Method for on-line monitoring of pressurized water reactor fuel cladding breakage and monitoring system thereof

Abstract

The present application relates to a pressurized water reactor fuel cladding breakage online monitoring method and a monitoring system thereof, the method comprising the following steps: step S1: obtaining a pre-established energy calibration relationship, which is a corresponding relationship between gamma ray energy and channel number; step S2: obtaining pre-calibrated efficiency curve parameters, which are obtained through active efficiency calibration based on a simulated tubular standard radioactive source, the material, outer diameter and inner diameter of the simulated tubular standard radioactive source being consistent with those of the measured pipeline; step S3: obtaining gamma energy spectrum data through a back-to-back coincidence measurement technology, and automatically deconvoluting the gamma energy spectrum data according to the energy calibration relationship and the efficiency curve parameters to calculate the type and activity concentration of fissile nuclides in the measured pipeline; and step S4: determining a corresponding preset threshold value according to the type of fissile nuclides, and comparing the activity concentration with the preset threshold value to obtain a monitoring result of pressurized water reactor fuel cladding breakage.

Description

Technical Field

[0001] This invention relates to nuclear reactor radioactivity monitoring technology, and more specifically, to a method and system for online monitoring of fuel cladding damage in pressurized water reactors. Background Technology

[0002] The fuel cladding is the first barrier protecting radioactive materials in a nuclear power plant. When it ruptures, fissile nuclides are released into the primary coolant, causing an abnormal increase in their radioactivity levels. The main fissile nuclides include Kr-85m (0.151 MeV), Kr-87 (0.403 MeV), Kr-88 (0.196 MeV, 0.835 MeV), Xe-133 (0.081 MeV), Xe-135 (0.25 MeV), I-131 (0.364 MeV), I-133 (0.53 MeV, 0.875 MeV, 1.3 MeV), Cs-134 (0.605 MeV, 0.796 MeV), and Cs-137 (0.662 MeV). Monitoring fuel cladding damage is a very important task, allowing nuclear power plants to understand the integrity of the fuel cladding. This is generally achieved by measuring the radiochemical indicators of the primary coolant water.

[0003] Currently, the main methods for monitoring fuel cladding damage in domestic nuclear power plants are regular manual sampling and radiochemical analysis to measure total gamma and gamma spectrum (twice a week), online leak tests, and offline leak tests. None of these methods can monitor and respond to fuel damage in real time. Once cladding damage occurs, there is no rapid response, which will increase the radiation dose to personnel during operation and the risk of accidental radiation exposure. In severe cases, it may even lead to unplanned unit shutdown.

[0004] Online automatic monitoring of fuel cladding damage in pressurized water reactors (PWRs) would effectively alleviate this problem. Different PWR types require different online monitoring technologies and solutions. Currently, most nuclear power units in China are PWRs. Given the characteristics of the primary coolant piping in PWRs, it is necessary to develop an online monitoring method that can be directly installed around the piping, possessing high sensitivity and low background interference, to achieve real-time detection and analysis of the characteristic gamma rays of fission nuclides. Summary of the Invention

[0005] The technical problem to be solved by the present invention is to provide an online monitoring method for fuel cladding damage in pressurized water reactors, addressing the aforementioned technical deficiencies in the existing technology. The method includes the following steps: Step S1: Obtain a pre-established energy scale relationship, which is the correspondence between gamma ray energy and channel number; Step S2: Obtain the pre-calibrated efficiency curve parameters, which are obtained by active efficiency calibration based on a simulated tubular standard radioactive source. The simulated tubular standard radioactive source has the same material, outer diameter, and inner diameter as the pipe being tested. Step S3: Obtain gamma spectrum data obtained through anti-coincidence measurement technology, and automatically interpret the gamma spectrum data according to the energy scale relationship and the efficiency curve parameters to calculate the type and activity concentration of fission nuclides in the tested pipeline. Step S4: Determine the corresponding preset threshold according to the type of fission nuclide, and compare the activity concentration with the preset threshold to obtain the monitoring results of pressurized water reactor fuel cladding damage.

[0006] Further, step S1 includes: The detector assembly was calibrated using at least two known energies to establish a correspondence between gamma-ray energy and channel number.

