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

Abstract

The invention discloses a reactor building radioactive waste liquid measuring device, a sampling measuring system and a method. The reactor building radioactive waste liquid measuring device comprises a measuring module, the measuring module comprises a measured liquid container, a first liquid level monitoring module and a radioactivity measurement and control module, the first liquid level monitoring module is arranged on the peripheral side of the measured liquid container, the radioactivity measurement and control module is arranged below the measured liquid container, and the first liquid level monitoring module is connected with the radioactivity measurement and control module. 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 radioactivity measurement and control module is used for measuring the radioactivity concentration of the radioactive waste liquid; the first quantitative sample injection module is connected with the tested liquid container, and the first quantitative sample injection module is used for quantitatively injecting radioactive waste liquid into the tested liquid container. According to the technical scheme, measurement can be carried out without human intervention, and radiation safety of workers is guaranteed.

Description

Technical Field

[0001] The present invention is applied to the technical field of nuclear power plant waste liquid sampling and measurement under accident conditions, and in particular relates to a radioactive waste liquid measuring device, a sampling and measuring system and a method. Background Art

[0002] When a design basis accident (such as a large break loss of coolant) or a severe accident (core damage) occurs in a nuclear power plant, a large amount of radioactive material is released into the reactor building in the form of gas and liquid. Gaseous radioactive material diffuses in the atmosphere of the reactor building, and liquid radioactive material flows into the pit and is collected. In order to accurately assess the severity of the nuclear accident in the reactor building, it can be directly characterized by the measurement data of the on-site temperature, humidity, pressure and radiation monitoring instruments in the reactor building, or by sampling the atmosphere of the reactor building and the waste liquid in the pit to the laboratory, and analyzing the concentration of radionuclides in the atmospheric samples and waste liquid samples to achieve an accurate assessment of the severity of the nuclear accident. The sampling and analysis results of the radioactive waste liquid in the reactor building are important decision-making bases for the emergency command center of the nuclear power plant to carry out the orderly and controlled discharge of radioactive waste liquid from the reactor building to the environment.

[0003] In severe accident conditions, the radioactivity concentration of the wastewater in the pit can reach up to 10 15 Bq / m 3 The dose rate near the sampling pipeline is above 5mSv / h, which will bring a large radiation dose to the workers 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 reactor building radioactive waste liquid measurement device, a sampling measurement system and a method, which are intended to solve the problem of radiation safety threats to workers when sampling and measuring radioactive waste liquid in a pit in a reactor building after a design basis accident or a serious accident occurs in a nuclear power plant under the prior art.

[0005] In the first aspect, an embodiment of the present invention provides a reactor building radioactive waste liquid measuring device comprising: a measuring 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 arranged on the peripheral side of the test liquid container, and the radioactivity measurement and control module is arranged 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; a first quantitative injection module, the first quantitative injection module is connected to the test liquid container, and the first quantitative injection module is used to quantitatively inject radioactive waste liquid into the test liquid container.

[0006] In a second aspect, an embodiment of the present invention further provides a reactor building radioactive waste liquid sampling and measurement system, comprising: a radioactive waste liquid measuring unit, comprising the reactor building radioactive waste liquid measuring device as described above; a liquid inlet unit, the liquid inlet unit being used to allow the radioactive waste liquid in the reactor building pit to pass through the containment shell into the sampling and measurement system; a liquid discharge unit, connected to the radioactive waste liquid measuring unit, the liquid discharge unit being used to discharge the liquid in the sampling and measurement system through the containment shell into the reactor building pit; a radioactive waste liquid sampling unit, connected to the liquid inlet unit and the radioactive waste liquid measuring unit, the radioactive waste liquid sampling unit being used to sample radioactive waste liquid; a control unit, electrically connected to the radioactivity measuring unit, the radioactive waste liquid sampling unit, the liquid inlet unit and the liquid discharge unit, the control unit being used to control the operation of the sampling and measurement system.

[0007] In a third aspect, an embodiment of the present invention further provides a method for sampling and measuring radioactive waste liquid in a reactor building, which is applied to the reactor building radioactive waste liquid sampling and measuring system as described above, and comprises: opening the liquid inlet unit and the liquid discharge unit to allow the radioactive waste liquid in the reactor building pit to enter the reactor building radioactive liquid sampling and measuring system; obtaining the gamma dose caused by the radioactive waste liquid in the reactor building pit by a gamma dose rate meter, and selecting a sampling mode according to the gamma dose caused by the radioactive waste liquid in the reactor building pit; starting the radioactive waste liquid sampling unit, and sampling the radioactive waste liquid according to the sampling mode; starting the radioactive waste liquid measuring unit, and measuring the radioactivity concentration of the radioactive waste liquid according to the sampling mode; closing the liquid inlet unit and the liquid discharge unit, and taking sample liquid from the radioactive waste liquid sampling unit.

[0008] Compared with the prior art, the present invention has the following beneficial effects:

[0009] In the technical solution of the present invention, a radioactive waste liquid measuring device is provided, a test liquid container for containing radioactive waste liquid is provided therein, quantitative injection of radioactive waste liquid into the test liquid container is achieved through a first quantitative injection module, a first liquid level monitoring module is arranged on the periphery of the test liquid container to detect the liquid level, and a radioactivity measurement and control module is arranged at the bottom of the test liquid container to measure the radioactivity concentration of the radioactive waste liquid, thereby obtaining the radioactivity concentration of the radioactive waste liquid in the pit in the reactor building, so that the radioactivity concentration of the radioactive waste liquid in the pit in the reactor building can be measured without the intervention of staff, thereby ensuring the radiation safety of the staff. BRIEF DESCRIPTION OF THE DRAWINGS

[0010] In order to more clearly illustrate the technical solutions of the embodiments of the present invention, the accompanying drawings required for use in the description of the embodiments will be briefly introduced below. Obviously, the accompanying drawings described below are some embodiments of the present invention. For ordinary technicians in this field, other accompanying drawings can be obtained based on these accompanying drawings without paying any creative work.

