Reactor building radioactive waste liquid measuring device, sampling measuring system and method

Abstract

The application discloses a reactor plant radioactive waste liquid measuring device, a sampling measuring system and a method. The reactor plant radioactive waste liquid measuring device comprises a measuring module, a first quantitative sampling module and a radioactive control module. The measuring module comprises a measured liquid container, a first liquid level monitoring module and a radioactive control module. The first liquid level monitoring module is arranged on the side of the measured liquid container. The radioactive control module is arranged below the measured liquid container. The measured liquid container is used for containing the measured radioactive waste liquid. The first liquid level monitoring module is used for monitoring the liquid level in the measured liquid container. The radioactive control module is used for measuring the radioactive concentration of the radioactive waste liquid. The first quantitative sampling module is connected with the measured liquid container. The first quantitative sampling module is used for quantitatively injecting the radioactive waste liquid into the measured liquid container. The above technical scheme realizes the measurement without the intervention of the staff, and guarantees the radiation safety of the staff.

Description

Technical Field

[0001] This invention relates to the field of nuclear power plant waste liquid sampling and measurement technology under accident conditions, and particularly to a radioactive waste liquid measuring device, sampling and measurement system and method. Background Technology

[0002] In the event of a design basis accident (such as a large breach loss-of-coolant accident) or a severe accident (core damage), a nuclear power plant may release large amounts of radioactive material into the reactor building in both gaseous and liquid forms. Gaseous radioactive material diffuses into the reactor building atmosphere, while liquid radioactive material flows into a sump and is collected. To accurately assess the severity of a nuclear accident within the reactor building, measurements from on-site temperature, humidity, pressure, and radiation monitoring instruments can be used. Alternatively, samples can be taken from the reactor building atmosphere and sump waste liquid and analyzed in a laboratory to determine the concentration of radionuclides. The sampling and analysis results of radioactive waste liquid from the reactor building are crucial decision-making information for the nuclear power plant's emergency command center regarding the orderly and controlled release of radioactive waste liquid into the environment.

[0003] Under severe accident conditions, the radioactivity concentration of the waste liquid in the sump can reach as high as 10. 15 Bq / m 3 The dose rate near the sampling pipeline is above 5 mSv / h, which will bring a large radiation dose to the staff who sample and analyze the waste liquid. Therefore, a safer radioactive waste liquid sampling and measurement system is needed. Summary of the Invention

[0004] The embodiments of the present invention provide a device, sampling and measurement system and method for measuring radioactive waste liquid in a reactor building, which aims to solve the problem that sampling and measuring radioactive waste liquid in the sump of a reactor building after a design basis accident or a serious accident in a nuclear power plant poses a radiation safety threat to workers.

[0005] In a first aspect, embodiments of the present invention provide a reactor building radioactive waste liquid measurement device comprising: a measurement module, including a test liquid container, a first liquid level monitoring module, and a radioactivity measurement and control module, wherein the first liquid level monitoring module is disposed on the periphery of the test liquid container, and the radioactivity measurement and control module is disposed below the test liquid container, wherein the test liquid container is used to contain the radioactive waste liquid to be tested, the first liquid level monitoring module is used to monitor the liquid level in the test liquid container, and the radioactivity measurement and control module is used to measure the radioactivity activity of the radioactive waste liquid; and a first quantitative injection module, the first quantitative injection module being connected to the test liquid container, and the first quantitative injection module being used to quantitatively inject radioactive waste liquid into the test liquid container.

[0006] Secondly, embodiments of the present invention also provide a sampling and measurement system for radioactive waste liquid in a reactor building, comprising: a radioactive waste liquid measurement unit, including the radioactive waste liquid measurement device for a reactor building as described above; a liquid inlet unit, the liquid inlet unit being used to allow radioactive waste liquid in the reactor building sump to pass through the containment vessel and enter the sampling and measurement system; a liquid outlet unit, connected to the radioactive waste liquid measurement unit, the liquid outlet unit being used to discharge the liquid in the sampling and measurement system through the containment vessel and into the reactor building sump; a radioactive waste liquid sampling unit, connected to the liquid inlet unit and the radioactive waste liquid measurement unit, the radioactive waste liquid sampling unit being used to sample the radioactive waste liquid; and a control unit, electrically connected to the radioactive measurement unit, the radioactive waste liquid sampling unit, the liquid inlet unit and the liquid outlet unit, the control unit being used to control the operation of the sampling and measurement system.

[0007] Thirdly, embodiments of the present invention also provide a method for sampling and measuring radioactive waste liquid in a reactor building, applied to the radioactive waste liquid sampling and measurement system for a reactor building as described above, comprising: opening the inlet unit and the outlet unit to allow radioactive waste liquid in the reactor building sump to enter the radioactive liquid sampling and measurement system for a reactor building; obtaining the gamma dose caused by the radioactive waste liquid in the reactor building sump using a gamma dose rate meter, and selecting a sampling mode based on the gamma dose caused by the radioactive waste liquid in the reactor building sump; starting the radioactive waste liquid sampling unit and sampling the radioactive waste liquid according to the sampling mode; starting the radioactive waste liquid measurement unit and measuring the radioactive concentration of the radioactive waste liquid according to the sampling mode; closing the inlet unit and the outlet unit, and taking sample liquid from the radioactive waste liquid sampling unit.

[0008] Compared with the prior art, the beneficial effects of the present invention are:

[0009] In the technical solution of this invention, a radioactive waste liquid measuring device is set up, in which a test liquid container for containing radioactive waste liquid is set up. The radioactive waste liquid is quantitatively injected into the test liquid container through a first quantitative injection module. A first liquid level monitoring module is arranged around the test liquid container to detect the liquid level. A radioactive measurement and control module is arranged at the bottom of the test liquid container to measure the radioactive concentration of the radioactive waste liquid, thereby obtaining the radioactive concentration of the radioactive waste liquid in the pit in the reactor building. This allows the measurement of the radioactive concentration of the radioactive waste liquid in the pit in the reactor building to be completed without the intervention of personnel, ensuring the radiation safety of personnel. Attached Figure Description

[0010] To more clearly illustrate the technical solutions of the embodiments of the present invention, the drawings used in the following description of the embodiments will be briefly introduced. Obviously, the drawings described below are some embodiments of the present invention. For those skilled in the art, other drawings can be obtained based on these drawings without creative effort.

[0011] Figure 1 A schematic diagram of the radioactive waste sampling and measurement system for reactor buildings provided by the present invention;

[0012] Figure 2 A schematic diagram of the second multi-way valve of the reactor building radioactive waste liquid sampling and measurement system provided by the present invention in quantitative mode;

[0013] Figure 3 A schematic diagram of the first multi-way valve of the reactor building radioactive waste liquid sampling and measurement system provided by the present invention in quantitative mode;

[0014] Figure 4 A schematic diagram illustrating the steps of the method for sampling and measuring radioactive waste liquid in a reactor building provided by the present invention;

[0015] Figure 5 A schematic diagram of a sub-step in the method for sampling and measuring radioactive waste liquid in a reactor building provided by the present invention;

[0016] Figure 6 A schematic diagram of another sub-step in the method for sampling and measuring radioactive waste liquid in a reactor building provided by the present invention;

[0017] Figure 7 A schematic diagram of the control relationship of the measurement system control platform in an embodiment of the radioactive waste liquid sampling and measurement method for reactor buildings provided by the present invention;

[0018] Figure 8 A flowchart illustrating the sampling and measurement process of highly radioactive waste liquid in a reactor building under accident conditions in an embodiment of the radioactive waste liquid sampling and measurement method for reactor buildings provided by the present invention.

[0019] Figure 9 A control flowchart of the monitoring module controller in an embodiment of the method for sampling and measuring radioactive waste liquid in a reactor building provided by the present invention;

[0020] Explanation of reference numerals in the attached figures:

[0021] 10. Liquid inlet module; 101. Electric valve No. 101; 102. Electric valve No. 102; 103. Electric valve No. 103; 104. Electric valve No. 104; 11. Liquid drain module; 105. Electric valve No. 105; 106. Electric valve No. 106; 107. Electric valve No. 107; 108. Electric valve No. 108;

[0022] 20. Flow control module; 21. Flushing module; 201. Gamma dose rate meter; 202. Electric valve No. 202; 203. First electric sampling pump; 204. Sampling cooler; 205. Electric valve No. 205; 206. Electric valve No. 206; 207. Flow meter No. 207;

[0023] 31. Radioactive waste liquid sampling unit; 32. Second quantitative injection module; 33. Sampling module; 34. First quantitative injection module; 301. Second multi-port valve; 302. Second quantitative loop; 303. Electric valve No. 303; 304. Electric valve No. 304; 305. Flow meter No. 305; 306. Electric valve No. 306; 307. Second quick connector; 308. Collection container; 309. Second shielding cover; 310. Third quick connector; 311. Second electric sampling pump; 312. Electric valve No. 312; 313. Electric valve No. 313; 314. Flow meter No. 314; 315. First quantitative loop; 316. Electric valve No. 316; 317. Electric valve No. 317; 318. Flow meter No. 318; 319. Second liquid level monitoring module; 320. First multi-port valve;

[0024] 30101, Second inlet; 30102, Second outlet; 30103, Second connection end of the second metering loop; 30104, Second diluent injection end; 30105, Collection container connection end; 30106, First connection end of the second metering loop;

[0025] 32001, First liquid inlet; 32002, First liquid outlet; 32003, Second connection end of the first metering loop; 32004, First diluent injection end; 32005, Connection end of the test liquid container; 32006, First connection end of the first metering loop;

[0026] 40. Measurement module; 401. Electric valve No. 401; 402. Electric valve No. 402; 403. First quick connector; 404. Test liquid container; 405. First liquid level monitoring module; 406. High-purity germanium detector; 407. Cooler; 408. Temperature and humidity controller; 409. First shielding shell; 410. Top cover of the first shielding shell; 411. Measurement workbench;

[0027] 50. Injection module; 501. Electric valve No. 501; 502. Injector; 503. Electric valve No. 503;

[0028] 60. Auxiliary components; 601. Electric valve No. 601; 602. Electric valve No. 602; 603. Electric valve No. 603; 604. Electric valve No. 604; 605. Electric valve No. 605; 606. Electric valve No. 606; 607. Electric valve No. 607; 608. Electric valve No. 608; 610. Diluent source;

[0029] 70. Measurement system control platform; 701. Electric valve controller; 702. Sampling pump controller; 703. Monitoring module controller; 704. Main amplifier of gamma-ray spectroscopy measurement system; 705. Multichannel analyzer; 706. Gamma-ray spectroscopy analysis software; 707. High-purity germanium spectrometer controller. Detailed Implementation

[0030] The technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the described embodiments are only some, not all, of the embodiments of the present invention. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without creative effort are within the scope of protection of the present invention.

[0031] It should be understood that, when used in this specification and the appended claims, the terms "comprising" and "including" indicate the presence of the described features, integrals, steps, operations, elements and / or components, but do not exclude the presence or addition of one or more other features, integrals, steps, operations, elements, components and / or collections thereof.

[0032] It should also be understood that the terminology used in this specification is for the purpose of describing particular embodiments only and is not intended to limit the invention. As used in this specification and the appended claims, the singular forms “a,” “an,” and “the” are intended to include the plural forms unless the context clearly indicates otherwise.

