Method for rapidly identifying cracking damage of plasma facing part interface in Tokamak

Abstract

The invention discloses a method for rapidly identifying cracking damage of a plasma facing part interface in tokamak. According to the method, surface temperature field distribution and evolution data of a plasma-oriented component are obtained in real time through a high-resolution thermal infrared imager, and temperature evolution characteristic parameters (such as a heating rate, a peak temperature deviation and a cooling rate) are used for contrastive analysis; and the early diagnosis of the cracking damage of the dissimilar material interface welding area of the plasma component during the operation period of the tokamak is realized. According to the method, the defect that the damage of the part cannot be detected on line in real time in a traditional detection technology is overcome, the method for rapidly and conveniently completing damage response by utilizing temperature data of the high-resolution thermal infrared imager is provided, and the method is particularly suitable for evaluating the interface integrity of the dissimilar material of the plasma part in fusion Tokamak devices such as EAST and ITER.

Description

Technical Field

[0001] The present invention relates to the field of nuclear fusion research, and in particular to a method for rapidly identifying crack damage on the interface of a plasma-facing component in a tokamak. Background Art

[0002] In magnetic confinement fusion devices such as tokamaks, plasma-facing components (PFCs) are subject to significant heat loads, radiation, and particle bombardment. Consequently, damage to these components, such as cracking and surface peeling, can directly impact the operational stability and safety of the device. Due to the high temperatures and intense radiation exposure of plasma, traditional damage detection methods have significant limitations, particularly during experiments, where they cannot perform comprehensive, real-time damage identification. Existing damage detection techniques, such as ultrasonic testing, X-ray imaging, and scanning electron microscopy (SEM), mostly require post-experimental testing and are unable to detect potential damage or cracks in a timely manner. This non-real-time detection approach not only hinders timely repair and maintenance but can also cause damage to progress undetected, potentially leading to equipment failure or long-term damage.

[0003] Infrared thermal imaging is a non-contact, real-time temperature monitoring method. It has been widely used in various fields to detect surface damage in equipment. High-resolution infrared thermal imaging cameras can capture the temperature distribution and evolution of component surfaces in real time. In tokamaks, existing research typically focuses on monitoring the maximum temperature of PFC components to provide early warning of melting due to high heat flux. This monitoring method is primarily used to ensure the safety of PFC surfaces. Major facilities today, such as EAST, WEST, JET, and ITER, all utilize dissimilar material welding in the design of heat-resistant components. Tungsten metal, due to its high-temperature resistance, is currently the primary heat-carrying material in tokamaks. Copper or stainless steel, due to their excellent thermal conductivity, facilitates water-cooling of heat. Fusion scientists have demonstrated that tungsten-copper composite components possess excellent heat-carrying and heat-removal capabilities. However, as the weld zone is one of the primary areas of thermal damage in tungsten-copper components, a suitable method is needed to rapidly and non-destructively assess the normal heat transfer between component interfaces during tokamak operation and identify abnormally cracked components. Summary of the Invention

[0004] To address the above technical issues, the present invention provides a method for rapidly identifying interfacial cracking damage in plasma-facing components in tokamaks. This method is particularly applicable to plasma-facing components (PFCs) in magnetic confinement fusion devices such as tokamaks, where dissimilar materials are connected to the plasma. This method monitors the temperature distribution and evolution of the PFC surface in real time during plasma discharge. By analyzing the temperature evolution characteristics, it can promptly detect interfacial cracking caused by thermal fatigue, thereby enabling early identification of damage.

[0005] The specific technical solution is: a method for rapidly identifying crack damage on the interface of plasma-facing components in a tokamak, which specifically includes the following steps:

[0006] Step 1: During the plasma discharge process, the surface temperature field of the PFCs in the tokamak device is continuously scanned and monitored using an infrared thermal imager.

[0007] Step 2: Classify the PFCs and obtain the theoretical temperature evolution curves of PFCs in different regions during the plasma discharge process;

[0008] Step 3: Compare all modules of the same type and with the same theoretical surface temperature field in the area with the obtained temperature field curve. During the plasma discharge component surface temperature loading phase, if the PFCs temperature rise rate exceeds the temperature rise rate threshold of the theoretical curve, record the curve as a first-level temperature warning response. Otherwise, the component is considered to be in a healthy state.

[0009] Step 4: Select the component that received the first-level temperature warning in step 3. When the plasma discharge reaches the steady-state loading stage, if the steady-state temperature value appears multiple times and is continuously higher than the measured temperature value of healthy components at other locations around the ring, and exceeds the normal operating data fluctuation range threshold, record the curve and mark it as a second-level temperature warning response; otherwise, the component is considered to be in a healthy state.

[0010] Step 5: For the components that received the second-level temperature warning in step 4, if the time required to cool to room temperature is longer than the fixed threshold time of healthy components under the same working conditions during the cooling stage after the plasma discharge, it indicates that a one-way heat conduction obstruction problem has occurred. The components are identified and determined to have damage such as interface debonding or cracking, and are determined to have a third-level temperature anomaly response.