[0007] Furthermore, the automatic interpretation of the gamma spectrum data based on the energy calibration relationship and the efficiency curve parameters to calculate the type and activity concentration of fission nuclides in the tested pipeline includes: Identify the characteristic peaks in the gamma spectrum data, determine their characteristic gamma ray energies, and determine the corresponding fission nuclides based on the characteristic gamma ray energies; Obtain the net count of the total energy peaks of the characteristic peaks; The detection efficiency of the corresponding energy can be queried based on the efficiency curve parameters; The activity concentration of fission nuclides is calculated based on the net count of the total energy peak, the detection efficiency, the measurement time, and the geometric factor correction coefficient.

[0008] Further, the calculation of the activity concentration of the fissile nuclide based on the net count of the total energy peak, the detection efficiency, the measurement time, and the geometric factor correction coefficient includes: The activity concentration of the i-th fissile nuclide is calculated using the following formula. : in: Let be the activity concentration of the i-th fission nuclide; The net count of the total energy peaks corresponding to the characteristic peaks of the i-th fission nuclide; The characteristic gamma ray energy of the i-th fission nuclide The corresponding detection efficiency is given by: t, measurement time; K, geometric factor correction coefficient; and η, branch ratio.

[0009] The present invention also provides an online monitoring system for pressurized water reactor fuel cladding damage, including a cabinet assembly, a detector assembly, a signal processing assembly, and a data processing assembly; The cabinet assembly is equipped with casters at the bottom and an adjustable guide rail at the top. The detector assembly is mounted on the adjustable guide rail and can move along the guide rail to adjust the distance to the pipe being tested. The detector assembly includes a high-purity germanium main detector, an anti-Compton detector, and a collimator coaxially arranged from the inside to the outside. The collimator has a collimation hole at its front end, and the center of the collimation hole is aligned with the pipe being tested. The signal processing component includes a digital multichannel analyzer, the input of which is connected to the output of the high-purity germanium main detector and the anti-Compton detector, respectively, to receive and process the preamplifier signals output by the detectors and output gamma spectrum data through anti-coincidence measurement technology. The data processing component is connected to the output of the digital multichannel analyzer and is used to perform the following operations: Obtain a pre-established energy scale relationship, which is the correspondence between gamma ray energy and channel number; Obtain pre-calibrated efficiency curve parameters, which are obtained by active efficiency calibration based on a simulated tubular standard radioactive source. The material, outer diameter, and inner diameter of the simulated tubular standard radioactive source are consistent with those of the pipe being tested. The gamma energy spectrum data output by the signal processing component is acquired, and the gamma energy spectrum data is automatically despectrated according to the energy scale relationship and the efficiency curve parameters to calculate the type and activity concentration of fission nuclides in the tested pipeline. Based on the type of fission nuclide, a corresponding preset threshold is determined, and the activity concentration is compared with the preset threshold to obtain the monitoring results of pressurized water reactor fuel cladding damage.

[0010] Furthermore, the anti-Compton detector is a ring structure, surrounding the outside of the high-purity germanium main detector, and is made of bismuth germanate or cesium iodide scintillator material.

[0011] Furthermore, the online monitoring system also includes a counterweight block located inside the cabinet at the bottom, used to balance the overturning moment generated when the detector assembly extends out of the front end of the cabinet.

[0012] Furthermore, the pipeline under test is a vertical sampling line inside the glove box of the primary loop sampling system for measuring hydrogen.

[0013] Furthermore, the collimator is made of a high-density material, and the collimation hole at the front end of the collimator has a replaceable aperture structure for selecting collimation holes of different apertures according to different dose rate environments.

[0014] Furthermore, the cabinet assembly also includes a power module, which is located inside the cabinet and is used to provide operating power to the detector assembly, the signal processing assembly, and the data processing assembly.