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

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

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

[0014] Figure 4 A schematic diagram of 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 of 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 of 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 the embodiment of the reactor building radioactive waste liquid sampling and measurement method provided by the present invention;

[0018] Figure 8 A flow chart of sampling and measuring high-level radioactive waste liquid in a reactor building under accident conditions in an embodiment of the method for sampling and measuring radioactive waste liquid in a reactor building provided by the present invention;

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

[0020] Description of reference numerals:

[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 discharge 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. γ dose rate meter; 202. No. 202 electric valve; 203. First electric sampling pump; 204. Sampling cooler; 205. No. 205 electric valve; 206. No. 206 electric valve; 207. No. 207 flow meter;

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

[0024] 30101, second liquid inlet end; 30102, second liquid outlet end; 30103, second connection end of second quantitative ring; 30104, second diluent injection end; 30105, collection container connection end; 30106, first connection end of second quantitative ring;

[0025] 32001, first liquid inlet end; 32002, first liquid outlet end; 32003, second connection end of first quantitative loop; 32004, first diluent injection end; 32005, test liquid container connection end; 32006, first connection end of first quantitative loop;

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

[0027] 50, injection module; 501, No. 501 electric valve; 502, injector; 503, No. 503 electric valve;

[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. γ spectrum measurement system main amplifier; 705. Multi-channel analyzer; 706. γ spectrum analysis software; 707. High purity germanium spectrometer controller. DETAILED DESCRIPTION

[0030] The following will be combined with the drawings in the embodiments of the present invention to clearly and completely describe the technical solutions in the embodiments of the present invention. Obviously, the described embodiments are part of the embodiments of the present invention, not all of the embodiments. Based on the embodiments of the present invention, all other embodiments obtained by ordinary technicians in this field without creative work 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 "include" and "comprises" indicate the presence of described features, integers, steps, operations, elements and / or components, but do not exclude the presence or addition of one or more other features, integers, steps, operations, elements, components and / or combinations thereof.

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

[0033] It should be further understood that the term "and / or" used in the present description and the appended claims refers to any and all possible combinations of one or more of the associated listed items, and includes these combinations.

[0034] In order to solve the problem of radiation safety of workers caused by sampling and measuring radioactive waste liquid in a pit in a reactor building after a design basis accident or a serious accident occurs in a nuclear power plant under the prior art, the present invention proposes a reactor building radioactive waste liquid measuring device, which comprises: a measuring module 40, including a measured liquid container 404, a first liquid level monitoring module 405 and a radioactivity measurement and control module, the first liquid level monitoring module 405 is arranged on the periphery of the measured liquid container 404, and the radioactivity measurement and control module is arranged below the measured liquid container 404, wherein the measured liquid container 404 is used to accommodate the measured radioactive waste liquid, the first liquid level monitoring module 405 is used to monitor the liquid level in the measured liquid container 404, and the radioactivity measurement and control module is used to measure the radioactivity activity of the waste liquid in the measured liquid container 404; and a first quantitative sampling module 34, the first quantitative sampling module 34 is connected to the measured liquid container 404, and the first quantitative sampling module 34 is used to quantitatively inject radioactive waste liquid into the measured liquid container 404.

[0035] Reference Figure 1The so-called test liquid container 404 is the core part of the device, which is 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 so-called first liquid level monitoring module 405 is responsible for real-time monitoring of the changes in the liquid level in the container. The first liquid level monitoring module 405 is arranged on the periphery of the test liquid container 404 to ensure that the liquid level is within a safe range, prevent overflow or leakage, and ensure that the amount of radioactive waste liquid measured in the test liquid container 404 is at the preset required amount. In the embodiment of the present invention, the preset required amount of liquid level of the radioactive waste liquid in the test liquid container 404 is 100mL. In order 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 radioactive contamination caused by direct contact, in the embodiment of the present invention, the liquid level monitoring module is preferably a photoelectric liquid level sensor, a laser liquid level sensor or an ultrasonic liquid level sensor. The so-called radioactivity measurement and control module is specifically used to measure the radioactivity concentration of radioactive waste liquid, so as to provide real-time radiation level data support for the emergency command of the nuclear power plant to carry out the orderly and controlled discharge of radioactive waste liquid from the reactor building to the environment, providing an important decision-making basis. The radioactivity measurement and control module is located at the bottom of the liquid container 404 to be tested, because the radioactive substances in the radioactive waste liquid may form a concentration gradient in the liquid due to factors such as precipitation and stratification. Placing the radioactivity measurement and control module at the bottom of the container can more effectively detect the radioactivity concentration at the bottom of the liquid, ensuring that any possible high concentration area is captured. At the same time, the surface of the liquid may be affected by fluctuations, bubbles or steam, which may interfere with the measurement, and measuring at the bottom can reduce such interference and more accurately reflect the true radioactivity level of the liquid. Through the cooperation of liquid level monitoring and radioactivity measurement and control modules, online monitoring of radioactive waste liquid can be achieved. The first quantitative injection module 34 quantitatively injects the radioactive waste liquid in the liquid container 404 to be tested, which can accurately control the injection amount and ensure 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 quantitative radioactive waste liquid, thereby completing the measurement of the radioactivity concentration of the reactor building radioactive waste liquid after the accident without the participation of personnel.

[0036] Compared with the prior art, the reactor building radioactive liquid waste measuring device of the present invention is provided with a radioactive liquid waste measuring device, in which a test liquid container 404 for containing radioactive liquid waste is provided, quantitative injection of the radioactive liquid waste is achieved through a first quantitative sampling module 34, a first liquid level monitoring module 405 is arranged on the periphery of the test liquid container 404 to detect the liquid level, and a radioactivity measurement and control module is arranged 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 in the reactor building, so that the radioactivity concentration of the radioactive waste liquid in the pit in the reactor building can be measured without the intervention of staff, thereby ensuring the radiation safety of the staff.

