladding failure risk are of significant engineering importance.

[0005] Current fuel performance analysis mainly relies on experimental data and empirical models, which makes it difficult to accurately describe the three-dimensional distribution of stress and strain in the fuel rod cladding. Especially under complex operating conditions such as peak shaving or rapid power changes, there is a lack of analytical methods that can finely assess the contact damage margin between the pellet and the cladding. Existing methods usually use stress limits as the criterion, which fails to fully reflect the energy accumulation and mechanical response process of the cladding, thus having certain limitations in terms of applicability and assessment accuracy. Summary of the Invention

[0006] To address the problems existing in the prior art, this invention proposes an online calculation method for the contact damage margin of the entire reactor core fuel rod pellets and cladding. Based on the actual operating history of nuclear power plants, by coupling core physics analysis software and single-rod fuel performance analysis software, the three-dimensional circumferential stress and strain state of the entire reactor core fuel rod cladding can be described, and then the strain energy density can be used to assess the margin for cladding failure.

[0007] To achieve the above objectives, the technical solution adopted by the present invention is as follows: A method for online calculation of contact damage margin between fuel rod pellets and cladding in a pressurized water reactor core includes the following steps: Step 1: Based on the overall parameters of the pressurized water reactor core, establish the geometric and physical model of the core, and use core physics analysis software to update the core power distribution in real time to obtain the three-dimensional distribution of the linear power density of the fuel rods in the entire core. Step 2: Combining the geometry and material parameters of the fuel rods, perform fuel performance analysis on each fuel rod in the entire core area, and calculate the temperature distribution, circumferential stress, circumferential strain, air gap thickness and oxide layer thickness parameters of the fuel rods; Step 3: Using strain energy density as the core-cladding contact damage criterion, calculate the strain energy density of the cladding based on the circumferential stress and strain of the cladding, and compare it with the critical strain energy density determined by experiments to obtain the core-cladding contact damage margin.

[0008] Step 1 involves the following steps: Based on the overall parameters of the nuclear power plant core, such as the fuel loading scheme, control rod positions, core power, and coolant flow distribution, a geometric and physical model of the pressurized water reactor core is established. Then, using core physics analysis software, steady-state neutronics calculations are performed based on nuclear power plant operating parameters such as power level, rod position information, and burnup increment. Through fine rod power reconstruction, the three-dimensional distribution of the online full-core fuel rod linear power density is obtained.

[0009] Step 2 is as follows: Based on the three-dimensional distribution of linear power density of the fuel rods in the entire core, and combined with geometric and material parameters such as the outer diameter of the fuel rod pellets, the inner diameter of the cladding, the pellet enrichment, and the cladding material, the single-rod fuel performance analysis software is called online concurrently to perform fuel performance analysis on each fuel rod within the entire core range, and to obtain key parameters such as fuel rod temperature distribution, circumferential stress, circumferential strain, air gap thickness, and oxide layer thickness.

[0010] Step 3 involves further analyzing the three-dimensional distribution of circumferential stress and strain in the fuel rod cladding of the entire reactor core, and calculating the strain energy density as a criterion for contact damage between the fuel rod and the cladding. Formula (1) In the formula: Represents strain energy density; Indicates circumferential stress; Indicates circumferential strain; The obtained strain energy density value is compared with the critical strain energy density to calculate the core-cladding contact damage margin: Formula (2) In the formula: Indicates the chip-cladding contact damage margin; Indicates the critical strain energy density; Represents strain energy density; when When >0, it indicates that the casing is in a safe state; when If the value is ≤0, it indicates that the casing is at risk of failure.

[0011] Preferably, the core physics analysis software described in step 1 is Bamboo-C, an advanced pressurized water reactor (PWR) core physics analysis software. Bamboo-C software is based on the classic two-step PWR calculation process and employs advanced nuclear reactor physics calculation methods, making it applicable to different types of commercial PWRs. Bamboo-C software mainly includes three functional programs: the component program LOCUST, the core program SPARK, and the functionalization program LtoS. Among them, the core program SPARK can obtain the three-dimensional distribution of the fuel rod line power density across the entire core through a fine rod power reconstruction method.

[0012] Preferably, the single-bar fuel performance analysis software mentioned in step 2 is the FRAPCON fuel performance analysis software program. In this invention, to meet the computational efficiency requirements of full-core fuel performance analysis, some improvements are made to the coupling method and output module when the FRAPCON program is coupled with the Bamboo-C core physics analysis software.

[0013] Taking the M310 pressurized water reactor at Tianwan Nuclear Power Plant as an example, the reactor core contains 157 fuel assemblies, each with 264 fuel rods. The performance data for the entire reactor core needs to be analyzed for 41,448 fuel rods. Due to the enormous computational workload, parallel computing is employed to improve computational efficiency when performing single-rod fuel performance analysis. This allows the analysis of multiple fuel rods to proceed simultaneously, significantly shortening the overall computation time for the whole-core fuel rod pellet-cladding contact damage margin analysis.