[0015] The beneficial effects of this invention are that it relates to an online monitoring method and system for pressurized water reactor fuel cladding damage. The method includes the following steps: Step S1: Obtaining a pre-established energy calibration relationship, wherein the energy calibration relationship is the correspondence between gamma ray energy and channel number; Step S2: Obtaining pre-calibrated efficiency curve parameters, wherein the efficiency curve parameters are obtained based on active efficiency calibration using a simulated tubular standard radioactive source, wherein the material, outer diameter, and inner diameter of the simulated tubular standard radioactive source are consistent with the tested pipe; Step S3: Obtaining gamma spectrum data obtained through anti-coincidence measurement technology, and automatically interpreting the gamma spectrum data according to the energy calibration relationship and the efficiency curve parameters to calculate the type and activity concentration of fissile nuclides in the tested pipe; Step S4: Determining a corresponding preset threshold according to the type of fissile nuclide, and comparing the activity concentration with the preset threshold to obtain the monitoring result of pressurized water reactor fuel cladding damage. This invention uses a simulated pipeline-shaped dedicated radioactive source for efficiency calibration, combines anti-coincidence measurement technology to reduce background, and utilizes energy calibration and efficiency curve parameters to achieve automatic spectrum interpretation. Finally, the activity concentration of fission nuclides is calculated and compared to obtain monitoring results. Attached Figure Description

[0016] The present invention will be further described below with reference to the accompanying drawings and embodiments. In the accompanying drawings: Figure 1 This is a flowchart of the online monitoring method for fuel cladding damage in pressurized water reactors; Figure 2 This is a system principle block diagram of the pressurized water reactor fuel cladding damage online monitoring system; Figure 3 This is a schematic diagram of the structure of an online monitoring system for fuel cladding damage in pressurized water reactors; Figure 4 This is a cross-sectional structural diagram of the detector assembly; Figure 5 This is a diagram showing the suppression effect of the anti-Compton detector. Detailed Implementation

[0017] To provide a clearer understanding of the technical features, objectives, and effects of the present invention, specific embodiments of the invention are now described in detail with reference to the accompanying drawings. In the following description, specific details such as particular structures and techniques are set forth for illustrative purposes and not for limitation, so as to provide a thorough understanding of the embodiments of the invention. However, those skilled in the art will understand that the invention can also be implemented in other embodiments without these specific details. In other instances, detailed descriptions of well-known devices, circuits, and methods are omitted so as not to obscure the description of the invention with unnecessary detail.

[0018] like Figure 1 As shown, Figure 1 This is a logic flowchart of an online monitoring method for fuel cladding damage in pressurized water reactors.

[0019] The online monitoring method for pressurized water reactor fuel cladding damage provided by this invention includes the following steps: Step S1: Obtain the pre-established energy scale relationship, which is the correspondence between gamma ray energy and channel number; In this step, it should be noted that energy calibration is the basis for qualitative analysis of gamma ray spectroscopy. In this embodiment, at least two known energies of radioactive sources are used to perform energy calibration on the detection assembly consisting of a high-purity germanium detector and an anti-Compton detector, establishing the correspondence between gamma ray energy and the number of channels in the multichannel analyzer, so that the corresponding nuclide species can be determined subsequently based on the channel number where the characteristic peak is located.

[0020] Step S2: Obtain the pre-calibrated efficiency curve parameters. The efficiency curve parameters are obtained by active efficiency calibration based on a simulated tubular standard radioactive source. The material, outer diameter, and inner diameter of the simulated tubular standard radioactive source are consistent with the pipe being tested. In this step, it should be noted that the detection efficiency of the external non-destructive measurement method is significantly affected by the pipeline geometry (material, wall thickness, diameter). To ensure measurement accuracy, this embodiment uses a simulated measurement pipeline with the same material, outer diameter, and inner diameter as the pipeline being measured (i.e., the vertical sampling pipeline inside the REN hydrogen meter glove box). A mixed liquid standard radioactive source with known activity concentration is filled with this pipeline, and measurements are performed in a laboratory under geometric conditions identical to those in the field, thereby obtaining accurate efficiency curve parameters. These parameters are crucial for subsequently extrapolating the detector count rate to the coolant activity concentration within the pipeline.