[0033] It should also be further understood that the term "and / or" as used in this specification and the appended claims refers to any combination of one or more of the associated listed items and all possible combinations, and includes such combinations.

[0034] To address the radiation safety concerns posed by sampling and measuring radioactive waste liquid in the reactor building's sump after a design basis accident or severe accident at a nuclear power plant, this invention proposes a radioactive waste liquid measurement device for reactor buildings. The device comprises: a measurement module 40, including a test liquid container 404, a first liquid level monitoring module 405, and a radioactivity monitoring and control module. The first liquid level monitoring module 405 is located around the test liquid container 404, and the radioactivity monitoring and control module is located below the test liquid container 404. The test liquid container 404 is used to contain the radioactive waste liquid to be measured. The first liquid level monitoring module 405 is used to monitor the liquid level in the test liquid container 404, and the radioactivity monitoring and control module is used to measure the radioactivity activity of the waste liquid in the test liquid container 404. A first quantitative injection module 34 is connected to the test liquid container 404 and is used to quantitatively inject radioactive waste liquid into the test liquid container 404.

[0035] Reference Figure 1The test liquid container 404 is the core component of the device, used to contain the radioactive waste liquid to be tested. The test liquid container 404 has certain corrosion resistance and pressure resistance to ensure safe storage and operation. The first liquid level monitoring module 405 is responsible for real-time monitoring of changes in the liquid level within the container. The first liquid level monitoring module 405 is located around the periphery of the test liquid container 404 to ensure the liquid level is within a safe range, preventing overflow or leakage, and ensuring that the amount of radioactive waste liquid being measured in the test liquid container 404 is at a preset required level. In this embodiment of the invention, the preset required level of radioactive waste liquid in the test liquid container 404 is 100 mL. To better monitor the liquid level of the radioactive waste liquid in the test liquid container 404, considering the corrosiveness of the radioactive waste liquid and the potential for radioactive contamination upon direct contact, in this embodiment of the invention, the liquid level monitoring module preferably uses a photoelectric liquid level sensor, a laser liquid level sensor, or an ultrasonic liquid level sensor. The so-called radioactivity monitoring and control module is specifically designed to measure the radioactivity concentration of radioactive waste liquid, providing real-time radiation level data to support the nuclear power plant emergency command center's decision-making regarding the orderly and controlled discharge of radioactive waste liquid from the reactor building into the environment. The radioactivity monitoring and control module is located at the bottom of the test liquid container 404. This is because radioactive substances in the waste liquid may form a concentration gradient in the liquid due to factors such as precipitation and stratification. Placing the module at the bottom of the container allows for more effective detection of the radioactivity concentration at the bottom of the liquid, ensuring that any potentially high-concentration areas are captured. Simultaneously, the liquid surface may be affected by fluctuations, bubbles, or vapors, which can interfere with the measurement. Measuring at the bottom reduces this interference, providing a more accurate reflection of the true radioactivity level of the liquid. Through the cooperation of liquid level monitoring and the radioactivity monitoring and control module, online monitoring of the radioactive waste liquid can be achieved. The first quantitative injection module 34 quantitatively injects the radioactive waste liquid into the test liquid container 404, enabling precise control of the injection volume and ensuring the accuracy and consistency of the test. Under the action of the first quantitative sampling module 34, the reactor building radioactive waste liquid measuring device of the present invention can realize online monitoring of a quantitative amount of radioactive waste liquid, thereby completing the measurement of the radioactive concentration of the reactor building radioactive waste liquid after the accident without the need for personnel intervention.

[0036] Compared to existing technologies, the radioactive waste liquid measuring device for reactor buildings of the present invention, by setting up a radioactive waste liquid measuring device and setting up a test liquid container 404 to contain radioactive waste liquid, realizes the quantitative injection of radioactive waste liquid through a first quantitative injection module 34, arranges a first liquid level monitoring module 405 around the test liquid container 404 to detect the liquid level, and arranges a radioactivity measurement and control module at the bottom of the test liquid container 404 to measure the radioactivity concentration of the radioactive waste liquid, thereby obtaining the radioactivity concentration of the radioactive waste liquid in the pit of the reactor building, so that the radioactivity concentration of the radioactive waste liquid in the pit of the reactor building can be measured without the intervention of personnel, thus ensuring the radiation safety of personnel.

[0037] In one embodiment, reference is made to Figure 1The measurement module 40 includes a first shielding shell 409, with a test liquid container 404 disposed inside the first shielding shell 409. The radioactivity measurement and control module includes a high-purity germanium detector 406, a cooler 407, and a temperature and humidity controller 408. The high-purity germanium detector 406 is located below the test liquid container 404, the cooler 407 is connected to the high-purity germanium detector 406, and the temperature and humidity controller 408 is located around the high-purity germanium detector 406. The first shielding shell 409 is used to shield against gamma rays from the environment and the universe, and the high-purity germanium detector 406 is used to convert the gamma rays from the radioactive waste liquid into electrical signals. The first shielding shell 409 is set up to prevent interference from environmental and cosmic rays, especially gamma rays, on the measurement results. The first shielding shell 409 is made of lead or other high-density materials to ensure that the influence of environmental radiation and cosmic rays on the detector is minimized during the measurement process. The test liquid container 404 is housed within a first shielding shell 409, and a high-purity germanium detector 406 is positioned below it. In this configuration, the first shielding shell 409 not only shields the radioactive waste liquid in the test liquid container 404 from gamma rays but also protects the high-purity germanium detector 406 from external environmental radiation and cosmic rays, thus enhancing the safety and accuracy of the radioactive waste liquid measurement device in the reactor building. The high-purity germanium detector 406 is a highly sensitive gamma-ray detector capable of accurately measuring gamma rays emitted by the radioactive waste liquid and converting them into electrical signals. These electrical signals are used for real-time monitoring, data recording, and subsequent statistical analysis. The high-purity germanium detector 406 has high energy resolution, enabling it to identify gamma rays of different energies, facilitating the analysis of the nuclide types and radioactive concentrations in the radioactive waste liquid. Since the high-purity germanium detector 406 operates at low temperatures to achieve optimal performance, a cooler 407 is required and connected to it. Cooler 407 is responsible for cooling the high-purity germanium detector 406 to the required operating temperature, which is between -80℃ and -200℃, with an optimal temperature below -175℃. Maintaining a stable operating temperature reduces thermal noise and improves detection sensitivity. Temperature and humidity controller 408 monitors the temperature environment of the high-purity germanium detector 406 and ensures it operates in a dry state, protecting it from damage caused by condensation or ice buildup from the cooler 407. When the temperature and humidity controller 408 detects that the current temperature and humidity are unsuitable for the detector's operation, it feeds this information back through its connected control circuit, and cooler 407 adjusts the temperature accordingly. Through the solution of the present invention, the measurement device for radioactive waste liquid in reactor buildings has been significantly improved in terms of measurement accuracy and environmental adaptability. It can accurately measure gamma rays in radioactive waste liquid and then calculate the radioactive concentration of waste liquid in the sump in the reactor building.

[0038] Furthermore, a first shielding shell top cover 410 is provided above the first shielding shell 409, and the first shielding shell is mounted on the measuring workbench 411. The first shielding shell top cover 410 allows personnel to open the first shielding shell 409 to clean and resample the test sample liquid in the test liquid container 404. The measuring workbench 411 isolates the first shielding shell 409 from direct contact with the bottom surface, providing a stable placement platform for the measuring module 40 and isolating interference from the ground or other potentially contacting surfaces, thus improving the accuracy of the final measurement results.

[0039] In one embodiment, reference is made to Figure 1 The radioactive monitoring and control module also includes a signal conversion module, which is electrically connected to the high-purity germanium detector 406. This module converts electrical signals into digital signals. It receives electrical signals from the detector. The signal conversion module converts the analog electrical signals output by the high-purity germanium detector 406 into digital signals, improving the speed and accuracy of data processing. In practical applications, after converting analog signals to digital signals, digital signal processing techniques such as filtering, amplification, and analysis can be used to process the signals. After processing, the digital signals have less noise interference and improved measurement stability. Therefore, by converting electrical signals to digital signals, the actual state of the radioactive waste liquid can be reflected more accurately, reducing the error rate. Furthermore, the converted digital signals are easier to store and transmit, facilitating connection with computer systems or other monitoring equipment, enabling effective data recording, real-time monitoring, and data analysis. The introduction of the signal conversion module allows the entire reactor building's radioactive waste liquid measurement device to be better integrated with other monitoring and control systems, achieving more comprehensive monitoring and management.

[0040] In one embodiment, reference is made to Figure 1 and Figure 3The first quantitative sampling module 34 includes a first quantitative loop 315 and a first multi-way valve 320. The first multi-way valve 320 includes a first liquid inlet 32001, a first liquid outlet 32002, a first connection end 32006 of the first quantitative loop, a second connection end 32003 of the first quantitative loop, a first diluent injection end 32004, and a test liquid container connection end 32005. In the sampling mode, the first liquid inlet 32001, the first connection end 32006 of the first quantitative loop, the first quantitative loop 302, the second connection end 32003 of the first quantitative loop, and the first liquid outlet 32002 are connected in sequence. The first quantitative loop 302 is used for quantitative sampling of radioactive waste liquid. And / or, in the quantitative mode, the first diluent injection end 32004, the second connection end 32003 of the first quantitative loop, the first quantitative loop 315, the first connection end 32006 of the first quantitative loop, and the test liquid container connection end 32005 are connected in series. The first quantitative sampling module 34 includes a first quantitative loop 315. The quantitative loop is used to accurately measure and control the volume of the liquid, ensuring the consistency of sample volume in each sampling and reducing measurement errors caused by sample volume fluctuations. A multi-port valve is used to connect different connection ends to achieve connectivity between various functional pipelines, and the valve port switching changes the interconnection relationship between different connection ends. In this embodiment, a rotary multi-port valve is specifically used. In the rotary multi-port valve, the valve core is cylindrical or spherical. As the valve stem rotates, different faces of the valve core align with channels within the valve body, thereby achieving switching. To achieve quantitative measurement of radioactive waste liquid in the reactor building's underground pit after an accident, it is necessary to first quantitatively sample the radioactive waste liquid through the quantitative loop, and then transfer the radioactive waste liquid from the quantitative loop. Therefore, to achieve the above-mentioned quantitative sampling and treatment, at least one multi-port valve with six external ends is required. This multi-port valve is equipped with a first inlet end 32001 and a first outlet end 32002, providing channels for radioactive waste liquid sampling and transfer. The system also includes a first quantitative loop first connection end 32006 and a second quantitative loop second connection end 32003, which serve as channels for the radioactive waste liquid to enter and exit the first quantitative loop 315, respectively. It also includes a first diluent injection end 32004 and a test liquid container connection end 32005. The first diluent injection end 32004 is used to connect to a diluent source 610. In this embodiment, the diluent specifically refers to demineralized water, which is used to adjust the dilution ratio of the final sample liquid or test liquid obtained or measured. The test liquid container connection end 32005 is used to connect to a test liquid container 404, and is used to inject the quantitatively injected radioactive waste liquid from the quantitative loop into the test liquid container 404 for subsequent measurement. In this embodiment, the use of the quantitative loop to inject the radioactive waste liquid and the transfer of the quantitatively injected radioactive waste liquid sample from the quantitative loop to the test liquid container 404 are respectively referred to as the injection mode and the quantitative mode. In these two modes, the multi-port valve is in two different valve port switching states.When in sample introduction mode, the flow path of the radioactive waste liquid is: first inlet end 32001, first connection end 32006 of the first metering loop, first metering loop 315, second connection end 32003 of the first metering loop, and first outlet end 32002. This path ensures that the liquid is undisturbed after entering the metering loop, accurately maintaining the state of the introduced radioactive waste liquid. When in quantitative mode, the flow path of the desalinated water flowing from the diluent source 610 for transferring and diluting the radioactive liquid in the metering loop is: first diluent injection end 32004, second connection end 32003 of the first metering loop, first metering loop 315, first connection end 32006 of the first metering loop, and test liquid container connection end 32005. This path allows for necessary dilution of the radioactive waste liquid when it enters the test liquid container 404 to meet the concentration requirements for subsequent analysis. The reason for quantitative analysis is that the radioactivity of the analyte that the high-purity germanium detector 406 can analyze has a certain upper limit. If the waste liquid is not quantitatively analyzed, the radioactivity of the analyte will be much higher than the detection limit of the high-purity germanium detector, making it impossible for the high-purity germanium detector 406 to measure normally.