[0037] In one embodiment, referring to Figure 1The measurement module 40 includes a first shielding shell 409, and the measured liquid container 404 is arranged in the first shielding shell 409. The radioactive measurement and control module includes a high-purity germanium detector 406, a refrigerator 407, and a temperature and humidity controller 408. The high-purity germanium detector 406 is arranged below the measured liquid container 404, the refrigerator 407 is connected to the high-purity germanium detector 406, and the temperature and humidity controller 408 is arranged on the side of the high-purity germanium detector 406. The first shielding shell 409 is used to shield the gamma rays in the environment and the universe, and the high-purity germanium detector 406 is used to convert the gamma rays of the radioactive waste liquid into electrical signals. The first shielding shell 409 is set to prevent the interference of the environment and cosmic rays, especially gamma rays, on the measurement results. The material of the first shielding shell 409 is 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 arranged in the first shielding shell 409, and the high purity germanium detector 406 is arranged below the test liquid container 404. At this time, the first shielding shell 409 can not only shield the gamma rays released by the radioactive waste liquid in the test liquid container 404, but also protect the high purity germanium detector 406 from the influence of external environmental radiation and cosmic ray factors, thereby improving the safety and measurement accuracy of the reactor plant radioactive waste liquid measurement device. The high purity germanium detector 406 is a highly sensitive gamma ray detector that can accurately measure the gamma rays emitted by the radioactive waste liquid and convert them into electrical signals. The electrical signals are used for real-time monitoring, data recording and subsequent statistical analysis. The high purity germanium detector 406 has a high energy resolution and can identify gamma rays of different energies, which is convenient for analyzing the nuclide type and radioactivity concentration of the radioactive waste liquid. Since the high purity germanium detector 406 can only achieve optimal performance when working at low temperatures, it is necessary to set a refrigerator 407 to connect with the high purity germanium detector 406. The refrigerator 407 is responsible for cooling the high-purity germanium detector 406 to the required working temperature. The high-purity germanium detector 406 is cooled to the required working temperature between -80°C and -200°C, and the optimal temperature is below -175°C. The high-purity germanium detector 406 can reduce thermal noise and improve detection sensitivity when it is at a stable working temperature. Setting a temperature and humidity controller 408 can detect the temperature environment in which the high-purity germanium detector 406 is located, and can also ensure that the high-purity germanium detector 406 is in a dry working state to protect the high-purity germanium detector 406 from damage caused by condensation droplets or ice generated by the cooling of the refrigerator 407. When the temperature and humidity controller 408 detects that the current temperature and humidity are not suitable for the high-purity germanium detector 406 to work, it will feedback the situation through the control circuit to which it is connected, and the refrigerator 407 will adjust the temperature according to the feedback. Through the scheme of the embodiment of the present invention, the measurement accuracy and environmental adaptability of the reactor building radioactive waste liquid measuring device are significantly improved, and the accurate measurement of gamma rays in the radioactive waste liquid can be achieved, and then the radioactive concentration of the waste liquid in the pit in the reactor building can be calculated.

[0038] Furthermore, a first shielding shell top cover 410 is provided above the first shielding shell 409, and the first shielding shell is provided on a measuring workbench 411. The first shielding shell top cover 410 is provided so that the staff can open the first shielding shell 409 to clean and resample the sample liquid under test in the test liquid container 404. The measuring workbench 411 is provided to isolate the direct contact between the first shielding shell 409 and the bottom surface, and on the basis of providing a stable placement platform for the measuring module 40, it also isolates interference from the ground or other possible contact planes, thereby improving the accuracy of the final measurement result.

[0039] In one embodiment, referring to Figure 1 The radioactive measurement and control module also includes a signal conversion module, which is electrically connected to the high-purity germanium detector 406, and the signal conversion module is used to convert the electrical signal into a digital signal. The signal conversion module is electrically connected to the high-purity germanium detector 406 and receives the electrical signal from the detector. The signal conversion module converts the analog electrical signal output by the high-purity germanium detector 406 into a digital signal to improve the speed and accuracy of data processing. In actual use, after converting the analog signal into a digital signal, the signal can be processed using digital signal processing techniques such as filtering, amplification, and analysis. After processing, the digital signal will reduce noise interference and improve the stability of the measurement. Therefore, by converting the electrical signal into a digital signal, the actual state of the radioactive waste liquid can be more accurately reflected and the error rate can be reduced. Furthermore, the converted digital signal is easier to store and transmit, and is convenient for connection with a computer system or other monitoring equipment so that the data can be effectively recorded, monitored in real time, and analyzed. The introduction of the signal conversion module enables the entire reactor plant radioactive waste liquid measurement device to be better integrated with other monitoring and control systems to achieve more comprehensive monitoring and management.

[0040] In one embodiment, referring to Figure 1 and Figure 3The first quantitative injection module 34 includes a first quantitative ring 315 and a first multi-way valve 320, and the first multi-way valve 320 includes a first liquid inlet end 32001, a first liquid outlet end 32002, a first quantitative ring first connection end 32006, a first quantitative ring second connection end 32003, a first diluent injection end 32004 and a test liquid container connection end 32005; wherein, when in the injection mode, the first liquid inlet end 32001, the first quantitative ring first connection end 32006, the first quantitative ring 302, the first quantitative ring second connection end 32003, and the first liquid outlet end 32002 are connected in sequence, and the first quantitative ring 302 is used for quantitative sampling of radioactive waste liquid; and / or, when in the quantitative mode, the first diluent injection end 32004, the first quantitative ring second connection end 32003, the first quantitative ring 315, the first quantitative ring first connection end 32006, and the test liquid container connection end 32005 are connected in series in sequence. The first quantitative injection module 34 includes a first quantitative ring 315, which is used to accurately measure and control the volume of the liquid, ensure the volume consistency of the sample each time, and reduce the measurement error caused by the fluctuation of the sample volume. The so-called multi-way valve is used to connect different connection ends to achieve the connection between the functional pipelines, and to achieve the change of the intercommunication relationship between different connection ends by switching the valve port. In this embodiment, a rotary multi-way valve is used. In the rotary multi-way valve, the valve core is cylindrical or spherical. As the valve stem rotates, different faces of the valve core dock with the channels in the valve body, thereby achieving switching. In order to achieve quantitative measurement of the radioactive waste liquid in the pit in the reactor building after the accident, it is necessary to first quantitatively inject the radioactive waste liquid through the quantitative ring, and then transfer the radioactive waste liquid in the quantitative ring. To this end, to achieve the above-mentioned fixed sampling and sampling, at least one multi-way valve including six external ends is required to realize the function. On the multi-way valve, a first liquid inlet end 32001 and a first liquid outlet end 32002 are provided to provide a channel for the injection and transfer of radioactive waste liquid. A first connection end 32006 of the first quantitative loop and a second connection end 32003 of the first quantitative loop are also provided, which serve as channels for the radioactive waste liquid to enter and leave the first quantitative loop 315, respectively. A first diluent injection end 32004 and a test liquid container connection end 32005 are also included. The so-called first diluent injection end 32004 is used to connect the diluent source 610. The diluent in the embodiment of the present invention specifically refers to deionized water, which is used to adjust the dilution ratio of the sample liquid or the test liquid finally sampled or measured. The so-called test liquid container connection end 32005 is used to connect the test liquid container 404, and is used to inject the quantitative radioactive waste liquid transferred from the quantitative loop into the test liquid container 404 for subsequent measurement. In this embodiment, when the radioactive waste liquid is sampled using the quantitative loop and when the radioactive waste liquid sample quantitatively injected in the quantitative loop is transferred to the test liquid container 404, it is respectively referred to as the injection mode and the quantitative mode. In these two modes, the multi-way valve is in two different valve port switching states.When in the injection mode, the flow path of the radioactive waste liquid is the first liquid inlet end 32001, the first connection end 32006 of the first quantitative loop, the first quantitative loop 315, the second connection end 32003 of the first quantitative loop, and the first liquid outlet end 32002. Through this path, it can be ensured that the liquid is not disturbed after entering the quantitative loop, and the state of the injected radioactive waste liquid is accurately maintained. When in the quantitative mode, the flow path of the deionized water flowing out of the diluent source 610 for transferring and diluting the radioactive liquid in the quantitative loop is 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 connection end 32005 of the test liquid container. Through this path, the necessary dilution can be performed when the radioactive waste liquid enters the test liquid container 404 so as to meet the concentration requirements of subsequent analysis. The reason for quantification is that the radioactive activity of the substance being tested that can be accepted by the high-purity germanium detector 406 for analysis has a certain upper limit. If the waste liquid is not quantified, the radioactive activity of the obtained liquid being tested will be much higher than the detection upper limit of the high-purity germanium detector, making it impossible for the high-purity germanium detector 406 to measure normally.