[0011] The present invention has the following beneficial effects:

[0012] Compared with traditional detection methods, the present invention has the advantages of being non-contact, highly efficient and real-time. It can directly identify damage during the experiment, avoid possible delays and errors in traditional methods, and ensure the stable operation of the equipment.

[0013] Specifically, the present invention uses a high-resolution infrared thermal imager installed in a tokamak to collect real-time data on the surface temperature distribution and evolution of PFCs. This high-resolution imaging technique allows for the timely detection of temperature anomalies, particularly those with rapid temperature increases, higher peak surface temperatures, and slower cooling. Further screening and analysis of these anomalies allows for the accurate identification of potential damage sites, providing a reference for subsequent maintenance and adjustments. This method is applicable to a variety of tokamaks, including but not limited to EAST and ITER, and demonstrates strong versatility and applicability. BRIEF DESCRIPTION OF THE DRAWINGS

[0014] Figure 1 A schematic flow chart of the method for rapidly identifying interface cracking damage on plasma-facing components in a tokamak according to the present invention;

[0015] Figure 2 This is a comparison diagram of the surface temperature evolution curves of interface damaged components and normal heat transfer components over time;

[0016] Figure 3 This is a comparison chart of the temperature evolution over time of ordinary components of the lower divertor during the plasma experiment of the EAST device, the theoretical curve obtained by simulation, and the interface damaged component. DETAILED DESCRIPTION

[0017] In order to make the objectives, technical solutions, and advantages of the present invention more clearly understood, the present invention is further described in detail below with reference to the accompanying drawings and examples. It should be understood that the specific embodiments described herein are intended only to illustrate the present invention and are not intended to limit the present invention. In addition, the technical features involved in the various embodiments of the present invention described below may be combined with each other as long as they do not conflict with each other. To achieve the above-mentioned objectives, the present invention adopts the following technical solutions.

[0018] In order to better illustrate the method of real-time identification of interface crack damage of plasma-facing components in a tokamak based on temperature measurement with a high-resolution infrared thermal imager, the present invention is described in detail in conjunction with the accompanying drawings and embodiments in the EAST device.

[0019] like Figure 1 As shown, the present invention provides a method for quickly identifying crack damage on the interface of plasma-facing components in a tokamak, comprising the following steps:

[0020] Step 1: During the EAST plasma discharge process, a high-resolution infrared thermal imager (Telops, time resolution >100 Hz, spatial resolution 3 mm / pixel) was used to continuously scan and monitor the surface temperature field of all plasma-facing components (PFCs) in the tokamak device. Based on their location, the PFCs were divided into upper divertor, lower divertor, main limiter, and protection limiter. Using the surface temperature evolution of the plasma components as a benchmark, the theoretical module temperature evolution curves for different regions were obtained over time. Taking the EAST lower divertor as an example, at the 115371th shot discharge (pulse #115371), the surface temperature of most divertor components was Figure 3 The blue line in the middle remains close. Combining ANSYS finite element simulation software, a divertor model identical to that in the device is established, and the same boundary conditions (water cooling parameters, thermal loading conditions, etc.) are added to simulate the theoretical temperature-time evolution curve, as shown in Figure 3 The red line in the middle shows the health criteria of the components. Figure 2 As shown in the figure, it can be judged from the temperature evolution curve during the entire thermal loading process that compared with healthy components, components with interface damage tend to have faster temperature rise, higher maximum temperature, and longer cooling time.

[0021] Step 2: By comparing the temperature evolution curves obtained by infrared temperature measurement and simulation, all the same type of components along the annular direction in the device area are screened. Figure 3 For example, the black line in the middle shows the surface temperature evolution curve of the abnormal module during the discharge of pulse #115371, as recorded by an infrared thermal imager. Comparing the real-time infrared temperature evolution data reveals significant discrepancies between the temperature evolution of this abnormal module and the simulated theoretical results. At the beginning of the experiment, during the temperature rise period when the device injected heating power, the abnormal module experienced a temperature rise faster than the theoretical simulation results, approximately 1.5 times the temperature rise of the simulated results, with dT1 / dt1 ≈1.5 dT2 / dt2. This represents a first-level temperature warning response, and all modules exhibiting similar phenomena in the circular direction are screened and recorded.

[0022] Step 3: Continue to observe the modules that have temperature warning level 1 response in the circular direction. Figure 3 As shown by the black curve in the middle, at the end of the plasma discharge, the abnormal module reaches a maximum temperature of around 1250°C, while the theoretical simulation result is only 852°C, a difference of nearly 400°C. This initiates the secondary temperature warning response, screening and recording all modules in the ring that exhibit similar phenomena.

[0023] Step 4: Continue to observe the modules with the secondary temperature warning response in the circular direction. Figure 3As shown in the figure, during the cooling phase after the discharge, the abnormal module requires more time to cool to the initial temperature (about 100°C) compared to the simulation results. With a delay of about 2.2 seconds, it is possible to quickly and conveniently determine that the module has interface cracking damage.