[0014] Furthermore, the single-bar fuel performance analysis software generates numerous independent output files after execution, and reading these files one by one results in low data processing efficiency. To address this issue, the FRAPCON fuel performance analysis software was functionally improved, enabling it to generate additional structured data files containing key parameters such as fuel rod temperature, circumferential stress, circumferential strain, and air gap thickness when outputting calculation results. The results generated by each calculation unit are automatically integrated into a single full-core fuel performance result file within the system, thereby reducing the number of data read / write operations and improving the reading efficiency and overall analysis speed of subsequent core physics calculation software.

[0015] Compared with the prior art, the present invention has the following significant advantages: 1. By using coupled core physics analysis software and single-bar fuel performance analysis software, the three-dimensional distribution of circumferential stress and strain in the fuel rod cladding of the entire core can be described; 2. It can analyze different types of commercial pressurized water reactors online. By analyzing the real-time operating power history of the pressurized water reactor, it can obtain the online core-cladding contact damage margin and reflect the core operating safety status in a timely manner. 3. The method of the present invention uses strain energy density to assess the cladding safety status. Compared with single stress index, strain energy density takes into account information such as stress, strain and energy state, providing a more comprehensive, online and reliable technical means for pressurized water reactor operation safety. Attached Figure Description

[0016] Figure 1 This is a flowchart of the method of the present invention.

[0017] Figure 2 This is a three-dimensional distribution diagram of the circumferential stress in the cladding at the end of the reactor core's life.

[0018] Figure 3 This is a three-dimensional distribution diagram of strain energy density at the beginning of the reactor core's lifespan.

[0019] Figure 4 This is a three-dimensional distribution diagram of strain energy density during the core's lifetime.

[0020] Figure 5 This is a three-dimensional distribution diagram of strain energy density at the end of the reactor core's lifespan. Detailed Implementation

[0021] To make the technical solution of the present invention clearer, the present invention will be further described below in conjunction with the accompanying drawings and specific embodiments.

[0022] This invention proposes an online calculation method for the contact damage margin of the fuel rod pellets and cladding in a pressurized water reactor core. The specific implementation steps are as follows: Figure 1 As shown, it includes the following steps: Step 1: Based on the overall core parameters of the nuclear power plant, such as the fuel loading scheme, control rod positions, core power, and coolant flow distribution, a geometric and physical model of the pressurized water reactor (PWR) core is established using the advanced PWR core physics analysis software Bamboo-C. Then, steady-state neutronics calculations are performed using Bamboo-C based on nuclear power plant operating parameters such as power level, rod position information, and burnup increment. Bamboo-C features a fine-grained rod power reconstruction function, enabling online acquisition of the three-dimensional distribution of linear power density of the entire core fuel rods, providing input parameters for subsequent rod fuel performance analysis. Step 2: Based on the three-dimensional distribution of linear power density of the fuel rods in the entire core, and combined with geometric and material parameters such as the outer diameter of the fuel rod pellets, the inner diameter of the cladding, the pellet enrichment, and the cladding material, the single-rod fuel performance analysis software FRAPCON is called online concurrently to perform fuel performance analysis on each fuel rod in the entire core, and key parameters such as fuel rod temperature distribution, circumferential stress, circumferential strain, air gap thickness, and oxide layer thickness are obtained.

[0023] Step 3: Based on the fuel performance analysis results, further analysis is conducted on the circumferential stress and circumferential strain at various locations within the fuel rod cladding of the entire reactor core, and the corresponding strain energy density is calculated: Formula (1) In the formula: This represents the cladding strain energy density; This indicates the circumferential stress of the casing; Indicates the circumferential strain of the shell; Calculated strain energy density Compared with the experimentally determined critical strain energy density Comparison of chip-cladding contact damage margin Defined as: Formula (2) In the formula: Indicates the chip-cladding contact damage margin; Indicates the critical strain energy density; Represents strain energy density; when When >0, it indicates that the casing is in a safe state; when If the value is ≤0, it indicates that the casing at that location is at risk of failure.

[0024] By summarizing the calculation results of all fuel rods in the reactor core, the three-dimensional distribution of strain energy density and the corresponding core-cladding contact damage margin distribution can be obtained. These results can be used to identify the fuel assemblies and axial locations with the highest cladding failure risk, providing a basis for operational optimization.

[0025] This example uses the Tianwan Nuclear Power Plant M310 pressurized water reactor as an example, taking its actual operational burnup depth as input to perform full-lifecycle burnup tracking calculations. In this example, the impact of power variations during operation on strain energy density is not considered; it is assumed that the reactor operates at full power throughout its entire lifecycle to verify the feasibility of the online calculation method for the contact damage margin of the entire reactor core fuel rod pellet-cladding as described in this invention.