[0041] The present invention also provides a reactor building radioactive liquid sampling and measuring system, referring to Figures 1 to 3, which includes: a radioactive waste liquid measurement unit, including the above reactor building radioactive waste liquid measurement device; a liquid inlet unit, the liquid inlet unit is used to allow the radioactive waste liquid in the reactor building pit to pass through the containment into the sampling and measurement system; a liquid discharge unit, connected to the radioactive waste liquid measurement unit, the liquid discharge unit is used to discharge the liquid in the sampling and measurement system through the containment into the reactor building pit; a radioactive waste liquid sampling unit 31, connected to the liquid inlet unit and the radioactive waste liquid measurement unit, the radioactive waste liquid sampling unit 31 is used to sample the radioactive waste liquid; a control unit, electrically connected to the radioactive measurement unit, the radioactive waste liquid sampling unit 31, the liquid inlet unit and the liquid discharge unit, the control unit is used to control the operation of the sampling and measurement system. While measuring the radioactive waste liquid, sampling must be carried out and retained, and the radioactive activity of the radioactive waste liquid must be further measured and analyzed in the laboratory using a gamma 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, and ensure that the radioactive concentration of the liquid is within the safe range and will not cause harm to the staff, and then the sample liquid obtained by sampling will be manually sampled and measured and analyzed in the laboratory. Outside the reactor building, a containment is set up, which can effectively contain and prevent radioactive substances from leaking into the environment when an accident or failure occurs in the reactor. It is usually built with strong materials that can withstand high pressure and high temperature to ensure that it can remain sealed under extreme conditions. Therefore, in order to enable the reactor building radioactive liquid sampling and measurement system to obtain radioactive waste liquid from the pit of the reactor building after the accident, it is necessary to set up a liquid inlet unit to ensure that the radioactive liquid can be safely and effectively introduced into the system without destroying the protective performance of the containment. At the same time, 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 to cause radiation hazards, a liquid discharge unit is also required to discharge the liquid in the sampling system through the containment into the pit of the reactor building. The radioactive liquid waste sampling unit 31 is used to sample the radioactive liquid waste in the reactor building pit. A certain volume of liquid sample is obtained through a quantitative sampling device, and a diluent is injected to dilute it to reduce radioactivity. The radioactive liquid waste sampling unit 31 is designed to ensure the representativeness and accuracy of the sample for subsequent analysis and detection. The control unit is the control core of the entire sampling and measurement system, responsible for coordinating and controlling the work of each unit. It communicates with the radioactivity measurement unit, the radioactive liquid waste sampling unit 31, the liquid inlet unit and the liquid discharge unit through electrical connection, and can realize automatic 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 measuring system of the present invention, radioactive waste liquid is introduced into the sampling and measuring 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 measuring unit, and simultaneously measures the radioactivity level, and after the sampling is completed, the measured and sampled liquid is safely discharged back to the reactor building pit through the liquid discharge unit. In order to effectively control the various units, modules and components in the reactor building radioactive liquid sampling and measuring system to accurately realize the functions of the sampling and measuring system, a measuring system control platform 70 is also provided, which is electrically connected to the various electronic control components, receives feedback from the various components for automatic control, and also actively intervenes in the control according to the instructions issued by human input.

[0042] In one embodiment, referring to Figure 1 and Figure 2The radioactive waste liquid sampling unit 31 includes a second quantitative sampling 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 arranged in the second shielding cover 309. The second liquid level monitoring module 319 is arranged on the side of the collection container 308. The second quantitative sampling module is arranged outside the second shielding cover 309. The second quantitative sampling 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 sampling module 34. The second shielding cover 309 is used to shield the gamma rays released by the radioactive waste liquid in the collection container 308. The radioactive waste liquid sampling unit 31 needs to complete the two tasks of quantitative sampling and sampling of radioactive waste liquid, so it includes the second quantitative sampling module 32 and the sampling module 33. The second quantitative sampling module 32 is used to quantitatively introduce the radioactive waste liquid in the reactor building pit into the sample collection container 308, which ensures that the volume of the radioactive liquid entering the sampling module 33 is controlled during the sampling process to obtain an accurate sample volume. The sampling module 33 includes a sample collection container 308 for collecting radioactive waste liquid samples extracted from the reactor building. The sample collection container 308 must meet the safety standards of corrosion resistance and radiation resistance. The sampling module 33 also includes a second shielding cover 309 wrapped outside the sample collection container 308. The second shielding cover 309 is mainly used to shield the gamma rays emitted by the radioactive waste liquid in the collection container 308 to protect the radiation safety of the operator and the surrounding environment. The material and thickness of the second shielding cover 309 shielding cover must be sufficient to shield the gamma rays emitted by the radioactive waste liquid in the collection container 308 to ensure radiation safety. It is usually made of a material with a higher density, such as lead or concrete, to effectively shield the 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 will not overflow or be insufficient during the sampling process. In actual use, the liquid level monitoring function can be realized by a photoelectric liquid level sensor, a laser liquid level sensor or an ultrasonic liquid level sensor, so as to timely feedback the state of the collection container 308. In the scheme of the embodiment of the present invention, the second quantitative injection module 32 is connected to the input end of the collection container 308, responsible for quantitatively introducing the radioactive waste liquid, and the output end of the collection container 308 is connected to the first liquid inlet end 32001 of the first quantitative injection module 34, which is used to transfer the liquid sample to the first quantitative sample tube to prepare for the measurement of the reactor building radioactive waste liquid measurement device.