[0021] Step S3: Obtain gamma spectrum data obtained through anti-coincidence measurement technology, and automatically interpret the gamma spectrum data according to the energy scale relationship and efficiency curve parameters to calculate the type and activity concentration of fission nuclides in the tested pipeline. In this step, it should be noted that the anti-combination measurement technique using an anti-Compton detector and a high-purity germanium main detector effectively suppresses the Compton scattering background generated by high-energy gamma rays in the low-energy region, significantly improving the signal-to-noise ratio of the characteristic peaks of low-energy fission nuclides (such as 0.081 MeV of Xe-133). After acquiring clean gamma spectrum data, the measurement software automatically identifies characteristic peaks and determines the nuclide type based on the energy scale relationship. Then, combined with efficiency curve parameters and measurement time, it automatically calculates the activity concentration of each nuclide.

[0022] Step S4: Determine the corresponding preset threshold according to the type of fission nuclide, and compare the activity concentration with the preset threshold to obtain the monitoring results of pressurized water reactor fuel cladding damage.

[0023] In this step, it should be noted that for different fissile nuclides (such as I-131, Cs-137, Xe-133, etc.), corresponding early warning thresholds are pre-set based on nuclear power plant operation technical specifications and historical data. The system compares the real-time activity concentration calculated in step S3 with the corresponding threshold. When the measured value exceeds the threshold, the system automatically determines that pressurized water reactor fuel cladding may have been damaged and issues an alarm signal, realizing a leap from "manual sampling and analysis" to "online real-time early warning".

[0024] Further, step S1 includes: using at least two known energies of radiation sources to perform energy calibration on the detection component and establishing a correspondence between gamma ray energy and channel number.

[0025] Specifically, in this embodiment, point radioactive sources with low activity such as Co-60 (1173.2keV, 1332.5keV) and Cs-137 (661.6keV) are typically used. These sources are placed at the front end of the detector, and the energy spectrum is collected by a digital multichannel analyzer. The characteristic peak positions are identified, and linear or quadratic fitting of energy and channel number is completed to ensure the accuracy of energy measurement.

[0026] Furthermore, based on the energy scale relationship and efficiency curve parameters, the gamma spectrum data is automatically interpreted to calculate the types and activity concentrations of fission nuclides in the tested pipeline, including: identifying characteristic peaks in the gamma spectrum data, determining their characteristic gamma ray energies, and determining the corresponding fission nuclide types based on the characteristic gamma ray energies; obtaining the net count of the full-energy peaks of the characteristic peaks; and querying the detection efficiency of the corresponding energies based on the efficiency curve parameters. The activity concentration of fission nuclides is calculated based on the net count of the total energy peak, detection efficiency, measurement time, and geometric factor correction coefficient.

[0027] Specifically, firstly, peak finding is performed on the smoothed energy spectrum, and the net count of all-energy peaks Ni is obtained after subtracting the background. Then, the corresponding detection efficiency ε(Ei) is obtained by interpolation from the efficiency curve pre-stored in step S2 based on the peak energy Ei. Finally, the activity concentration is calculated by substituting the set measurement time t and the branching ratio η into the formula. Since a simulated tubular source is used for calibration, the geometric factor K is implicitly considered in the efficiency calibration process and is usually set to 1.

[0028] Furthermore, based on the net count of the full-energy peak, detection efficiency, measurement time, and geometric factor correction coefficient, the activity concentration of the fission nuclide is calculated as follows: The activity concentration of the i-th fissile nuclide is calculated using the following formula. : in: Let be the activity concentration of the i-th fission nuclide; The net count of the total energy peaks corresponding to the characteristic peaks of the i-th fission nuclide; The characteristic gamma ray energy of the i-th fission nuclide The corresponding detection efficiency is given by: t, measurement time; K, geometric factor correction coefficient; and η, branch ratio.

[0029] Specifically, in this embodiment, because the geometric distribution, medium density, and chemical composition of the calibration source (simulating a tubular source) and the sample to be tested (coolant inside the pipe) are highly consistent, the K value is very close to 1. The branching ratio η is an inherent decay parameter of the nuclide, obtained from a nuclear database. The system is set to a measurement period t of 1 hour, outputting the activity concentration measurement result once per hour, thus reducing the warning time from 3 days in the traditional method to less than 1 hour.