[0041] This invention also provides a system for sampling and measuring radioactive liquids in a reactor building, referring to... Figures 1 to 3The system includes: a radioactive waste liquid measurement unit, comprising the aforementioned reactor building radioactive waste liquid measurement device; a liquid inlet unit, used to allow radioactive waste liquid in the reactor building sump to pass through the containment vessel and enter the sampling and measurement system; a liquid outlet unit, connected to the radioactive waste liquid measurement unit, used to discharge the liquid in the sampling and measurement system through the containment vessel into the reactor building sump; a radioactive waste liquid sampling unit 31, connected to the liquid inlet unit and the radioactive waste liquid measurement unit, used to sample the radioactive waste liquid; and a control unit, electrically connected to the radioactive measurement unit, the radioactive waste liquid sampling unit 31, the liquid inlet unit, and the liquid outlet unit, used to control the operation of the sampling and measurement system. Simultaneously with the radioactive waste liquid measurement, sampling and retention are also required. Further measurement and analysis of the radioactive activity of the radioactive waste liquid will then be conducted in the laboratory using a gamma-ray spectrometer. The radioactive waste liquid measurement unit, including the reactor building radioactive waste liquid measurement device, can monitor the radioactive concentration of the waste liquid in real time. Only when the concentration remains within a safe range and poses no harm to personnel will the sampled liquid be manually taken and analyzed in a laboratory. Outside the reactor building, a containment vessel is installed to effectively contain and prevent radioactive materials from leaking into the environment in the event of a reactor accident or malfunction. It is typically constructed of robust materials capable of withstanding high pressure and high temperatures, ensuring it remains sealed even under extreme conditions. Therefore, to enable the reactor building radioactive liquid sampling and measurement system to obtain radioactive waste liquid from the reactor building sump after an accident, and to ensure the safe and effective introduction of radioactive liquid into the system without compromising the protective performance of the containment vessel, a liquid inlet unit is required. Simultaneously, since the reactor building radioactive waste liquid sampling and measurement system itself also requires a closed system to prevent the radioactive waste liquid entering the sampling system from leaking outside the containment vessel and causing radiation hazards, a liquid outlet unit is also needed. This unit drains the liquid from the sampling system through the containment vessel into the reactor building sump. The radioactive waste sampling unit 31 is used to sample radioactive waste from the reactor building sump. A quantitative sampling device obtains a specific volume of liquid sample, which is then diluted with a diluent to reduce radioactivity. The radioactive waste sampling unit 31 aims to ensure the representativeness and accuracy of the samples for subsequent analysis and testing. The control unit is the core of the entire sampling and measurement system, responsible for coordinating and controlling the operation of each unit. It communicates with the radioactivity measurement unit, the radioactive waste sampling unit 31, the liquid inlet unit, and the liquid outlet unit via electrical connections, enabling automated control, real-time monitoring, and data recording to ensure the safe and stable operation of the system.In the reactor building radioactive liquid sampling and measurement system of the present invention, radioactive waste liquid is introduced into the sampling and measurement system through the liquid inlet unit. The control unit instructs the radioactive waste liquid sampling unit 31 to take samples and injects the waste liquid into the radioactive waste liquid measurement unit, while simultaneously measuring the radioactivity level. After sampling is completed, the measured and sampled liquid is safely discharged back into the reactor building sump through the liquid outlet unit. In order to effectively control each unit, module, and component in the reactor building radioactive liquid sampling and measurement system to accurately realize the function of the sampling and measurement system, a measurement system control platform 70 is also provided. The measurement system control platform 70 is electrically connected to each electronic control component, receives feedback from each component for automatic control, and also actively intervenes and controls according to human input commands.

[0042] In one embodiment, reference is made to Figure 1 and Figure 2The radioactive waste sampling unit 31 includes a second quantitative injection module 32 and a sampling module 33. The sampling module 33 includes a collection container 308, a second shielding cover 309, and a second liquid level monitoring module 319. The collection container 308 is located inside the second shielding cover 309, and the second liquid level monitoring module 319 is located around the collection container 308. The second quantitative injection module is located outside the second shielding cover 309. The second quantitative injection module 32 is connected to the input end of the collection container 308, and the output end of the collection container 308 is connected to the first quantitative injection module 34. The second shielding cover 309 is used to shield the gamma rays released by the radioactive waste liquid inside the collection container 308. The radioactive waste sampling unit 31 is required to perform both quantitative injection and sampling of the radioactive waste liquid, hence it includes a second quantitative injection module 32 and a sampling module 33. The second quantitative sampling module 32 is used to quantitatively introduce radioactive waste liquid from the reactor building sump into the sample collection container 308. It ensures that the volume of radioactive liquid entering the sampling module 33 is controlled during the sampling process to obtain an accurate sample volume. The sampling module 33 includes the sample collection container 308 for collecting radioactive waste liquid samples extracted from the reactor building. The sample collection container 308 must meet safety standards for corrosion resistance and radiation resistance. The sampling module 33 also includes a second shielding cover 309 surrounding the sample collection container 308. The second shielding cover 309 is mainly used to shield the gamma rays emitted by the radioactive waste liquid inside the collection container 308, protecting the radiation safety of operators and the surrounding environment. The material and thickness of the second shielding cover 309 must be sufficient to shield the gamma rays emitted by the radioactive waste liquid inside the collection container 308 to ensure radiation safety. It is typically made of high-density materials, such as lead or concrete, to effectively shield gamma-ray radiation. The second liquid level monitoring module 319 is used to monitor the liquid level in the collection container 308 in real time to ensure that the liquid does not overflow or become insufficient during the sampling process. In practical use, liquid level monitoring can be achieved through photoelectric liquid level sensors, laser liquid level sensors, or ultrasonic liquid level sensors to provide timely feedback on the status of the collection container 308. In the embodiment of the present invention, the second quantitative sampling module 32 is connected to the input end of the collection container 308 and is responsible for quantitatively introducing radioactive waste liquid. The output end of the collection container 308 is connected to the first liquid inlet end 32001 of the first quantitative sampling module 34, which is used to transfer the liquid sample to the first sample tube to prepare for measurement by the reactor building radioactive waste liquid measuring device.

[0043] Furthermore, referring to Figure 1 and Figure 2The second quantitative sampling module 32 includes a second quantitative loop 302 and a second multi-way valve 301. The second multi-way valve 301 includes a second liquid inlet 30101, a second liquid outlet 30102, a first connection end 30106 of the second quantitative loop, a second connection end 30103 of the second quantitative loop, a second diluent injection end 30104, and a collection container connection end 30105. The second quantitative loop 302 is used for quantitative sampling of radioactive waste liquid. In the sampling mode, the second liquid inlet 30101, the first connection end 30106 of the second quantitative loop, the second quantitative loop 302, the second connection end 30103 of the second quantitative loop, and the second liquid outlet 30102 are connected in sequence. And / or, in the quantitative mode, the second diluent injection end 30104, the second connection end 30103 of the second quantitative loop, the second quantitative loop 302, the first connection end 30106 of the second quantitative loop, and the collection container connection end 30105 are connected in series. Similar to the first quantitative sampling module 34, the second quantitative sampling module 32 also includes a quantitative loop and a multi-way valve, namely a second quantitative loop 302 and a second multi-way valve 301. The second multi-way valve 301 is provided with a second inlet end 30101 and a second outlet end 30102, providing channels for the injection and transfer of radioactive waste liquid. It also includes a first connection end 30106 and a second connection end 30103 of the second quantitative loop, serving as channels for the radioactive waste liquid to enter and exit the second quantitative loop 302, respectively. It also includes a second diluent injection end 30104 and a collection container connection end 30105. The second diluent injection end 30104 is used to connect a dilution source. Similar to the first quantitative sampling module 34, the diluent specifically refers to demineralized water, which is used to adjust the dilution ratio of the final sample liquid or the liquid being measured. The collection container connection end 30105 is used to connect to the collection container 308, and is used to inject the quantitative radioactive waste liquid transferred from the second quantitative loop 302 into the collection container 308 for sampling. In this embodiment, the use of the quantitative loop to sample the radioactive waste liquid and the transfer of the quantitatively sampled radioactive waste liquid sample from the quantitative loop to the collection container 308 are respectively referred to as the sampling mode and the quantitative mode. In these two modes, the multi-port valve 301 is in two different valve port switching states. When in the sampling mode, the flow path of the radioactive waste liquid is the second inlet end 30101, the first connection end 30106 of the second quantitative loop, the second quantitative loop 302, the second connection end 30103 of the second quantitative loop, and the second outlet end 30102. Through this path, it can be ensured that the liquid in the reactor building sump is not disturbed after entering the quantitative loop, and the state of the sampled radioactive liquid is accurately maintained. When in quantitative mode, the flow path of the demineralized water flowing out of the diluent source 610 for transferring and diluting the radioactive liquid in the quantitative ring is: second diluent injection end 30104, second quantitative ring second connection end 30103, second quantitative ring 302, second quantitative ring first connection end 30106, and collection container connection end 30105.This pathway allows for necessary dilution of the radioactive waste liquid upon entry into collection container 308 to meet sampling concentration requirements. Dilution is necessary because the radioactive concentration of the waste liquid may be very high, potentially posing a radiation hazard to personnel transporting and collecting samples. Furthermore, excessively high radioactivity may prevent measurement and analysis by the gamma spectrometer in the laboratory, necessitating secondary dilution. This process carries the risk of radioactive waste liquid leakage, potentially leading to radiation safety accidents such as leakage and spread of radioactive liquid. Therefore, during sampling, the radioactive waste liquid is directly diluted to a certain proportion using demineralized water. The second quantitative sampling module 32, through a flexible multi-port valve, can quickly switch between sampling and quantitative modes, ensuring effective sampling and treatment of the radioactive waste liquid.