[0024] Step 5: Repeat steps 2-4, screening all modules in the loop for damaged components and recording them. Provide timely feedback to the Tokamak Engineering Team, and open the vacuum chamber for component maintenance if necessary.

[0025] In step 1, PFCs generally refer to the plasma-facing components in a tokamak device with active water cooling.

[0026] The components selected for obtaining the theoretical temperature evolution in step 2 and the components of the same type described in steps 3-5 should be consistent and must meet the following common requirements: (a) the material selection and geometric structure of the components are completely consistent; (b) the water cooling parameters inside the water pipes of the components are consistent; (c) the position of the components in the tokamak differs only in the circumferential direction; and (d) non-destructive testing before this discharge confirms that the interface is in good condition before the experiment begins and there are no obvious macroscopic defects.

[0027] The method of the present invention is applicable to tokamak devices, including but not limited to EAST, ITER and other magnetic confinement fusion experimental devices.

[0028] The present invention is used to quickly identify interface damage phenomena of tokamak internal components, which are targeted at components made of dissimilar welding materials with active water cooling. The region names include but are not limited to divertor and limiter; the structural types include but are not limited to through-tube components and flat plate types.

[0029] The results of the finite element simulation of the present invention are only used to calibrate the temperature measurement results of the health module, and are used to alleviate the influence of errors caused by signal disturbances, hardware jitter of the acquisition equipment, light radiation, etc. Slight errors are allowed in the simulation results.

[0030] The method of the present invention has strong real-time and high efficiency, and can timely detect and identify component damage during the experiment of magnetic confinement fusion devices such as tokamaks, providing effective support for equipment maintenance and safety assurance.

Claims

1. A method for rapidly identifying crack damage on the interface of plasma-facing components in a tokamak, characterized in that: The specific steps include: Step 1: During the plasma discharge process, the surface temperature field of the plasma-facing components PFCs in the tokamak device is continuously scanned and monitored using an infrared thermal imager. Step 2: Classify the PFCs and obtain the theoretical temperature evolution curves of PFCs in different regions during the plasma discharge process; Step 3: Compare all modules of the same type and with the same theoretical surface temperature field in the area with the obtained temperature field curve. During the plasma discharge component surface temperature loading phase, if the PFCs temperature rise rate exceeds the temperature rise rate threshold of the theoretical curve, record the curve as a first-level temperature warning response. Otherwise, the component is considered to be in a healthy state. Step 4: Select the component that received the first-level temperature warning in step 3. When the plasma discharge reaches the steady-state loading stage, if the steady-state temperature value appears multiple times and is continuously higher than the measured temperature value of healthy components at other locations around the ring, and exceeds the normal operating data fluctuation range threshold, record the curve and mark it as a second-level temperature warning response; Otherwise, the component is considered to be in a healthy state; Step 5: For the components that received the second-level temperature warning in step 4, if the time required to cool to room temperature is longer than the fixed threshold time of healthy components under the same working conditions during the cooling stage after the plasma discharge, it indicates that a one-way heat conduction obstruction problem has occurred. The components are identified and determined to have damage such as interface debonding or cracking, and are determined to have a third-level temperature anomaly response.

2. The method according to claim 1, characterized in that The temperature rise rate threshold is when the temperature rise rate is greater than 20% of the theoretical curve, that is, .

3. The method according to claim 1, characterized in that The normal operation data fluctuation range threshold is more than 15% beyond the normal operation data fluctuation range, that is, .

4. The method according to claim 1, wherein The fixed threshold time is 2s and above, that is, , Cooling time for abnormal parts, Cooldown time for health components.

5. The method according to claim 1, characterized in that In step 1, PFCs refer to the plasma-facing components in a tokamak device with active water cooling.

6. The method according to claim 1, characterized in that The areas include but are not limited to divertors and restrictors; the same types include but are not limited to tube-type components and flat-plate types.

7. The method according to claim 1, characterized in that The components selected when obtaining the theoretical temperature evolution in step 2 and the components of the same type described in steps 3-5 should be consistent and meet the following common requirements: (a) the material selection and geometric structure of the components are completely consistent; (b) the water cooling parameters inside the water pipes of the components are consistent; (c) the position of the components in the tokamak only differs in the circumferential direction; and (d) non-destructive testing before this discharge confirms that the interface is in good condition before the experiment begins and there are no obvious macroscopic defects.

8. The method according to claim 1, characterized in that The method is applicable to tokamak devices, including but not limited to EAST, ITER and other magnetic confinement fusion experimental devices.

9. The method according to claim 1, characterized in that The thermal imager has a temporal resolution of >100 Hz and a spatial resolution of 3 mm / pixel.

10. The method according to claim 1, characterized in that The theoretical temperature-time evolution curve is obtained by combining ANSYS finite element simulation software, establishing a completely identical PFCs model, adding the same boundary conditions, including water cooling parameters and thermal loading conditions, and simulating to obtain the theoretical temperature-time evolution curve.

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