[0030] like Figure 2 and Figure 3As shown, the present invention also provides an online monitoring system for pressurized water reactor fuel cladding damage, including a cabinet assembly 1, a detector assembly 2, a signal processing assembly 3, and a data processing assembly 4. The cabinet assembly 1 has casters at its bottom and an adjustable guide rail on its upper part. The detector assembly 2 is mounted on the adjustable guide rail and can move along the guide rail to adjust the distance to the pipe being measured. The detector assembly 2 includes a high-purity germanium main detector, an anti-Compton detector, and a collimator 5 coaxially arranged from the inside out. The collimator 5 has a collimation hole at its front end, the center of which is aligned with the pipe being measured. The signal processing assembly 3 includes a digital multichannel analyzer. The input terminals of the digital multichannel analyzer are connected to the output terminals of the high-purity germanium main detector and the anti-Compton detector, respectively, for receiving and processing the preamplifier signals output by the detectors, and outputting gamma energy through anti-coincidence measurement technology. The data processing component 4 is connected to the output of the digital multichannel analyzer and is used to perform the following operations: acquire a pre-established energy calibration relationship, which is the correspondence between gamma ray energy and channel number; acquire pre-calibrated efficiency curve parameters, which are obtained based on active efficiency calibration using a simulated tubular standard radioactive source, with the material, outer diameter, and inner diameter of the simulated tubular standard radioactive source being consistent with the tested pipe; acquire the gamma energy spectrum data output by the signal processing component 3, and automatically interpret the gamma energy spectrum data according to the energy calibration relationship and efficiency curve parameters to calculate the type and activity concentration of fission nuclides in the tested pipe; determine the corresponding preset threshold according to the type of fission nuclide, and compare the activity concentration with the preset threshold to obtain the monitoring results of pressurized water reactor fuel cladding damage.

[0031] like Figure 3 As shown, it should be noted that this system is a portable monitoring solution specifically designed for the limited space within the hydrogen meter glove box of the REN (Nuclear Array) system. The "sleeve-type" design of detector assembly 2 (HPGe main detector + annular anti-coincidence detector + outer collimator 5) effectively reduces background interference through anti-coincidence measurement while ensuring detection efficiency. The adjustable guide rails on the upper part of the cabinet allow on-site adjustment of the distance between the probe and the measured pipeline to cope with different dose rate levels; the casters at the bottom allow for flexible deployment and relocation of the equipment in confined spaces, facilitating equipment maintenance during overhauls.

[0032] The preamplifier signal refers to the raw pulse signal output by the detector. After detecting gamma rays, the high-purity germanium main detector and the anti-Compton detector generate weak charge signals. These are amplified by the preamplifier to form a voltage pulse signal, which contains the energy information of the gamma rays. The pulse amplitude is proportional to the energy of the incident gamma rays. Anti-coincidence measurement is a signal processing technique that reduces measurement background. When the high-purity germanium main detector and the ring-shaped anti-Compton detector simultaneously receive signals within a very short time, it indicates that the event is caused by Compton scattering, and the system classifies it as a scattering event and discards it. Only when the main detector has a signal but the anti-Compton detector has no signal is it considered a valid event and recorded, thus effectively suppressing the Compton plateau region of the energy spectrum and improving the signal-to-noise ratio of low-energy characteristic peaks. Automatic spectrum interpretation refers to the qualitative and quantitative analysis process of the gamma spectrum, including automatically identifying characteristic peaks, determining the nuclide type based on the energy scale, calculating the net count of the full-energy peaks, and calculating the activity concentration of fission nuclides by combining efficiency curve parameters and measurement time. The entire process requires no manual intervention. Active efficiency calibration refers to the process of establishing a correspondence between detector efficiency and gamma ray energy using a standard radioactive source with known activity concentration.

[0033] This invention employs a simulated tubular standard radiation source with the same material, outer diameter, and inner diameter as the pipe being measured. Measurements are performed in a laboratory setting with geometric conditions identical to those in the field, thus obtaining accurate efficiency curve parameters and fundamentally ensuring the accuracy of the measurement results. The distance between the probe and the pipe being measured can be manually adjusted. When the dose rate of the pipe is high, the distance between the probe and the pipe is increased; when the dose rate is low, the distance is decreased, ensuring that the intensity of the gamma rays received by the detector is within a suitable range. Figure 3 As shown, the guide rail for adjusting the detector distance using the handwheel 6 can be adjusted manually, electrically, or hydraulically.