[0044] Furthermore, referring to Figures 1 to 3 The second quantitative loop 302 has a volume of 0.1 ml, the first quantitative loop 315 has a volume of 1 mL, the collection container 308 has a volume of 1000 mL, and the test liquid container 404 has a volume of 100 ml. The radioactive waste liquid in the reactor building sump is first quantitatively injected through the second quantitative loop 302, then enters the collection container 308 and is diluted, then is drawn by the second electric sampling pump 311, flows through the first quantitative loop 315, and finally enters the test liquid container 404. During this process, the solution in the test liquid container 404 undergoes two stages of dilution, resulting in a lower radioactivity. Therefore, even when the radioactivity concentration of the radioactive waste liquid in the reactor building sump is very high, the two-stage dilution ensures that the radioactivity of the diluted sample entering the test liquid container 404 is within the measurement range of the high-purity germanium detector 406, thus avoiding measurement failure due to exceeding the measurement and analysis range.

[0045] In one embodiment, reference is made to Figure 1The liquid discharge unit includes a spray module 50 and a liquid discharge module 11. The output end of the spray module 50 is connected to the liquid discharge module 11, and the input end of the spray module 50 is connected to the first quantitative injection module 34 and the second quantitative injection module 32. The liquid discharge module 11 is used to discharge the liquid in the sampling system through the containment vessel into the reactor building pit, and the spray module 50 is used to pressurize the liquid to be discharged from the sampling system. In order to allow the liquid in the reactor building radioactive liquid sampling and measurement system of the present invention to be discharged into the reactor building sump to reduce the residual radioactive material in the system and overcome the pressure difference that may exist inside and outside the containment, it is necessary to set up a liquid discharge module 11 that passes through the containment and a spray module 50 for pressurizing and assisting the liquid injection in the system. The function of the spray module 50 is mainly realized by the injector 502. The injector 502 is connected to the diluent source 610. The demineralized water in the diluent source 610 exists as a pressurizing and cleaning fluid. The pipeline connecting the injector 502 and the diluent source 610 is also equipped with a No. 501 electric valve 501. This electric valve is used to prevent the demineralized water in the diluent source 610 from flowing out when the injector 502 is not working, thereby reducing waste and preventing backflow. When the liquid in the reactor building's radioactive liquid sampling and measurement system reaches ejector 502, ejector 502 mixes with demineralized water and injects it under pressure, then passes through the containment vessel from the discharge module 11 and is sprayed into the reactor building's sump. After the radioactive liquid is discharged, to prevent backflow in the pipeline, ejector 502 continues to inject a certain amount of demineralized water to fill the pipeline between the discharge module 11 and the end discharge port, and closes electric valve 503 to achieve isolation, thereby preventing backflow of radioactive waste liquid along the pipeline of discharge module 11 due to pressure changes within the reactor building. Four electric valves are also installed on the discharge module 11: electric valve 105 (No. 105), electric valve 106 (No. 106), electric valve 107 (No. 107), and electric valve 108 (No. 108). Electric valves 105 and 107 are located inside the containment vessel, while electric valves 106 and 108 are located outside the containment vessel. Electric valves 106 (No. 106) and 105 (No. 105) are connected along the liquid flow direction and form a set of flow safety control valves to control the discharge of liquid outside the containment into the reactor building sump. When the discharge module 11 needs to discharge liquid, electric valves 106 (No. 106) and 105 (No. 105) will be opened to discharge the liquid. However, in the event of a design basis accident or a severe accident, the electric valves are likely to be damaged, especially electric valve 105 (No. 105) and its connected piping located inside the containment. Furthermore, under accident conditions, it is impossible to inspect or replace components and piping inside the containment.To ensure the sampling system can continue to drain liquid into the reactor building's underground sump after an accident, electric valves 107 (No. 107) and 108 (No. 108) are installed as redundancy for electric valves 105 (No. 105) and 106 (No. 106), ensuring the normal operation of radioactive liquid sampling and measurement in the reactor building. It should be noted that the redundancy is not limited to electric valves 107 and 108. Typically, at least three sets of redundancy are installed for electric valves 106 (No. 106) and 105 (No. 105) to ensure the normal operation of the sampling system. This embodiment only shows one set of redundancy in the accompanying drawings for illustrative purposes.

[0046] In one embodiment, reference is made to Figure 1 The liquid inlet unit includes a liquid inlet module 10 and a flow control module 20. The input end of the flow control module 20 is connected to the liquid inlet module 10, and the output end of the flow control module 20 is connected to the input end of the radioactive waste liquid measurement unit and the second liquid inlet end 30101. The liquid inlet module 10 is used to allow the radioactive waste liquid in the reactor building pit to enter the sampling and measurement system, and the flow control module 20 is used to control the flow rate of the liquid in the sampling and measurement system.

[0047] Reference Figure 1To facilitate the entry of radioactive waste liquid from the reactor building's sump into the sampling and measurement system, a liquid inlet module 10 and a flow control module 20 were installed. The liquid inlet module 10 is equipped with four electrically operated valves: valve 101 (No. 101), valve 102 (No. 102), valve 103 (No. 103), and valve 104 (No. 104). Valve 101 and valve 103 are located inside the containment vessel, while valves 102 and 104 are located outside. Valve 101 and valve 102 are connected along the liquid flow direction, forming a set of flow safety control valves to control the entry of liquid from inside and outside the containment vessel. When the liquid inlet module 10 requires radioactive waste liquid to enter, valves 101 and 102 will be opened, allowing the radioactive liquid from the reactor building's sump to enter. However, in the event of a design basis accident or a severe accident, the electric valves are likely to be damaged, especially electric valve 101 (No. 101) and its connected piping located within the containment. Furthermore, under accident conditions, it is impossible to inspect or replace components and piping within the containment. To ensure that the sampling and measurement system can still normally feed liquid into the reactor building sump after an accident, electric valves 103 (No. 103) and 104 (No. 104) are provided as redundancy for electric valves 101 (No. 101) and 102 (No. 102), ensuring the normal operation of the reactor building radioactive liquid sampling and measurement system. It should be noted that the redundancy does not only include electric valves 103 (No. 103) and 104 (No. 104). Typically, at least three sets of redundancy are provided for electric valves 101 (No. 101) and 102 (No. 102) to ensure the normal operation of the sampling system. This embodiment only shows one set of redundancy in the accompanying drawings for illustrative purposes. The flow control module 20 is equipped with an online gamma dose rate meter 201. The gamma dose rate meter 201 is used to perform a preliminary measurement of the gamma dose rate caused by the radioactive waste liquid entering the sampling and measurement system, so as to provide a basis for the subsequent measurement system control platform 70 to control the sampling and measurement modes of the second quantitative injection module 32, the sampling module 33, the first quantitative injection module 34 and the measurement module 40.

[0048] Reference Figure 1Due to design specifications for the reactor building, the reactor building's sump must be positioned lower than the reactor building's radioactive liquid sampling and measurement system. Therefore, to allow radioactive waste liquid from the sump to enter the sampling and measurement system, a power source must be provided for the radioactive waste liquid. Thus, a first electric sampling pump 203 is installed in the flow control module 20. An electric valve 202 is installed at the front end of the first electric sampling pump 203 along the liquid flow direction to control the on / off state of the pipeline. A sampling cooler 204 is installed at the rear end of the first electric sampling pump 203 along the liquid flow direction. During a prolonged period following a design-basis accident or severe accident at a nuclear power plant, the reactor building environment within the containment vessel remains at a high temperature due to heat leakage from the reactor. At this time, the temperature of the radioactive waste liquid in the reactor building's sump will be very high. High-temperature liquids can affect the operational safety of the sampling and measurement system and the required operating temperature of the high-purity germanium detector 406 in the measurement module 40. Therefore, the radioactive waste liquid entering the sampling and measurement system must be cooled by the sampling cooler 204, ultimately reducing the temperature to below 30°C. In embodiments of the present invention, temperature sensors (not shown in the figure) are also provided at the front and rear ends of the sampling cooler 204 along the liquid flow direction. Based on the feedback from these temperature sensors, the measurement system control platform 70 controls the flow rate of the coolant inside the sampling cooler 204 to ensure that the temperature of the radioactive waste liquid leaving the sampling cooler 204 is below 30°C. Considering the existing liquid cooling pipeline conditions around the nuclear reactor building, in actual use, the sampling cooler 204 will be directly connected to the existing liquid cooling pipeline in the nuclear power plant and supplied with cooling water.

[0049] In one embodiment, reference is made to Figure 1The reactor building radioactive liquid sampling and measurement system of the present invention also includes a flushing module 21. The flushing module 21 is connected to the radioactive waste liquid sampling unit 31, the radioactive waste liquid measurement unit, and the discharge unit. The flushing module 21 is used to flush the sampling and measurement system. The flushing module 21 includes a diluent source 610 and a No. 206 electric valve 206 for controlling the opening and closing of the diluent source 610 pipeline. A No. 205 electric valve 205 is provided between the flow control module 20 and the flushing module 21. Through the diluent source 610, the pipeline can be flushed with demineralized water in the diluent source 610 before sampling and measurement work is performed, reducing measurement errors caused by residues in the sampling and measurement system pipeline and improving the reliability and accuracy of the measured data. In case of abnormal conditions or detection of contamination, the flushing module 21 can be quickly activated to clean and reduce risks. Specifically, the flow control module 20 is also equipped with a flow meter 207 (No. 207) to monitor the flow rate before entering the second quantitative injection module 32, preventing excessive pressure in the sampling and measurement system pipeline due to excessive flow rate, which could lead to pipe bursting and damage. When the pressure is too high, the pipeline can be quickly disconnected by controlling the pipeline opening and closing via an electric valve 205 (No. 205) located between the flow meter 207 (No. 207) and the sampling cooler 204, and the pipeline pressure can be reduced in conjunction with the demineralized water output from the diluent source 610.

[0050] In one embodiment, reference is made to Figure 1 The output of the sampling cooler 204 is also connected to the input of the ejector 502, and an electric valve 607 (No. 607) is installed between them to control the on / off state. Under certain special circumstances, when the radioactive waste liquid in the sump is at an extremely high temperature or high radiation level, sampling and measurement operations must be abandoned. In this case, after the temperature sensors at the back end of the gamma dose rate meter 201 and the sampling cooler 204 return an over-temperature or over-radiation threshold alarm, electric valve 205 (No. 205) will close, and electric valve 607 (No. 607) will open, allowing the radioactive waste liquid to be directly discharged back into the sump of the reactor building through the ejector module 50 and the discharge module 11.

[0051] In one embodiment, reference is made to Figure 7The measurement system control platform 70 includes an electric valve controller 701, a sampling pump controller 702, a liquid level monitoring module controller 703, a gamma-ray spectrometer main amplifier 704, a multichannel analyzer 705, gamma-ray spectrometer analysis software 706, and a high-purity germanium spectrometer controller 707. The electric valve controller 701 controls the opening and closing of the electric valve, regulating the flow rate of the radioactive waste liquid. This ensures that the liquid flow is controllable during sampling or measurement. The sampling pump controller 702 controls the operation of the sampling pump, precisely adjusting its working state according to the required sampling volume and speed. The monitoring module controller 703 monitors and displays real-time changes in various indicators from various monitors, such as the gamma dose rate meter 201, the liquid level monitor, the temperature and humidity controller 408, and the temperature sensor. The gamma-ray spectrometer main amplifier 704 enhances the signal strength, enabling clearer results in subsequent analysis and processing. The multichannel analyzer 705 is used to digitize the amplified signal, separate gamma-ray signals of different energies, and convert them into analyzable data. The gamma-ray energy spectrum analysis software 706 further processes and analyzes the data output from the multichannel analyzer 705, generating energy spectra and extracting relevant radionuclide information and concentration data. The high-purity germanium spectrometer controller 707 is used to control the operation of the high-purity germanium spectrometer, including energy calibration and temperature control, to ensure the accuracy and stability of the measurements.