[0043] Further, refer to Figure 1 and Figure 2The second quantitative injection module 32 includes a second quantitative ring 302 and a second multi-way valve 301, the second multi-way valve 301 includes a second liquid inlet end 30101, a second liquid outlet end 30102, a second quantitative ring first connection end 30106, a second quantitative ring second connection end 30103, a second diluent injection end 30104 and a collection container connection end 30105, and the second quantitative ring 302 is used for quantitative sampling of radioactive waste liquid; wherein, when in the injection mode, the second liquid inlet end 30101, the second quantitative ring first connection end 30106, the second quantitative ring 302, the second quantitative ring second connection end 30103, and the second liquid outlet end 30102 are connected in sequence; and / or, when in the quantitative mode, the second diluent injection end 30104, the second quantitative ring second connection end 30103, the second quantitative ring 302, the second quantitative ring first connection end 30106, and the collection container connection end 30105 are connected in series in sequence. The second quantitative injection module 32 is the same as the first quantitative injection module 34, and also includes a quantitative ring and a multi-way valve, which are a second quantitative ring 302 and a second multi-way valve 301. Among them, the second multi-way valve 301 is provided with a second liquid inlet end 30101 and a second liquid outlet end 30102, which provide a channel for the injection and transfer of radioactive waste liquid. A second quantitative ring first connection end 30106 and a second quantitative ring second connection end 30103 are also provided, which serve as channels for the radioactive waste liquid to enter and leave the second quantitative ring 302 respectively. It also includes a second diluent injection end 30104 and a collection container connection end 30105. The so-called second diluent injection end 30104 is used to connect a dilution source. The diluent is the same as the first quantitative injection module 34. The diluent specifically refers to deionized water, which is used to adjust the dilution ratio of the sample liquid or the measured liquid finally sampled or measured. The so-called collection container connection end 30105 is used to connect the collection container 308, and is used to inject the quantitative radioactive waste liquid transferred from the second quantitative ring 302 into the collection container 308 to achieve sampling. In this embodiment, when the radioactive waste liquid is sampled using the quantitative ring and when the radioactive waste liquid sample in the quantitative ring is transferred to the collection container 308 after quantitative injection, it is respectively referred to as the injection mode and the quantitative mode. In these two modes, the multi-way valve 301 is in two different valve port switching states. When in the injection mode, the flow path of the radioactive waste liquid is the second liquid inlet end 30101, the first connection end 30106 of the second quantitative ring, the second quantitative ring 302, the second connection end 30103 of the second quantitative ring, and the second liquid outlet end 30102. Through this path, it can be ensured that the liquid in the reactor pit is not disturbed after entering the quantitative ring, and the state of the injected radioactive liquid is accurately maintained. When in the quantitative mode, the flow path of the deionized water flowing out from the diluent source 610 for transferring and diluting the radioactive liquid in the quantitative ring is the second diluent injection end 30104, the second quantitative ring second connection end 30103, the second quantitative ring 302, the second quantitative ring first connection end 30106, and the collection container connection end 30105.Through this path, necessary dilution can be performed when the radioactive waste liquid enters the collection container 308 so as to meet the concentration requirements for sampling. The reason for dilution is that the radioactive concentration of the radioactive waste liquid may be very high, which may cause radiation hazards to the staff who transport and collect the samples. At the same time, due to the excessively high radioactivity, it cannot be measured and analyzed by the gamma spectrometer in the laboratory, and a secondary dilution is required. In this process, there is a risk of leakage of radioactive waste liquid, which may cause radiation safety accidents such as leakage and diffusion of radioactive liquid. For this reason, the radioactive waste liquid is directly diluted by a certain proportion with deionized water when sampling. The second quantitative injection module 32 can quickly switch between injection and quantitative modes through a flexible multi-way valve, ensuring the effective sampling and treatment of radioactive waste liquid.

[0044] Further, refer to Figures 1 to 3 , the volume of the second quantitative ring 302 is 0.1ml, the volume of the first quantitative ring 315 is 1ml, the volume of the collection container 308 is 1000ml, and the volume of the test liquid container 404 is 100ml. The radioactive waste liquid in the reactor plant pit is first quantitatively sampled by the second quantitative ring 302, then enters the collection container 308 and is diluted, and then is extracted by the second electric sampling pump 311, and then flows through the first quantitative ring 315, and finally enters the test liquid container 404. In this process, the solution in the test liquid container 404 is diluted in two stages and has a lower radioactivity. Therefore, when the radioactivity concentration of the radioactive waste liquid in the reactor plant pit is very high, the radioactivity of the diluted sample that finally enters the test liquid container 404 can be within the measurement range of the high-purity germanium detector 406 through two-stage dilution, thereby avoiding measurement failure caused by exceeding the measurement and analysis range.

[0045] In one embodiment, referring to Figure 1The drainage unit includes an injection module 50 and a drainage module 11, the output end of the injection module 50 is connected to the drainage module 11, and the input end of the injection module 50 is connected to the first quantitative injection module 34 and the second quantitative injection module 32, wherein the drainage module 11 is used to allow the liquid in the sampling system to pass through the containment and be discharged into the reactor building pit, and the injection module 50 is used to pressurize the liquid to be discharged from the sampling system. In order to enable the liquid in the reactor building radioactive liquid sampling and measuring system of the present invention to be discharged into the reactor building pit to reduce the residual radioactive material in the system and overcome the possible pressure difference between the inside and outside of the containment, it is necessary to set a drainage module 11 that passes through the containment and an injection module 50 for injecting liquid in the pressurized assist system. The function of the injection module 50 is mainly realized by the injector 502. The injector 502 is connected to the diluent source 610. The desalted water in the diluent source 610 exists as a boosting and cleaning liquid. A No. 501 electric valve 501 is also provided on the pipeline connecting the injector 502 and the diluent source 610. The electric valve is used to prevent the desalted 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 radioactive liquid sampling and measurement system reaches the ejector 502, the ejector 502 mixes the desalted water and performs pressurized ejection, and then the liquid is ejected from the drainage module 11 through the containment into the pit of the reactor building. When the radioactive liquid is discharged, in order to prevent the pipeline from flowing back, the ejector 502 will continue to eject a certain amount of desalted water so that the pipeline between the drainage module 11 and the terminal outlet is filled, and the No. 503 electric valve 503 is closed to achieve isolation, thereby preventing the radioactive waste liquid from flowing back along the pipeline of the drainage module 11 due to the pressure change in the reactor building. Four electric valves are also arranged on the drainage module 11, namely, No. 105 electric valve 105, No. 106 electric valve 106, No. 107 electric valve 107 and No. 108 electric valve 108. Among them, No. 105 electric valve 105 and No. 107 electric valve are located inside the containment, and No. 106 electric valve 106 and No. 108 electric valve 108 are located outside the containment. The No. 106 electric valve 106 and the No. 105 electric valve 105 are connected along the liquid flow direction, and are a group of flow safety control valves that control the discharge of liquid outside the containment to the reactor plant pit. When the liquid discharge module 11 needs to discharge liquid, the No. 106 electric valve 106 and the No. 105 electric valve 105 will be opened to discharge the liquid. However, when a design basis accident or a serious accident occurs, the electric valve is likely to be damaged, especially the No. 105 electric valve 105 and the connected pipelines in the containment, and under accident conditions, it is impossible to repair and replace the components and pipelines in the containment.In order to ensure that the sampling system can still discharge liquid to the pit in the reactor building normally after an accident, the No. 107 electric valve 107 and the No. 108 electric valve 108 are set as redundant equipment for the No. 105 electric valve 105 and the No. 106 electric valve 106 to ensure the normal operation of the sampling and measurement of radioactive liquid in the reactor building. It should be pointed out that the redundant equipment does not only include the No. 107 electric valve 107 and the No. 108 electric valve 108. For the No. 106 electric valve 106 and the No. 105 electric valve 105, at least three groups of redundant equipment are usually set to ensure the normal operation of the sampling system function. This embodiment only sets one group of redundant equipment in the attached figure for explanation.