[0034] like Figure 4 As shown, Figure 4 This is a cross-sectional view of detector assembly 2. The anti-Compton detector is a ring-shaped structure that surrounds the outside of the high-purity germanium main detector and is made of bismuth germanate or cesium iodide scintillator material.

[0035] Specifically, this embodiment preferably uses a cesium iodide (CsI) ring detector as the anti-Compton detector, which has high density and good light yield, and can effectively capture gamma photons scattered by the main detector's Compton scattering. Through an anti-coincidence circuit, when the main detector and the anti-Compton detector receive signals simultaneously within a very short time, the system determines the event as a Compton scattering event and eliminates it, thereby significantly suppressing the Compton plateau region of the energy spectrum. Experimental data show that with this design, the Compton plateau suppression coefficient for Co-60 can reach over 3.0, greatly improving the detection limit for low-energy nuclides such as I-131 and Xe-133.

[0036] Furthermore, the online monitoring system also includes a counterweight located at the bottom of the cabinet to balance the overturning moment generated when the detector assembly 2 extends out of the front of the cabinet.

[0037] Specifically, because the monitoring point (inside the hydrogen meter glove box) requires the detector and collimator 5 to "protrude forward" for measurement, the center of gravity of the entire cabinet shifts forward, posing a risk of tipping over. This system places a certain weight of lead blocks as a counterweight in the lower rear space of the cabinet, effectively balancing the tipping torque caused by the probe's forward extension and ensuring the safety of the equipment for long-term stable operation in a confined space.

[0038] Furthermore, the pipeline under test is a vertical sampling line inside the glove box of the primary loop sampling system for measuring hydrogen.

[0039] Specifically, the selection of the vertical sampling pipeline within the glove box of the REN system's hydrogen meter as the monitoring point was a creative discovery based on the compact layout of the M310 / CPR1000 unit. This location, originally intended for manual hydrogen sampling, offers the following unique advantages: 1) the pipeline carries a representative sample of the primary coolant; 2) its location within the glove box ensures a relatively clean environment, good personnel accessibility, and facilitates equipment installation and maintenance; and 3) the dose rate is moderate, guaranteeing both the measurement signal strength and minimizing the risk of excessive radiation damage to the detector.

[0040] Furthermore, the collimator 5 is made of high-density material, and the collimation hole at the front end of the collimator 5 has a replaceable aperture structure for selecting collimation holes of different apertures according to different dose rate environments.

[0041] Specifically, in this embodiment, the collimator 5 is made of high-density materials such as tungsten or lead, effectively shielding background radiation from non-target directions. The collimating aperture is designed as a replaceable structure, allowing for the replacement of collimating apertures of different diameters on-site based on the actual radioactivity level of the coolant in the pipeline. When the primary circuit dose rate is high (e.g., in the later stages of damage), a small aperture is selected to reduce the count rate entering the detector and prevent peak accumulation; when the dose rate is low (e.g., in the early stages of damage), a large aperture is selected to improve detection sensitivity. The collimator 5 has a certain thickness to shield surrounding gamma rays, providing a low-background measurement environment.

[0042] Furthermore, the cabinet assembly 1 also includes a power module, which is located inside the cabinet and is used to provide operating power to the detector assembly 2, the signal processing assembly 3, and the data processing assembly 4.

[0043] Specifically, to ensure the continuity and stability of monitoring, a UPS (Uninterruptible Power Supply) is integrated inside the cabinet. In the event of a main power failure or a temporary power outage on site, the UPS can provide backup power to the system, ensuring the preservation of critical data and safe system shutdown, thus avoiding the risk of data loss and equipment damage due to power failure.

[0044] like Figure 5 As shown, Figure 5 This is a diagram illustrating the suppression effect of the anti-Compton detector. The method employs the non-destructive external gamma spectroscopy monitoring provided by this invention. This solution requires no modification to the existing systems of the nuclear power plant, and the measurement frequency is significantly increased compared to traditional manual sampling radiochemical analysis methods. This significantly enhances the nuclear power plant's early warning capability for pressurized water reactor fuel cladding damage, reducing the warning time from the traditional 3 days to less than 1 hour. This helps the nuclear power plant to promptly detect damage and quickly respond to formulate countermeasures, effectively controlling and reducing the diffusion, migration, and deposition of radionuclides in the main loop and auxiliary systems, thereby reducing the radiation dose level of the power plant, reducing the emission activity of gaseous and liquid effluents, effectively protecting the radiation safety of workers, and ensuring the safety of the surrounding environment and public health.