[0026] in Figure 2 The figure shows the three-dimensional distribution of circumferential stress in the fuel rods of the entire core at the end of its service life. It can be seen from the figure that even without considering the influence of power changes on the core, there are significant differences in stress among the fuel rods of the entire core at the end of its service life. The maximum stress reaches -68.27 MPa, which indicates that the cladding is mainly subjected to external coolant pressure at this time. Figure 3 , Figure 4 and Figure 5 The three-dimensional distribution maps of strain energy density (SED) at the beginning, middle, and end of the entire core's lifespan are shown. The results indicate that the strain energy density gradually increases with burnup, and the differences between different fuel rods also gradually increase. The maximum strain energy density at the end of the lifespan is 0.37 MPa, which, compared to the critical strain energy density of 2.87 MPa, still provides a 2.50 MPa margin for pellet-cladding contact damage, indicating that the cladding still has sufficient safety margin.

Claims

1. A method for on-line calculation of the core barrel cladding contact damage allowance of a pressurized water reactor full core fuel rod, characterized in that: It comprises the following steps: Step 1: based on the overall parameters of the pressurized water reactor core, the geometric and physical model of the core is established, the core power distribution is updated in real time by using the core physical analysis software, and the three-dimensional distribution of the line power density of the fuel rod in the whole core is obtained; Step 2: combined with the geometric structure and material parameters of the fuel rod, the fuel performance of each fuel rod in the whole core is analyzed, and the temperature distribution, circumferential stress, circumferential strain, air gap thickness and oxidation layer thickness parameters of the fuel rod are calculated; Step 3: taking the strain energy density as the core-pellet contact damage criterion, the strain energy density of the cladding is calculated based on the circumferential stress and circumferential strain of the cladding, and compared with the critical strain energy density determined by experiment, to obtain the core-pellet contact damage margin.

2. The method of claim 1, wherein the method is characterized by: The specific process of step 1 is: according to the fuel loading scheme, the control rod position, the core power and the coolant flow distribution, the geometric and physical model of the pressurized water reactor core is established, and then the steady-state neutron calculation is carried out by using the core physical analysis software through the power level, rod position information and fuel burnup increment, and the online three-dimensional distribution of the line power density of the fuel rod in the whole core is obtained through fine rod power reconstruction.

3. The method of claim 1, wherein the method is characterized by: The specific process of step 2 is: based on the three-dimensional distribution of the line power density of the fuel rod in the whole core, combined with the geometric and material parameters such as the outer diameter of the fuel rod pellet, the inner diameter of the cladding, the pellet enrichment and the cladding material, the single-rod fuel performance analysis software is called online and concurrently to analyze the fuel performance of each fuel rod in the whole core, and the key parameters such as the temperature distribution, circumferential stress, circumferential strain, air gap thickness and oxidation layer thickness of the fuel rod are obtained.

4. The method of claim 3, wherein the method further comprises: determining the contact damage margin of the fuel rod pellet-clad contact in the PWR. The single-rod fuel performance analysis software is the FRAPCON program, and when the FRAPCON program is coupled with the core physical analysis software Bamboo-C, based on the calculation efficiency requirement, the FRAPCON program output module is optimized to additionally generate a structured data file containing the key parameters such as the temperature, circumferential stress, circumferential strain, air gap thickness and oxidation layer thickness of the fuel rod, and automatically integrate into a single whole-core result file after calculation is completed, so as to improve the data reading and processing efficiency.

5. The method of claim 1, wherein the method is characterized by: The specific process of step 3 is: for the three-dimensional distribution of the circumferential stress and strain of the fuel rod cladding in the whole core, the core-pellet contact damage criterion strain energy density is calculated: Equation (1) In the formula: wherein represents the strain energy density; represents the circumferential stress; represents the circumferential strain; The strain energy density value obtained is compared with the critical strain energy density, and the core-pellet contact damage margin is calculated: Equation (2) In the formula: represents the pellet-cladding contact damage allowance; represents the critical strain energy density; represents the strain energy density; When > 0, it indicates that the cladding is in a safe state; when ≤ 0, it indicates that the cladding is at risk of failure.

6. The method of claim 1, wherein the method is characterized by: The core physical analysis software in step 1 is the advanced pressurized water reactor core physical analysis software Bamboo-C.

7. The method of claim 1, wherein the method is characterized by: In the whole-core fuel performance calculation process of step 2, in order to improve the calculation efficiency, parallel computing is adopted to make multiple fuel rod analysis processes proceed at the same time, so as to significantly shorten the overall calculation time of the whole-core fuel performance analysis.

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