[0046] In one embodiment, referring 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 measuring unit and the second liquid inlet end 30101, wherein the liquid inlet module 10 is used to allow the radioactive waste liquid in the reactor building pit to enter the sampling and measuring system, and the flow control module 20 is used to control the flow rate of the liquid in the sampling and measuring system.

[0047] Reference Figure 1In order to allow the radioactive waste liquid in the pit in the reactor building to smoothly enter the sampling and measuring system, a liquid inlet module 10 and a flow control module 20 are set. The liquid inlet module 10 is also provided with four electric valves, namely, electric valve No. 101 101, electric valve No. 102 102, electric valve No. 103 103 and electric valve No. 104. Among them, electric valve No. 101 101 and electric valve No. 103 103 are located inside the containment, and electric valve No. 102 102 and electric valve No. 104 are located outside the containment. Electric valve No. 101 101 and electric valve No. 102 102 are connected along the liquid flow direction, and are a group of flow safety control valves that control the liquid entering the containment. When the liquid inlet module 10 needs radioactive waste liquid to enter, electric valve No. 101 101 and electric valve No. 102 102 will be opened, and the radioactive liquid in the pit in the reactor building will enter. However, when a design basis accident or a serious accident occurs, the electric valve is likely to be damaged, especially the No. 101 electric valve 101 and the connected pipelines in the containment, and under accident conditions, it is impossible to repair and replace the components and pipelines in the containment. In order to ensure that the sampling and measurement system can still normally take in liquid from the reactor building pit after an accident, the No. 103 electric valve 103 and the No. 104 electric valve are set as redundant equipment for the No. 101 electric valve 101 and the No. 102 electric valve 102 to ensure the normal operation of the reactor building radioactive liquid sampling and measurement system. It should be pointed out that the redundant equipment does not only include the No. 103 electric valve 103 and the No. 104 electric valve. For the No. 101 electric valve 101 and the No. 102 electric valve 102, at least three groups of redundant equipment are usually set to ensure the normal operation of the sampling system function. This embodiment only sets a group of redundant equipment in the accompanying drawings for explanation. A gamma dose rate meter 201 capable of online monitoring is provided in the flow control module 20. The gamma dose rate meter 201 is used to perform preliminary measurement of the gamma dose rate caused by the radioactive waste liquid entering the sampling and measurement system, and provide a basis for the subsequent measurement system control platform 70 to control the sampling and measurement modes of the second quantitative sampling module 32, the sampling module 33, the first quantitative sampling module 34 and the measurement module 40.

[0048] Reference Figure 1Due to the design basis of the reactor building, the pit of the reactor building must be at a low position relative to the sampling and measuring system of the radioactive liquid in the reactor building. Therefore, if the radioactive waste liquid in the pit is to enter the sampling and measuring system, it is necessary to provide power for the radioactive waste liquid, so a first electric sampling pump 203 is set in the flow control module 20. A No. 202 electric valve 202 is set at the front end of the first electric sampling pump 203 along the liquid flow direction to control the on-off of the pipeline. A sampling cooler 204 is set at the rear end of the first electric sampling pump 203 along the liquid flow direction. When the nuclear power plant has a design basis accident or a serious accident for a long time, the reactor building environment in the containment is in a high temperature state due to the heat spillage of the reactor. At this time, the temperature of the radioactive waste liquid in the pit of the reactor building will be very high. High-temperature liquid will affect the working safety of the sampling and measuring system and the working environment temperature conditions required by the high-purity germanium detector 406 in the measuring module 40. Therefore, the radioactive waste liquid entering the sampling and measuring system must be cooled by the sampling cooler 204, and finally the temperature is cooled to below 30°C. In the embodiment of the present invention, the front and rear ends of the sampling cooler 204 along the liquid flow direction are also provided with temperature sensors (not shown in the figure). Through the feedback of these temperature sensors, the measurement system control platform 70 controls the flow rate of the coolant in 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 liquid cooling pipeline conditions that exist around the nuclear reactor building itself, in actual use, the sampling cooler 204 will be directly connected to the liquid cooling pipeline in the existing nuclear power plant and supplied with cooling water.

[0049] In one embodiment, referring to Figure 1The reactor building radioactive liquid sampling and measuring system of the present invention also includes a flushing module 21, which is connected to the radioactive waste liquid sampling unit 31, the radioactive waste liquid measuring unit and the drainage unit. The flushing module 21 is used to flush the sampling and measuring system. The flushing module 21 includes a diluent source 610 and a No. 206 electric valve 206 for controlling the on-off of the diluent source 610 pipeline. A No. 205 electric valve 205 is provided between the flow control module 20 and the connecting pipeline of the flushing module 21. Through the diluent source 610, the pipeline can be flushed with the deionized water in the diluent source 610 before the sampling and measuring system performs sampling and measurement work, thereby reducing the measurement error caused by the residue in the sampling and measuring system pipeline and improving the reliability and accuracy of the measured data. When an abnormal situation occurs or contamination is detected, the flushing module 21 can be quickly started for cleaning to reduce risks. In particular, the flow control module 20 is also provided with a No. 207 flow meter 207 for monitoring the flow before entering the second quantitative sampling module 32, to prevent the sampling measurement system pipeline from bursting and being damaged due to excessive pressure in the pipeline due to excessive flow. When the pressure is too high, the No. 205 electric valve 205 provided between the No. 207 flow meter 207 and the sampling cooler 204 can be used to control the pipeline on and off to achieve rapid disconnection, and the pipeline pressure can be reduced in coordination with the output of deionized water from the diluent source 610.