[0052] Reference Figure 1 To further enhance the functionality of the reactor building's radioactive waste sampling and measurement system, auxiliary components 60 are required, including electric valves 601-602, 603-604, 605-606, 607-608. These electric valves are controlled by the measurement system control platform 70, managing the on / off state of their respective connecting pipelines, and function as electrically operated isolation valves. Auxiliary components 60 also include the aforementioned diluent source 610. The diluent source 610 can be a single component or multiple components can be installed depending on the pipeline layout. It serves both to output demineralized water to dilute the radioactive waste and to flush the sampling and measurement system pipelines.

[0053] This invention also provides a method for sampling and measuring radioactive waste liquid in reactor buildings, applied to the aforementioned radioactive liquid sampling and measurement system in reactor buildings, with reference to... Figures 4 to 6 It includes:

[0054] S801. Open the liquid inlet unit and liquid outlet unit to allow the radioactive waste liquid in the reactor building pit to enter the reactor building radioactive liquid sampling and measurement system.

[0055] The measurement system control platform 70 controls the start of the liquid inlet unit and simultaneously activates the liquid outlet unit to ensure liquid flow within the system, facilitating subsequent gamma dose rate measurements. At this point, radioactive waste liquid enters the system.

[0056] S802. Obtain the gamma dose caused by the radioactive waste liquid in the reactor building pit through the gamma dose rate meter 201, and select the sampling mode according to the gamma dose caused by the radioactive waste liquid in the reactor building pit.

[0057] The gamma dose rate meter 201 installed in the liquid inlet unit is used to acquire gamma dose rate data of radioactive waste liquid in real time. The measured gamma dose rate value is recorded by the measurement system control platform 70 to provide a basis for subsequent selection of sampling mode.

[0058] S803. Start the radioactive waste liquid sampling unit 31 and sample the radioactive waste liquid according to the sampling mode;

[0059] Based on the acquired gamma dose rate data, the radioactivity level of the radioactive waste liquid is determined according to the preset threshold or scheme, thereby selecting a suitable sampling mode (such as high radioactivity concentration mode, low radioactivity concentration mode, etc.).

[0060] In one embodiment, the steps of activating the radioactive waste sampling unit 31 and sampling the radioactive waste according to the sampling mode include:

[0061] S8031. Start radioactive waste liquid sampling unit 31;

[0062] S8032a When the γ dose is greater than the first preset value, the first sampling mode is used, and the radioactive waste liquid sampling unit 31 dilutes and samples the radioactive waste liquid according to the first preset dilution ratio of the first sampling mode.

[0063] S8032b When the γ dose is less than the first preset value and greater than the second preset value, the second sampling mode is used, and the radioactive waste liquid sampling unit 31 dilutes and samples the radioactive waste liquid according to the second preset dilution ratio of the second sampling mode.

[0064] S8032c When the γ dose is less than the second preset value, the third sampling mode is used, and the radioactive waste liquid sampling unit 31 dilutes and samples the radioactive waste liquid according to the third preset dilution ratio of the third sampling mode.

[0065] Based on the comparison between the measured gamma dose value and the first and second preset values, the corresponding sampling mode is selected and executed. When the gamma dose is greater than the first preset value, the radioactive waste sampling unit 31 dilutes the radioactive waste according to the first preset dilution ratio set in the first sampling mode. When the gamma dose is less than the first preset value but greater than the second preset value, the radioactive waste sampling unit 31 dilutes and samples according to the second preset dilution ratio set in the second sampling mode. When the gamma dose is less than the second preset value, the radioactive waste sampling unit 31 dilutes and samples according to the third preset dilution ratio set in the third sampling mode.

[0066] S804. Start the radioactive waste liquid measurement unit and measure the types of radionuclides and their radioactive concentrations in the radioactive waste liquid according to the sampling mode.

[0067] The radioactive waste liquid measurement unit is activated, and the radioactive concentration of the sampled waste liquid is measured according to the selected sampling mode. The measured nuclide types and their radioactive concentration data are recorded through the measurement system control platform 70, providing a basis for subsequent analysis and processing.

[0068] In one embodiment, the step of activating the radioactive waste liquid measurement unit and measuring the radioactive concentration of the radioactive waste liquid according to the sampling mode includes:

[0069] S8041. Inject the radioactive waste liquid into the test liquid container 404, and measure the radionuclide types and radioactivity information of the radioactive waste liquid through the radioactivity measurement and control module.

[0070] S8042a. When the γ dose is greater than the first preset value, the concentration of each radionuclide in the waste liquid in the reactor building pit is calculated based on the radioactivity information according to the first preset formula of the first sampling mode.

[0071] S8042b: When the γ dose is less than the first preset value and greater than the second preset value, the concentration of each radionuclide in the waste liquid in the reactor building pit is calculated based on the radioactivity information according to the second preset formula of the second sampling mode.

[0072] S8042c: When the γ dose is less than the second preset value, the concentration of each radionuclide in the waste liquid in the reactor building pit is calculated based on the radioactivity information according to the third preset formula of the third sampling mode.

[0073] Diluted radioactive waste liquid is injected into the test liquid container 404. The injected radioactive waste liquid is measured using a radioactivity monitoring module. Based on the measured gamma dose and radioactivity activity information, different formulas are used to calculate the concentration of each radionuclide in the reactor building sump waste liquid. When the gamma dose is greater than a first preset value, the concentration of each radionuclide in the reactor building sump waste liquid is calculated using the first preset formula of the first sampling mode and the radioactivity information. When the gamma dose is less than the first preset value but greater than a second preset value, the concentration of each radionuclide in the reactor building sump waste liquid is calculated using the second preset formula of the second sampling mode, combined with the radioactivity information. When the gamma dose is less than the second preset value, the concentration of each radionuclide in the reactor building sump waste liquid is calculated using the third preset formula of the third sampling mode, based on the radioactivity information. The formulas used need to consider the radioactivity activity of each radionuclide in the waste liquid and the influence of environmental conditions on the measurement results. Depending on the changes in gamma dose, the formulas also need to be adjusted according to the characteristics of different gamma doses to improve measurement accuracy.

[0074] S805. Close the liquid inlet unit and the liquid outlet unit, and take sample liquid from the radioactive waste liquid sampling unit 31.

[0075] After sampling and measurement are completed, the inlet and outlet units are closed to stop the flow of waste liquid, and the pipeline is flushed using the flushing module 21. After closing the inlet and outlet units, the sampled liquid is removed from the radioactive waste liquid sampling unit 31.

[0076] To more clearly illustrate the reactor building atmosphere sampling system and method of the present invention, refer to 1 to 2000. Figure 9 The following example illustrates a complete post-accident reactor building sump waste liquid sampling and measurement procedure:

[0077] During normal operation of a nuclear power plant, the waste liquid sampling and measurement system in the reactor building sump after an accident is shut down. During a nuclear power plant shutdown and overhaul, the waste liquid sampling and measurement system in the reactor building sump after an accident is activated to inspect and test the equipment status and operational reliability. If any components fail during the inspection and testing, they are replaced and repaired in a timely manner.

[0078] Within 24 hours of a nuclear power plant accident, waste liquid samples must be taken from the sump and analyzed for radioactivity to provide radiation data support for the nuclear power plant emergency command center to conduct an accident progress assessment.

[0079] In the event of a large breach and loss of water or a serious accident at a nuclear power plant, staff immediately remotely activate the high-purity germanium detector 406, cooler 407, and temperature and humidity controller 408 of the measurement module 40 via the high-purity germanium spectrometer controller 707 on the measurement system control platform 70 to prepare for the radioactivity measurement of the sample.

[0080] In the event of a large-break loss-of-coolant accident or a severe accident, a large amount of radioactive material released from the primary coolant or due to core damage is collected in the reactor building's sump. During accident conditions, the sampling and measurement of highly radioactive waste in the reactor building's sump is conducted on the waste liquid within the reactor building's sump.

[0081] The sampling and measurement methods for high-level radioactive waste liquid in the reactor building under accident conditions are as follows. Before sampling and measuring the waste liquid in the reactor building sump, the staff remotely controlled and confirmed that all sampling pumps and electric valves (including electric flow regulating valves and electric isolation valves) of the high-level radioactive waste liquid sampling and measurement system in the reactor building were in the closed state through the electric valve controller 701 and sampling pump controller 702 of the measurement system control platform 70.

[0082] The electric valves controlling the liquid inlet module 10 and the liquid outlet module 11 are opened. After the nuclear power plant emergency command center issues an instruction to sample and measure the waste liquid in the reactor building sump, the operator in the main control room remotely controls the opening of electric valves 101 (No. 101), 102 (No. 102), 105 (No. 105), and 106 (No. 106) of the waste liquid sampling pipeline in the sump to open. If any one of these electric valves fails to open properly, the operator in the main control room will immediately receive an indication signal that the corresponding electric isolation valve is not open. When an indication signal indicating that electric valve 101 or electric valve 102 is not open is received, the operator remotely controls electric valve 103 or electric valve 104 to open; similarly, when an indication signal indicating that electric valve 105 or electric valve 106 is not open is received, the operator remotely controls electric valve 107 or electric valve 108 to open.

[0083] Waste liquid gamma dose rate measurement. After the operator in the nuclear power plant's main control room remotely controls the opening of the electric valves of the inlet module 10 and outlet module 11 of the waste liquid sampling pipeline in the reactor building, relevant personnel are notified via the nuclear power plant's internal network, plant broadcast, or telephone to carry out waste liquid sampling and measurement work in reactor building sump a. After receiving the instruction from the main control room, the staff initiates the waste liquid sampling and measurement work.

[0084] The staff sequentially opened electric valve 202, first electric sampling pump 203, and sampling cooler 204 through the electric valve controller 701 and sampling pump controller 702 of the measurement system control platform 70, opened electric valve 607, and remotely opened electric valves 501 and 503 of the injection module 50. The first electric sampling pump 203 extracts waste liquid from reactor building pit a. The waste liquid is injected into reactor building pit b through electric valves 101 and 102 or 103 and 104, electric valve 202, first electric sampling pump 203, sampling cooler 204, electric valve 607, injector 502 and electric valve 503 of injection module 50, electric valve 105 and electric valve 106 or electric valve 107 and electric valve 108.

[0085] The gamma dose rate meter 201 measures the gamma dose rate caused by the radioactive liquid in the sampling pipeline and transmits the measurement results in real time to the monitoring module controller 703 of the measurement system control platform 70 via a measurement cable, and displays the measurement results of the gamma dose rate meter 201 in real time.

[0086] Sampling of radioactive waste liquid from the crater following the accident. Workers viewed the measurement data from the gamma dose rate meter 201 via the data display console of the monitoring module controller 703 on the measurement system control platform 70. After the gamma dose rate meter 201's measurement data stabilized, workers analyzed the data... Select appropriate methods for quantifying, diluting, and sampling the waste liquid in the reactor building pit a.