[0045] By adopting this invention, fully automated monitoring and early warning are achieved, significantly reducing the frequency of periodic manual sampling and radiochemical analysis. This not only reduces the workload of personnel but also significantly lowers the risk of contamination and the radiation dose received by personnel. Furthermore, the system's real-time monitoring capabilities help reduce the risk of unplanned reactor shutdowns and minimize significant economic losses caused by such shutdowns.

[0046] In terms of technical performance, this invention employs a combination of a high-purity germanium (HPGe) detector and a ring-shaped anti-Compton detector. Compared to using an HPGe detector alone, this system effectively reduces the Compton plateau region of the gamma spectrum through anti-coincidence measurement technology, significantly improving the signal quality of low-energy gamma rays and lowering the detection limit of low-energy fissile nuclides. Taking Co-60 as an example, the Compton plateau suppression coefficient reaches over 3.0 after using the anti-Compton detector.

[0047] In terms of engineering practicality, the online monitoring system adopts a wheeled mobile design, which has high flexibility and facilitates equipment inspection and maintenance.

[0048] In summary, the main innovations of this invention are reflected in the following four aspects: Innovative Solution: The primary loop water sampling riser inside the hydrogen meter glove box of the REN system is used as an online monitoring point. This riser serves as a natural sampling delay pipeline. By installing a high-purity germanium or sodium iodide detector directly on the outside of the riser and configuring a local shielding device, real-time online detection of characteristic gamma rays of fission nuclides (such as I-131, Xe-133, etc.) in the primary loop coolant flowing through this section of the pipeline can be achieved.

[0049] Innovative detector combination: The combination of a high-purity germanium detector and a ring-shaped anti-Compton detector for anti-coincidence measurement has been used to specifically improve the signal-to-noise ratio of low-energy characteristic gamma rays in the primary coolant.

[0050] Innovative structural design: The mobile cabinet design integrates adjustable guide rails, counterweight balance, and wheeled movement, achieving stable and flexible measurement within a limited space.

[0051] Innovative calibration method: The active efficiency is calibrated using a "simulated tubular standard radioactive source" that is completely consistent with the actual pipe material, outer diameter, and inner diameter, which fundamentally ensures the accuracy of the measurement results.

[0052] It is understood that the above embodiments only illustrate preferred embodiments of the present invention, and their descriptions are relatively specific and detailed, but they should not be construed as limiting the scope of the present invention. It should be noted that those skilled in the art can freely combine the above technical features without departing from the concept of the present invention, and can also make several modifications and improvements, all of which fall within the protection scope of the present invention. Therefore, all equivalent transformations and modifications made with respect to the scope of the claims of the present invention should fall within the scope of the claims of the present invention.

Claims

1. A method for online monitoring of fuel cladding damage in pressurized water reactors, characterized in that, The method includes the following steps: Step S1: Obtain a pre-established energy scale relationship, which is the correspondence between gamma ray energy and channel number; Step S2: Obtain the pre-calibrated efficiency curve parameters, which are obtained by active efficiency calibration based on a simulated tubular standard radioactive source. The simulated tubular standard radioactive source has the same material, outer diameter, and inner diameter as the pipe being tested. Step S3: Obtain gamma spectrum data obtained through anti-coincidence measurement technology, and automatically interpret the gamma spectrum data according to the energy scale relationship and the efficiency curve parameters to calculate the type and activity concentration of fission nuclides in the tested pipeline. Step S4: Determine the corresponding preset threshold according to the type of fission nuclide, and compare the activity concentration with the preset threshold to obtain the monitoring results of pressurized water reactor fuel cladding damage.

2. The online monitoring method for pressurized water reactor fuel cladding damage according to claim 1, characterized in that, Step S1 includes: The detector assembly was calibrated using at least two known energies to establish a correspondence between gamma-ray energy and channel number.