[0050] In one embodiment, referring to Figure 1 The output end of the sampling cooler 204 is also connected to the input end of the ejector 502, and a No. 607 electric valve 607 is provided between the two for controlling the on-off. In some special cases, when the radioactive waste liquid in the pit is at an extremely high temperature or high radiation, the sampling and measurement operations must be abandoned. At this time, after the temperature sensor at the rear end of the γ dose rate meter 201 and the sampling cooler 204 returns an over-temperature or over-radiation threshold alarm, the No. 205 electric valve 205 will be closed, the No. 607 electric valve 607 will be opened, and the radioactive waste liquid will be directly discharged back to the pit of the reactor building through the ejector module 50 and the drain module 11.

[0051] In one embodiment, referring 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 γ spectrum measurement system main amplifier 704, a multi-channel analyzer 705, a γ spectrum analysis software 706, and a high-purity germanium spectrometer controller 707. Among them, the electric valve controller 701 controls the opening and closing of the electric valve and adjusts the flow rate of the flow of the radioactive waste liquid. This can ensure that the flow of the liquid is controllable during the sampling or measurement process. The sampling pump controller 702 controls the operation of the sampling pump and accurately adjusts the working state of the sampling pump according to the required sampling amount and sampling speed. The monitoring module controller 703 is used to monitor and obtain the real-time changes of various indicators from various monitors, such as the γ dose rate meter 201, the liquid level monitor, the temperature and humidity controller 408, the temperature sensor, etc., and display them. The main amplifier 704 of the γ spectrum measurement system is used to increase the strength of the signal so that subsequent analysis and processing can obtain clearer results. The multi-channel analyzer 705 is used to digitize the amplified signal, separate the gamma ray signals of different energies, and convert them into analyzable data. The gamma spectrum analysis software 706 further processes and analyzes the data output by the multi-channel analyzer 705, generates an energy spectrum, and extracts 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, temperature control, etc., to ensure the accuracy and stability of the measurement.

[0052] Reference Figure 1 In order to make the function of the reactor building radioactive waste liquid sampling and measurement system more complete, auxiliary components 60 are also needed, including No. 601 electric valve 601, No. 602 electric valve 602, No. 603 electric valve 603, No. 604 electric valve 604, No. 605 electric valve 605, No. 606 electric valve 606, No. 607 electric valve 607, and No. 608 electric valve 608. These electric valves are controlled by the measurement system control platform 70 to manage the on and off of their respective connecting pipelines, and are electric isolation valves. The auxiliary component 60 also includes the diluent source 610 mentioned above. The diluent source 610 can be a single component, or multiple components can be set according to the needs of pipeline layout. It is used to output demineralized water to dilute the radioactive waste liquid, and also plays a role in flushing the sampling and measurement system pipeline.

[0053] The present invention also provides a method for sampling and measuring radioactive waste liquid in a reactor building, which is applied to the above-mentioned radioactive liquid sampling and measuring system in a reactor building, referring to Figures 4 to 6 , which includes:

[0054] S801, opening the liquid inlet unit and the liquid discharge 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 the liquid discharge unit to ensure that the liquid flows in the system for subsequent gamma dose rate measurement. At this time, the radioactive waste liquid enters the system.

[0056] S802, obtaining the gamma dose caused by the radioactive waste liquid in the reactor building pit by using the gamma dose rate meter 201, and selecting a 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 obtain the gamma dose rate data of the radioactive waste liquid in real time, and the measured gamma dose rate value is recorded through the measurement system control platform 70 to provide a basis for the subsequent selection of the sampling mode.

[0058] S803, starting the radioactive waste liquid sampling unit 31, sampling 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 judged according to a preset threshold or scheme, so as to select an appropriate sampling mode (such as a high radioactivity concentration mode, a low radioactivity concentration mode, etc.).

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

[0061] S8031, start the radioactive waste liquid sampling unit 31;

[0062] S8032a, when the gamma 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 gamma 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 gamma 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] According to the comparison between the measured γ dose value and the first preset value and the second preset value, the corresponding sampling mode is selected and executed. When the γ dose is greater than the first preset value, the radioactive waste liquid sampling unit 31 dilutes the radioactive waste liquid according to the first preset dilution ratio set in the first sampling mode. When the γ dose is less than the first preset value and greater than the second preset value, the radioactive waste liquid sampling unit 31 performs dilution sampling according to the second preset dilution ratio set in the second sampling mode. When the γ dose is less than the second preset value, the radioactive waste liquid sampling unit 31 performs dilution sampling according to the third preset dilution ratio set in the third sampling mode.

[0066] S804, starting the radioactive waste liquid measurement unit to measure the type of nuclides and the radioactivity concentration of the radioactive waste liquid according to the sampling mode;

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

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

[0069] S8041, injecting radioactive waste liquid into the test liquid container 404, and measuring the type of nuclides and radioactivity information of the radioactive waste liquid through the radioactivity measurement and control module;

[0070] S8042a, when the gamma dose is greater than a first preset value, the concentration of each radioactive nuclide in the waste liquid in the reactor plant pit is calculated based on the radioactivity information according to a first preset formula of the first sampling mode;

[0071] S8042b, when the gamma dose is less than the first preset value and greater than the second preset value, the concentration of each radioactive nuclide in the waste liquid in the reactor plant pit is calculated according to the radioactivity information using a second preset formula of the second sampling mode;

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

[0073] The diluted radioactive waste liquid is injected into the liquid container 404 to be tested, and the injected radioactive waste liquid is measured using the radioactive measurement and control module. According to the measured γ dose and radioactivity information, different formulas are used to calculate the concentration of each radioactive nuclide in the reactor plant waste liquid. When the γ dose is greater than the first preset value, the concentration of each radioactive nuclide in the reactor plant waste liquid is calculated using the radioactivity information according to the first preset formula of the first sampling mode. When the γ dose is less than the first preset value and greater than the second preset value, the concentration of each radioactive nuclide in the reactor plant waste liquid is calculated according to the second preset formula of the second sampling mode in combination with the radioactivity information. When the γ dose is less than the second preset value, the concentration of each radioactive nuclide in the reactor plant waste liquid is calculated according to the radioactivity information using the third preset formula of the third sampling mode. Among them, the formula used needs to take into account the radioactivity of each nuclide in the waste liquid and the influence of environmental conditions on the measurement results. According to the change of γ dose, the formula also needs to be adjusted according to the characteristics of different γ doses to improve the measurement accuracy.