[0087] when At this time, sampling mode 1 is used. After the measurement data of the gamma dose rate meter 201 stabilizes, when the measurement data of the gamma dose rate meter 201... When sampling is performed, sampling is carried out according to sampling mode 1.

[0088] The staff remotely controls the following electric isolation valves sequentially via the electric valve controller 701 on the measurement system control platform 70: electric valve 316 (No. 316), electric valve 303 (No. 303), electric valve 604 (No. 604), electric valve 603 (No. 603), electric valve 605 (No. 605), electric valve 306 (No. 306), electric valve 317 (No. 317), electric valve 304 (No. 304), electric valve 312 (No. 312), electric valve 601 (No. 601), and electric valve 206 (No. 206). The second electric sampling pump 311 is remotely activated via the sampling pump controller 702. Flowmeter 314 (No. 314) transmits its measurement data to the second electric sampling pump 311 in real time via electrical transmission. The second electric sampling pump 311 automatically adjusts its sampling flow rate based on the feedback data from flowmeter 314, ensuring that the sampling flow rate of the second electric sampling pump 311 is stably controlled at 100 mL / min. Flowmeters 305 and 318 transmit their measurement data to electric valves 304 and 317 via electrical transmission. Based on the measurement data, electric valves 304 and 317 automatically adjust their valve openings to ensure that the flow rate of demineralized water entering the second diluent injection end 30104 of the second multi-way valve 301 and the first diluent injection end 32004 of the first multi-way valve 320 is controlled at 100 mL / min.

[0089] The sample inlet circuit is flushed. Demineralized water enters the second multi-way valve 301 via electric valve 206 (No. 206), flow meter 207 (No. 207), and electric valve 601 (No. 601). It then enters the waste liquid injection module 50 via the second inlet 30101, the first connection end of the second metering loop 30106, the first metering loop (0.1 mL) 302, the second connection end of the second metering loop 30103, the port of the second outlet 30102, and electric valve 303 (No. 303). The injector 502 of the injection module 50 draws the waste liquid discharged from the second outlet 30102 and injects it into the reactor building pit b via electric valve 503 (No. 503), electric valve 105 (No. 105), and electric valve 106 (No. 106), or electric valve 107 (No. 107) and electric valve 108 (No. 108), forming a flushing circuit.

[0090] The demineralized water enters the second multi-way valve 301 via electric valve 304 (No. 304) and flow meter 305 (No. 305), then enters the collection container 308 via the second diluent injection end 30104, the collection container connection end 30105, electric valve 306 (No. 306), and the second quick connector 307. The demineralized water entering the collection container 308 enters the first multi-way valve 320 via the third quick connector 310, the second electric sampling pump 311, electric valve 312 (No. 312), and flow meter 314 (No. 314), then enters the waste liquid injection and spray module 50 via the first inlet end 32001, the first connection end 32006 of the first metering ring, the 1mL first metering ring 315, the second connection end 32003 of the first metering ring, the first outlet end 32002, and electric valve 316 (No. 316). The ejector 502 of the waste injection module 50 draws the waste liquid discharged from the first outlet end 32002, and injects it into the reactor building pit b through electric valves 503, 105, and 106 or 107 and 108, forming a flushing circuit.

[0091] Demineralized water enters the first multi-way valve 320 via electric valve 317 (No. 317) and flow meter 318 (No. 318), then enters the injection module 50 via the first diluent injection end 32004, the test liquid container connection end 32005, electric valve 604 (No. 604), electric valve 603 (No. 603), and electric valve 605 (No. 605). The injector 502 of the injection module 50 draws the test liquid from the test liquid container connection end 32005, and injects it into the reactor building pit b via electric valve 503 (No. 503), electric valve 105 (No. 105), and electric valve 106 (No. 106), or electric valve 107 (No. 107) and electric valve 108 (No. 108), forming a flushing loop.

[0092] Demineralized water is used to flush any residual radioactive waste or radionuclide contamination on the pipe walls in the above flushing circuit at a flow rate of 100 mL / min for 5 minutes. The flushing fluid enters the injection module 50. The injector 502 draws in, injects into, and discharges the waste liquid generated during the above flushing into the reactor building pit b.

[0093] Sample injection, dilution, and quantification are performed. After flushing the above-mentioned flushing circuit with deionized water at a flow rate of 100 mL / min for 5 minutes, the operator remotely controls the electric isolation valves 206 and 607 to close and open the electric valve 205 via the electric valve controller 701 of the measurement system control platform 70. The flow meter 207 transmits its measurement data to the first electric sampling pump 203 in real time via electrical transmission. The first electric sampling pump 203 automatically adjusts its sampling flow rate based on the measurement data fed back by the flow meter 207, ensuring that the sampling flow rate of the first electric sampling pump 203 is stably controlled at 100 mL / min, thus stabilizing the flow rate of the sample entering the second inlet 30101 at 100 mL / min.

[0094] like Figure 2 As shown, the first electric sampling pump 203 extracts waste liquid from the reactor building pit a. The waste liquid passes through electric valves 101 and 102 or electric valves 103 and 104, electric valve 202, first electric sampling pump 203, sampling cooler 204, electric valve 205, flow meter 207, and electric valve 601 to enter the second multi-way valve 301. It then enters the injection module 50 through the second inlet end 30101, the first connection end 30106 of the second metering ring, the 0.1mL first metering ring 302, the second connection end 30103 of the second metering ring, the second outlet end 30102, and electric valve 303, forming a sampling circuit.

[0095] At this time, the second multi-way valve 301 is in injection mode, with an injection flow rate of 100 mL / min and an injection duration of 5 min, ensuring that the radioactivity concentration of the liquid at all points in this injection loop is the same as that of the waste liquid in reactor building sump a. This completes the injection of 0.1 mL of waste liquid from reactor building sump a into the 0.1 mL second metering loop 302 of the second multi-way valve 301. At this point, the radioactivity concentration of the liquid within the 0.1 mL second metering loop 302 is the same as that of the waste liquid in reactor building sump a.

[0096] After the second multi-way valve 301 operates in sample injection mode for 5 minutes, the operator remotely controls the second multi-way valve 301 to switch from sample injection mode to quantitative mode via the electric valve controller 701 of the measurement system control platform 70. At the same time, the operator remotely shuts down the second electric sampling pump 311 via the sampling pump controller 702. In quantitative mode, demineralized water enters the second multi-way valve 301 via electric valve 304 (No. 304) and flow meter 305 (No. 305). It then enters the collection container 308 via the second diluent injection end 30104, the second connection end 30103 of the second quantitative loop, the 0.1mL first quantitative loop 302, the first connection end 30106 of the second quantitative loop, the collection container connection end 30105, electric valve 306 (No. 306), and the second quick connector 307, forming a quantitative sampling waste liquid dilution loop. Flow meter 305 transmits its measurement data to electric valve 304 via electrical transmission. Based on the measurement data from flow meter 305, electric valve 304 automatically adjusts the valve opening to ensure that the flow rate of demineralized water entering the second diluent injection end 30104 is controlled at 100 mL / min.

[0097] The second liquid level monitoring module 319 transmits the liquid level data of the collection container 308 to the monitoring module controller 703 of the measurement system control platform 70 in real time via a measuring cable. When the liquid level of the collection container 308 reaches 1000mL, the monitoring module controller 703 sends an alarm signal to the electric valve controller 701. The electric valve controller 701 automatically closes the electric valves 304 and 306 to ensure that the liquid level of the collection container 308 is 1000mL.

[0098] The flow rate of demineralized water entering the second multi-way valve 301 is 100 mL / min. 0.1 mL of waste liquid with the same radioactivity concentration as the waste liquid in reactor building sump a is flushed into the collection container 308 over a period of 10 minutes, thus diluting the 0.1 mL sample of waste liquid from reactor building sump a. At this point, the radioactivity concentration of the liquid in the collection container 308 is one ten-thousandth of the radioactivity concentration of the waste liquid from reactor building sump a.

[0099] After diluting the 0.1 mL sample, the operator adjusts the injection mode of the first multi-way valve 320 via the electric valve controller 701 of the measurement system control platform 70, and remotely controls the second electric sampling pump 311 to start via the sampling pump controller 702. The liquid in the collection container 308 enters the first multi-way valve 320 via the third quick connector 310, the second electric sampling pump 311, the No. 312 electric valve, and the No. 314 flow meter. It then enters the injection module 50 via the first inlet end 32001, the first connection end 32006 of the first metering loop, the 1 mL first metering loop 315, the second connection end 32003 of the first metering loop, the first outlet end 32002, and the No. 316 electric valve, forming an injection loop. If the demineralized water is insufficient, it can also be injected via the diluent source 610 through the No. 313 electric valve.

[0100] Flowmeter 314 transmits its measurement data to the second electric sampling pump 311 in real time via electrical transmission. The second electric sampling pump 311 automatically adjusts its sampling flow rate based on the feedback data from flowmeter 314, ensuring that the sampling flow rate of the second electric sampling pump 311 is stably controlled at 100 mL / min and the injection flow rate of the first multi-way valve 320 is also 100 mL / min, with an injection duration of 5 minutes. This ensures that the radioactivity concentration of the liquid at all points in the injection loop is the same as that of the liquid in collection container 308. This completes the injection of 1 mL of liquid from collection container 308 into the first multi-way valve 320. At this point, the radioactivity concentration of the liquid at all points in the injection loop is the same as that of the liquid in collection container 308, which is one ten-thousandth of the radioactivity concentration of the waste liquid in reactor building sump a.

[0101] After the 1 mL first quantitative loop 315 of the first multi-way valve 320 is injected, the staff remotely controls the 603 electric valve and the 316 electric valve to close via the electric valve controller 701 of the measurement system control platform 70, thereby changing the first multi-way valve 320 from the injection mode to the quantitative mode, opening the 401 electric valve, and simultaneously remotely closing the second electric sampling pump 311 via the sampling pump controller 702.

[0102] like Figure 2As shown, the demineralized water enters the first multi-way valve 320 via the No. 317 electric valve 317, the No. 318 flow meter 318, the first diluent injection end 32004, the second connection end of the first metering ring 32003, the 1mL first metering ring 315, the first connection end of the first metering ring 32006 and the connection end of the test liquid container 32005, the No. 604 electric valve 604 and the No. 401 electric valve 401, and the first quick connector 403, and then enters the test liquid container 404. The flow meter 318 transmits its measurement data to the electric valve 317 via electrical transmission. Based on the measurement data from the flow meter 318, the electric valve 317 automatically adjusts the valve opening to ensure that the demineralized water enters the first multi-way valve 320 through the electric valve 317 and flushes 1 mL of liquid (with a radioactive concentration of one ten-thousandth of the waste liquid in the reactor building pit a) from the first metering loop 315 to the test liquid container 404 at a flow rate of 100 mL / min. The flushing time is 1 minute.

[0103] like Figure 3 As shown, the first liquid level monitoring module 405 transmits the liquid level data of the test liquid container 404 to the monitoring module controller 703 of the measurement system control platform 70 in real time via a measuring cable. When the liquid level of the test liquid container 404 reaches 100mL, the monitoring module controller 703 issues an alarm indicating that the test liquid container 404 has reached the measurement condition. The alarm lasts for 30 seconds and then automatically shuts off. At the same time, it sends an alarm signal to the electric valve controller 701, which automatically closes electric valves 317 (No. 317) and 401 (No. 401) to ensure that the liquid level of the test liquid container 404 is 100mL. At this time, the radioactivity concentration of the liquid in the test liquid container 404 is one part per million of the waste liquid in the reactor building sump a.