3. The online monitoring method for pressurized water reactor fuel cladding damage according to claim 1, characterized in that, The automatic interpretation of the gamma spectrum data based on the energy calibration relationship and the efficiency curve parameters, and the calculation of the types and activity concentrations of fission nuclides in the tested pipeline, include: Identify the characteristic peaks in the gamma spectrum data, determine their characteristic gamma ray energies, and determine the corresponding fission nuclides based on the characteristic gamma ray energies; Obtain the net count of the total energy peaks of the characteristic peaks; The detection efficiency of the corresponding energy can be queried based on the efficiency curve parameters; The activity concentration of fission nuclides is calculated based on the net count of the total energy peak, the detection efficiency, the measurement time, and the geometric factor correction coefficient.

4. The online monitoring method for pressurized water reactor fuel cladding damage according to claim 3, characterized in that, The calculation of the activity concentration of the fission nuclide based on the net count of the total energy peak, the detection efficiency, the measurement time, and the geometric factor correction coefficient includes: The activity concentration of the i-th fissile nuclide is calculated using the following formula. : in: Let be the activity concentration of the i-th fission nuclide; The net count of the total energy peaks corresponding to the characteristic peaks of the i-th fission nuclide; The characteristic gamma ray energy of the i-th fission nuclide The corresponding detection efficiency, where t is the measurement time and K is the geometric factor correction coefficient; The branching ratio.

5. An online monitoring system for pressurized water reactor fuel cladding damage, characterized in that, This includes cabinet components, detector components, signal processing components, and data processing components; The cabinet assembly is equipped with casters at the bottom and an adjustable guide rail at the top. The detector assembly is mounted on the adjustable guide rail and can move along the guide rail to adjust the distance to the pipe being tested. The detector assembly includes a high-purity germanium main detector, an anti-Compton detector, and a collimator coaxially arranged from the inside to the outside. The collimator has a collimation hole at its front end, and the center of the collimation hole is aligned with the pipe being tested. The signal processing component includes a digital multichannel analyzer, the input of which is connected to the output of the high-purity germanium main detector and the anti-Compton detector, respectively, to receive and process the preamplifier signals output by the detectors and output gamma spectrum data through anti-coincidence measurement technology. The data processing component is connected to the output of the digital multichannel analyzer and is used to perform the following operations: Obtain a pre-established energy scale relationship, which is the correspondence between gamma ray energy and channel number; Obtain pre-calibrated efficiency curve parameters, which are obtained by active efficiency calibration based on a simulated tubular standard radioactive source. The material, outer diameter, and inner diameter of the simulated tubular standard radioactive source are consistent with those of the pipe being tested. The gamma energy spectrum data output by the signal processing component is acquired, and the gamma energy spectrum data is automatically despectrated according to the energy scale relationship and the efficiency curve parameters to calculate the type and activity concentration of fission nuclides in the tested pipeline. Based on the type of fission nuclide, a corresponding preset threshold is determined, and the activity concentration is compared with the preset threshold to obtain the monitoring results of pressurized water reactor fuel cladding damage.

6. The pressurized water reactor fuel cladding damage online monitoring system according to claim 5, characterized in that, The anti-Compton detector is a ring-shaped structure, surrounding the outside of the high-purity germanium main detector, and is made of bismuth germanate or cesium iodide scintillator material.

7. The pressurized water reactor fuel cladding damage online monitoring system according to claim 5, characterized in that, The online monitoring system also includes a counterweight block located inside the cabinet at the bottom, used to balance the overturning moment generated when the detector assembly extends out of the front end of the cabinet.

8. The pressurized water reactor fuel cladding damage online monitoring system according to claim 5, characterized in that, The pipeline under test is a vertical sampling line inside the glove box of a primary loop sampling system for measuring hydrogen.

9. The pressurized water reactor fuel cladding damage online monitoring system according to claim 5, characterized in that, The collimator is made of high-density material, and the collimation hole at the front end of the collimator has a replaceable aperture structure for selecting collimation holes of different apertures according to different dose rate environments.

10. The pressurized water reactor fuel cladding damage online monitoring system according to claim 5, characterized in that, The cabinet assembly also includes a power module, which is located inside the cabinet and is used to provide operating power to the detector assembly, the signal processing assembly, and the data processing assembly.

Images