[0104] Sampling mode 2 is used when 5 μSv / h <·D ≤ 5 mSv / h. When the measurement data of the γ dose rate meter 201 is 5 μSv / h <·D ≤ 5 mSv / h, sampling is performed according to mode 2.

[0105] The staff remotely controlled the electric valve controller 701 of the measurement system control platform 70 to sequentially open electric valves 316, 604, 603, 605, 317, 312, 602, 606, and 206, and remotely opened the second electric sampling pump 311 through the sampling pump controller 702.

[0106] Flowmeter 314 (No. 314) transmits its measurement data to the second electric sampling pump 311 in real time via electrical transmission. The second electric sampling pump 311 automatically adjusts its sampling flow rate based on the feedback data from flowmeter 314, ensuring that the sampling flow rate is stably controlled at 100 mL / min. Flowmeter 318 (No. 318) feeds back its measurement data to electric valve 317 (No. 317) via electrical transmission. Based on the measurement data from flowmeter 318, electric valve 317 automatically adjusts its valve opening to ensure that the flow rate of demineralized water entering the first diluent injection end 32004 is controlled at 100 mL / min.

[0107] Flushing of the sample inlet circuit. Demineralized water enters the collection container 308 via electric valve 206, flow meter 207, electric valve 602 and electric valve 606, and second quick connector 307. The liquid in the collection container 308 enters the first multi-way valve 320 via the third quick connector 310, second electric sampling pump 311, electric valve 312, and flow meter 314. It then enters the spray module 50 via the first inlet end 32001, the first connection end 32006 of the first metering loop, the 1mL first metering loop 315, the second connection end 32003 of the first metering loop, the first outlet end 32002, and electric valve 316. The injector 502 of the injection module 50 draws the waste liquid discharged from the first liquid outlet 32002, and injects it into the reactor building pit b through electric valves 503, 105, and 106 or 107 and 108, forming a flushing circuit.

[0108] Demineralized water enters the first multi-way valve 320 via electric valve 317 (No. 317) and flow meter 318 (No. 318). It then enters the injector 502 of the injection module 50 via the first diluent injection end 32004, the test liquid container connection end 32005, electric valve 604 (No. 604), electric valve 603 (No. 603), and electric valve 605 (No. 605) to draw the waste liquid discharged from the test liquid container connection end 32005. The waste liquid is then injected into the reactor building pit b via electric valve 503 (No. 503), electric valve 105 (No. 105), and electric valve 106 (No. 106) or electric valve 107 (No. 107) and electric valve 108 (No. 108), forming a flushing loop.

[0109] Demineralized water is used to flush any residual radioactive waste or radionuclide contamination on the pipe walls in the above flushing circuit at a flow rate of 100 mL / min for 5 minutes. The flushing fluid enters the injection module 50. The injector 502 draws in, injects into, and discharges the waste liquid generated during the above flushing into the reactor building pit b.

[0110] Sample injection and quantification: After flushing the above-mentioned flushing circuit with demineralized water at a flow rate of 100 mL / min for 5 minutes, the operator remotely controls the closing of electric valves 206 (No. 206) and 607 (No. 607) and the opening of electric isolation valve 205 (No. 205) via the electric valve controller 701 of the measurement system control platform 70. Flowmeter 207 (No. 207) transmits its measurement data to the first electric sampling pump 203 in real time via electrical transmission. The first electric sampling pump 203 automatically adjusts its sampling flow rate based on the measurement data fed back by flowmeter 207 (No. 207), ensuring that the sampling flow rate of the first electric sampling pump 203 is stably controlled at 100 mL / min, and ensuring that the flow rate of the sample entering the collection container 308 and the sample entering the first inlet 32001 is stably controlled at 100 mL / min.

[0111] like Figure 1 As shown, the first electric sampling pump 203 extracts waste liquid from the reactor building pit a. The waste liquid passes through electric valves 101 and 102 or electric valves 103 and 104, electric valve 202, first electric sampling pump 203, sampling cooler 204, electric valve 205, flow meter 207, electric valve 602, electric valve 606, and second quick connector 307 into collection container 308. The liquid in the collection container 308 enters the first multi-way valve 320 via the third quick connector 310, the second electric sampling pump 311, the No. 312 electric valve 312, and the No. 314 flow meter. It then enters the injection module 50 via the first inlet end 32001, the first connection end 32006 of the first metering loop, the 1mL first metering loop 315, the second connection end 32003 of the first metering loop, the first outlet end 32002, and the No. 316 electric valve, forming a sampling loop. At this time, the first multi-way valve 320 is in sampling mode, with a sampling flow rate of 100mL / min and a sampling duration of 5min, ensuring that the radioactivity concentration of the liquid at each point in this sampling loop is the same as that of the waste liquid in reactor building sump a. Thus, 1mL of waste liquid from reactor building sump a is sampled into the 1mL first metering loop 315, at which point the radioactivity concentration of the liquid in the 1mL first metering loop 315 is the same as that of the waste liquid in reactor building sump a.

[0112] After the 1mL second quantitative loop 315 of the first multi-port valve 320 is injected, the operator remotely controls the 603 electric valve to close via the electric valve controller 701 of the measurement system control platform 70, switching the first multi-port valve 320 from injection mode to quantitative mode, and simultaneously opening the 401 electric valve. The demineralized water enters the first multi-port valve 320 via the 317 electric valve 317, the 318 flow meter, and then flows through the first diluent injection end 32004, the second connection end of the first quantitative loop 32003, the 1mL first quantitative loop 315, the first connection end of the first quantitative loop 32006, the connection end of the test liquid container 32005, the 604 electric valve 604, the 401 electric valve 401, and the first quick connector 403, ultimately entering the test liquid container 404. The flow meter 318 transmits its measurement data to the electric valve 317 via electrical transmission. Based on the measurement data, the electric valve 317 automatically adjusts the valve opening to ensure that the demineralized water enters the first multi-way valve 320 through the electric valve 317 and flushes 1 mL of liquid (with the same radioactive concentration as the waste liquid in reactor building pit a) from the first metering loop 315 to the test liquid container 404 at a stable flow rate of 100 mL / min. The flushing time is 1 minute.

[0113] The first liquid level monitoring module 405 transmits the liquid level data of the test liquid container 404 to the monitoring module controller 703 of the measurement system control platform 70 in real time via a measuring cable. When the liquid level of the test liquid container 404 reaches 100mL, the monitoring module controller 703 issues an alarm indicating that the test liquid container 404 has reached the measurement condition. The alarm lasts for 30 seconds and then automatically shuts off. At the same time, it sends an alarm signal to the electric valve controller 701, which automatically closes electric valves 317 (No. 317) and 401 (No. 401) to ensure that the liquid level of the test liquid container 404 is 100mL. At this time, the radioactivity concentration of the liquid in the test liquid container 404 is one percent of the waste liquid in the reactor building sump a.

[0114] when Sampling mode 3. After the γ dose rate meter 201 measurement data stabilizes, as follows... Figure 4 The data shown is from the measurement of the gamma dose rate meter 201. When sampling is performed, it is done according to mode 3.

[0115] The staff remotely controlled the electric valve controller 701 of the measurement system control platform 70 to sequentially open electric valves 605 (605), 608 (608), 602 (602), and 206 (206).

[0116] The sample inlet circuit is flushed. Demineralized water enters the spray module 50 via electric valve 206, flow meter 207, electric valve 602, electric valve 608, and electric valve 605, forming a flushing circuit.

[0117] Demineralized water is used to flush any residual radioactive waste or radionuclide contamination on the pipe walls in the above flushing circuit at a flow rate of 100 mL / min for 5 minutes. The flushing fluid enters the injection module 50. The injector 502 of the injection module 50 draws the waste liquid from the outlet of electric valve 605 (No. 605), and injects it into the reactor building pit b via electric valves 503 (No. 503), 105 (No. 105), and 106 (No. 106), or electric valves 107 (No. 107) and 108 (No. 108), thus forming a flushing circuit.

[0118] Sample collection was performed. After flushing the above-mentioned flushing circuit with deionized water at a flow rate of 100 mL / min for 5 minutes, the operator remotely controlled the electric valve controller 701 of the measurement system control platform 70 to close electric valves 206 and 607 and open electric valve 205. Flowmeter 207 transmitted its measurement data to the first electric sampling pump 203 in real time via electrical transmission. The first electric sampling pump 203 automatically adjusted its sampling flow rate based on the measurement data fed back by flowmeter 207, ensuring that the sampling flow rate of the first electric sampling pump 203 was stably controlled at 100 mL / min.

[0119] The first electric sampling pump 203 extracts waste liquid from reactor building sump a. The waste liquid passes through electric valves 101 and 102 or 103 and 104, electric valve 202, the first electric sampling pump 203, the sampling cooler 204, electric valve 205, flow meter 207, electric valves 602, 608, and 605 into the injection module 50. The entire process lasts 5 minutes. The radioactivity concentration of the sampled in the entire loop is the same as that of the waste liquid in reactor building sump a. Personnel remotely control the opening of electric valve 402 and the closing of electric valve 608 via the electric valve controller 701 on the measurement system control platform 70.

[0120] The first electric sampling pump 203 extracts the waste liquid from reactor building sump a. The waste liquid passes through electric valves 101 and 102 or 103 and 104, electric valve 202, the first electric sampling pump 203, sampling cooler 204, electric valve 205, flow meter 207, electric valve 602, electric valve 402, and the first quick connector 403 into the test liquid container 404. At this time, the radioactivity concentration of the liquid in the test liquid container 404 is the same as that of the waste liquid in reactor building sump a.

[0121] The first liquid level monitoring module 405 transmits the liquid level data of the test liquid container 404 to the monitoring module controller 703 of the measurement system control platform 70 in real time via a measuring cable. When the liquid level of the test liquid container 404 reaches 100mL, the monitoring module controller 703 issues an alarm indicating that the test liquid container 404 has reached the measurement condition. The alarm lasts for 30 seconds and then automatically shuts off. At the same time, it sends an alarm signal to the electric valve controller 701, which automatically closes electric valves 317 (No. 317) and 401 (No. 401) to ensure that the liquid level of the test liquid container 404 is 100mL. At this time, the radioactivity concentration of the liquid in the test liquid container 404 is the same as that of the waste liquid in the reactor building sump a.

[0122] Radioactivity measurements are performed on the sampled material. After the alarm is triggered when the test liquid container 404 reaches the measurement conditions, the staff checks whether the temperature of the high-purity germanium detector 406 has reached stable operating conditions through the high-purity germanium spectrometer controller 707 on the measurement system control platform 70. If the operating conditions have not been met, the cooler 407 continues to cool the high-purity germanium detector 406 until its temperature reaches the low temperature of -175℃ required for normal operation.

[0123] If the high-purity germanium detector 406 has reached its normal operating low-temperature condition, the operator initiates the gamma-ray spectrum measurement of the sample in the test liquid container 404 via the gamma-ray spectrum analysis software 706 configured on the measurement system control platform 70. The high-purity germanium detector 406 converts the measured gamma rays into corresponding voltage pulse amplitudes according to their energy levels. This voltage signal is then transmitted via a measurement cable to the main amplifier 704 of the gamma-ray spectrum measurement system on the measurement system control platform 70 for further amplification into a 0-5V voltage signal. The amplified voltage signal is then processed into a digital signal by the multichannel analyzer 705, achieving analog-to-digital conversion. The multichannel analyzer 705 transmits the digital signal to the gamma-ray spectrum analysis software 706 to collect gamma-ray spectrum data during the measurement of the sample in the test liquid container 404. It is generally recommended to set the data acquisition time to 3600 seconds or 5000 seconds (the data acquisition time is adjustable). After data acquisition is complete, the gamma-ray spectrum analysis software 706 automatically completes the analysis of the gamma-ray spectrum data and outputs the nuclide type i and radioactivity information A of the sample in the test liquid container 404.i .

[0124] Calculation of radioactive concentration of waste liquid in reactor building sump a. Following the corresponding sampling mode and using the γ-ray spectrometry analysis software 706, the analysis automatically obtained the nuclide types and radioactivity information of the sample taken from container 404. i The concentrations of each radionuclide in the waste liquid of reactor building pit a were calculated.

[0125] For sampling mode 1, i.e. At that time, according to sampling mode 1, the radioactivity concentration of the liquid in the test container 404 was one part per million of the waste liquid in reactor building sump a. Therefore, the relationship between the concentration of each radionuclide i in the waste liquid of reactor building sump a and the radioactivity concentration of each radionuclide i in the liquid in the test container 404 can be established:

[0126]

[0127] C i =10 10 ×A i (Bq / m 3 )

[0128] For sampling mode 2: i.e., 5 μSv / h < D ≤ 5 mSv / h. According to sampling mode 2, the radioactivity concentration of the liquid in the test container 404 is one percent of the waste liquid in reactor building sump a. Therefore, the relationship between the concentration of each radionuclide i in the waste liquid of reactor building sump a and the radioactivity concentration of each radionuclide i in the liquid in the test container 404 can be established:

[0129]

[0130] C i =10 6 ×A i (Bq / m 3 )

[0131] For sampling mode 3, i.e. At that time, according to sampling mode 3, the radioactivity concentration of the liquid in the tested liquid container 404 was the same as the radioactivity concentration of the waste liquid in reactor building sump a. The concentrations of each radionuclide i in the waste liquid of reactor building sump a were:

[0132]

[0133] C i =10 4 ×A i (Bq / m 3 )

[0134] Among them, A iThe radionuclide species i and its radioactivity were measured in Bq for the sample inside the test liquid container 404; C i The radioactivity concentration of each radionuclide i in the waste liquid of reactor building sump a is given in Bq / m³. 3 Bq / mL and Bq / m 3 These are all units for radioactivity concentration.

[0135] The above description is merely a specific embodiment of the present invention, but the scope of protection of the present invention is not limited thereto. Any person skilled in the art can easily conceive of various equivalent modifications or substitutions within the technical scope disclosed in the present invention, and these modifications or substitutions should all be covered within the scope of protection of the present invention. Therefore, the scope of protection of the present invention should be determined by the scope of the claims.

Claims

1. A reactor building radioactive liquid waste measuring device, characterized by, The application relates to a reactor building radioactive waste liquid measuring device. The measuring module comprises a measured liquid container, a first liquid level monitoring module and a radioactive measuring and controlling module, the first liquid level monitoring module is arranged on the side of the measured liquid container, the radioactive measuring and controlling module is arranged below the measured liquid container, the measured liquid container is used for containing measured radioactive waste liquid, the first liquid level monitoring module is used for monitoring the liquid level in the measured liquid container, and the radioactive measuring and controlling module is used for measuring the radioactive activity of the radioactive waste liquid. The measuring module comprises a first shielding shell, the measured liquid container is arranged in the first shielding shell, the radioactive measuring and controlling module comprises a high-purity germanium detector, a refrigerator and a temperature and humidity controller, the high-purity germanium detector is arranged below the measured liquid container, the refrigerator is connected with the high-purity germanium detector, and the temperature and humidity controller is arranged on the side of the high-purity germanium detector, the first shielding shell is used for shielding the environmental gamma rays and the cosmic gamma rays, and the high-purity germanium detector is used for converting the gamma rays of the radioactive waste liquid into electric signals. The radioactive measuring and controlling module further comprises a signal conversion module which is electrically connected with the high-purity germanium detector, and is used for converting the electric signals into digital signals.

2. The reactor building radioactive liquid waste measuring apparatus according to claim 1, characterized by, The first quantitative sampling module comprises a first quantitative ring and a first multi-way valve, the first multi-way valve comprises a first liquid inlet end, a first liquid outlet end, a first quantitative ring first connecting end, a first quantitative ring second connecting end, a first diluent injection end and a measured liquid container connecting end, and the first quantitative ring is used for quantitatively sampling the radioactive waste liquid; when in a sampling mode, the first liquid inlet end, the first quantitative ring first connecting end, the first quantitative ring, the first quantitative ring second connecting end and the first liquid outlet end are sequentially connected; and / or when in a quantitative mode, the first diluent injection end, the first quantitative ring second connecting end, the first quantitative ring, the first quantitative ring first connecting end and the measured liquid container connecting end are sequentially connected.

3. The reactor building radioactive liquid waste measuring apparatus according to any one of claims 1 to 2, characterized by, The application further relates to a reactor building radioactive waste liquid measuring device.

4. A reactor building radioactive liquid sampling measurement system, characterized by, The application relates to a reactor building radioactive waste liquid measuring device. The application relates to a reactor building radioactive waste liquid measuring device. The application relates to a reactor building radioactive waste liquid measuring device. The application relates to a reactor building radioactive waste liquid measuring device. ​ ​ 5. The reactor building radioactive liquid sampling and measurement system according to claim 4, characterized in that, The radioactive waste liquid sampling unit comprises a second quantitative sampling module and a sampling module, the sampling module comprises a collection container, a second shielding cover and a second liquid level monitoring module, the collection container is arranged in the second shielding cover, the second liquid level monitoring module is arranged on the side of the collection container, the second quantitative sampling module is arranged outside the second shielding cover, the second quantitative sampling module is connected with the input end of the collection container, the output end of the collection container is connected with the first quantitative sampling module, and the second shielding cover is used for shielding the γ rays released by the radioactive waste liquid in the collection container.

6. The reactor building radioactive liquid sampling and measurement system according to claim 5, characterized in that, The second quantitative sampling module comprises a second quantitative ring and a second multi-way valve, the second multi-way valve comprises a second liquid inlet end, a second liquid outlet end, a second quantitative ring first connecting end, a second quantitative ring second connecting end, a second diluent injection end and a collection container connecting end, and the second quantitative ring is used for quantitatively sampling the radioactive waste liquid; when in the sampling mode, the second liquid inlet end, the second quantitative ring first connecting end, the second quantitative ring, the second quantitative ring second connecting end and the second liquid outlet end are sequentially connected; and / or when in the quantitative mode, the second diluent injection end, the second quantitative ring second connecting end, the second quantitative ring, the second quantitative ring first connecting end and the collection container connecting end are sequentially connected in series.

7. The reactor building radioactive liquid sampling and measurement system according to claim 6, characterized in that, The liquid discharge unit comprises a spraying module and a liquid discharge module, the output end of the spraying module is connected with the liquid discharge module, and the input end of the spraying module is connected with the first quantitative sampling module and the second quantitative sampling module; the liquid discharge module is used for discharging the liquid in the sampling system into the reactor building pit through the containment, and the spraying module is used for pressurizing the liquid to be discharged from the sampling system.

8. The reactor building radioactive liquid sampling and measurement system according to claim 6, characterized in that, The liquid inlet unit comprises a liquid inlet module and a flow control module, the input end of the flow control module is connected with the liquid inlet module, and the output end of the flow control module is connected with the input end of the radioactive waste liquid measuring unit and the second liquid inlet end; the liquid inlet module is used for making the radioactive waste liquid in the reactor building pit enter the sampling and measuring system, and the flow control module is used for controlling the flow rate of the liquid in the sampling and measuring system.

9. The reactor building radioactive liquid sampling and measurement system according to any one of claims 4 to 8, characterized in that, The flushing module is further arranged, the flushing module is connected with the radioactive waste liquid sampling unit, the radioactive waste liquid measuring unit and the liquid discharge unit, and the flushing module is used for flushing the sampling and measuring system.

10. A method of sampling and measuring radioactive waste liquid in a reactor building, characterized by, The reactor building radioactive liquid sampling and measuring system is applied to any one of claims 4 to 9, comprising: opening the liquid inlet unit and the liquid discharge unit to make the radioactive waste liquid in the reactor building pit enter the reactor building radioactive liquid sampling and measuring system; obtaining the γ dose caused by the radioactive waste liquid in the reactor building pit by means of a γ dose rate meter, and selecting a sampling mode according to the γ dose caused by the radioactive waste liquid in the reactor building pit; starting the radioactive waste liquid sampling unit to sample the radioactive waste liquid according to the sampling mode; starting the radioactive waste liquid measuring unit to measure the nuclide species and the radioactivity concentration of the radioactive waste liquid according to the sampling mode; The liquid inlet unit and the liquid outlet unit are closed, and the sample liquid is taken from the radioactive waste liquid sampling unit.

11. The method of claim 10, wherein the method further comprises: The step of starting the radioactive waste liquid sampling unit to sample the radioactive waste liquid according to the sampling mode comprises: starting the radioactive waste liquid sampling unit; when the gamma dose is greater than a first preset value, using a first sampling mode, the radioactive waste liquid sampling unit samples the radioactive waste liquid according to a first preset dilution ratio of the first sampling mode; when the gamma dose is less than the first preset value and greater than a second preset value, using a second sampling mode, the radioactive waste liquid sampling unit samples the radioactive waste liquid according to a second preset dilution ratio of the second sampling mode; when the gamma dose is less than the second preset value, using a third sampling mode, the radioactive waste liquid sampling unit samples the radioactive waste liquid according to a third preset dilution ratio of the third sampling mode.

12. The method of claim 11, wherein the method further comprises: determining a volume of the radioactive liquid waste in the reactor building; and determining a volume of the sample of the radioactive liquid waste. The step of starting the radioactive waste liquid measuring unit to measure the nuclide species and the radioactivity concentration of the radioactive waste liquid according to the sampling mode comprises: injecting the radioactive waste liquid into a measured liquid container, and measuring the nuclide species and the radioactivity information of the radioactive waste liquid by a radioactive control module; when the gamma dose is greater than a first preset value, calculating the concentration of each radioactive nuclide in the waste liquid of the reactor plant pit according to a first preset formula of the first sampling mode and based on the radioactivity information of each nuclide; when the gamma dose is less than the first preset value and greater than a second preset value, calculating the concentration of each radioactive nuclide in the waste liquid of the reactor plant pit according to a second preset formula of the second sampling mode and based on the radioactivity information of each nuclide; when the gamma dose is less than the second preset value, calculating the concentration of each radioactive nuclide in the waste liquid of the reactor plant pit according to a third preset formula of the third sampling mode and based on the radioactivity information of each nuclide.

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