[0074] S805 , closing the liquid inlet unit and the liquid discharge unit, and taking sample liquid from the radioactive waste liquid sampling unit 31 .

[0075] After the sampling and measurement are completed, the liquid inlet unit and the liquid discharge unit are closed, the flow of the waste liquid is stopped, and the pipeline is flushed using the flushing module 21. After the liquid inlet and liquid discharge units are closed, the sampled liquid sample is taken out from the radioactive waste liquid sampling unit 31.

[0076] In order to more clearly illustrate the reactor building atmosphere sampling system and method of the present invention, refer to 1 to Fig. 9 The following is a complete example of a post-accident reactor building wastewater sampling and measurement program:

[0077] During the normal operation of a nuclear power plant, the post-accident reactor plant underground pit waste liquid sampling and measurement system is in a shutdown state; during the shutdown and overhaul of a nuclear power plant, the post-accident reactor plant underground pit waste liquid sampling and measurement system is started to inspect and test the equipment status and operational reliability. If any component failure occurs during the inspection and testing, it will be replaced and repaired in a timely manner.

[0078] Sampling of pit waste liquid and radioactivity analysis must be carried out within 24 hours after a nuclear power plant accident occurs, providing radiation data support for the nuclear power plant emergency command center to conduct accident progress assessment.

[0079] After a large break loss of coolant or a serious accident occurs in a nuclear power plant, the staff immediately remotely starts the high-purity germanium detector 406, the refrigerator 407, and the temperature and humidity controller 408 of the measurement module 40 through the high-purity germanium spectrometer controller 707 of the measurement system control platform 70 to prepare for the radioactivity measurement of the sampled samples.

[0080] In the case of a large break loss of coolant or severe accident, high-level radioactive waste in the primary coolant or a large amount of radioactive material released by core damage is collected in the reactor pit. The object of sampling and measurement of high-level radioactive waste in the reactor pit under accident conditions is the waste in the reactor pit.

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

[0082] The electric valves of the liquid inlet module 10 and the liquid discharge module 11 are controlled to open. After the nuclear power plant emergency command issued an instruction to perform sampling and measurement of the waste liquid in the reactor plant pit, the operator of the main control room of the nuclear power plant remotely controlled the opening of the No. 101 electric valve 101, No. 102 electric valve 102, No. 105 electric valve 105 and No. 106 electric valve 106 of the waste liquid sampling pipeline in the main control room. When any one of the No. 101 electric valve 101, No. 102 electric valve 102, No. 105 electric valve 105 and No. 106 electric valve 106 cannot be opened normally, the operator of the main control room will immediately receive an indication signal that the corresponding electric isolation valve is not opened. When receiving an indication signal that electric valve No. 101 or electric valve No. 102 is not open, the operator remotely controls electric valve No. 103 or electric valve No. 104 to open; similarly, when receiving an indication signal that electric valve No. 105 and electric valve No. 106 are not open, the operator remotely controls electric valve No. 107 or electric valve No. 108 to open.

[0083] Sampling waste liquid gamma dose rate measurement. After the operator in the main control room of the nuclear power plant remotely controls the electric valves of the inlet module 10 and the outlet module 11 of the pit waste liquid sampling pipeline to open, the operator notifies the relevant personnel through the internal network of the nuclear power plant, the in-plant broadcast or the telephone to carry out the waste liquid sampling and measurement work in the reactor plant pit a. After receiving the instructions from the main control room, the staff starts the waste liquid sampling and measurement work.

[0084] The staff opens the electric valve No. 202, the first electric sampling pump 203, the sampling cooler 204 in sequence through the electric valve controller 701 and the sampling pump controller 702 of the measurement system control platform 70, opens the electric valve No. 607 607, and remotely opens the electric valve No. 501 and the electric valve No. 503 503 of the injection module 50. The first electric sampling pump 203 extracts waste liquid from the reactor building pit a, and the waste liquid is injected into the reactor building pit b through electric valve No. 101 101 and electric valve No. 102 102 or electric valve No. 103 103 and electric valve No. 104 104, electric valve No. 202 202, the first electric sampling pump 203, the sampling cooler 204, electric valve No. 607 607, the injector 502 of the injection module 50 and electric valve No. 503 503, electric valve No. 105 105 and electric valve No. 106 106 or electric valve No. 107 107 and electric valve No. 108 108.

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

[0086] Sampling of radioactive waste liquid from the pit after the accident. The staff checked the measurement data of the gamma dose rate meter 201 through the data display of the monitoring module controller 703 of the measurement system control platform 70. After the measurement data of the gamma dose rate meter 201 stabilized, the staff Select appropriate waste liquid quantification, dilution and sampling methods for the waste liquid in reactor building pit a.

[0087] when When the measured data of the γ dose rate meter 201 is stable, the sampling mode 1 is used. When , sampling is performed according to sampling mode 1.

[0088] The staff remotely controls the electric isolation valve No. 316 electric valve 316, No. 303 electric valve 303, No. 604 electric valve 604, No. 603 electric valve 603, No. 605 electric valve 605, No. 306 electric valve 306, No. 317 electric valve 317, No. 304 electric valve 304, No. 312 electric valve 312, No. 601 electric valve 601 and No. 206 electric valve 206 in sequence through the electric valve controller 701 of the measurement system control platform 70, and remotely controls the second electric sampling pump 311 through the sampling pump controller 702. The No. 314 flowmeter 314 transmits its measurement data to the second electric sampling pump 311 in real time by telex mode, and the second electric sampling pump 311 automatically adjusts its sampling flow rate according to the feedback measurement data of the No. 314 flowmeter 314, ensuring that the sampling flow rate of the second electric sampling pump 311 is stably controlled at 100mL / min. Flowmeter No. 305 305 and flowmeter No. 318 318 feed back their measurement data to electric valve No. 304 304 and electric valve No. 317 317 respectively via telex. According to the measurement data of flowmeter No. 305 305 and flowmeter No. 318 318, electric valve No. 304 304 and electric valve No. 317 317 automatically adjust the valve opening to ensure that the flow rate of desalted 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 injection loop is flushed. The deionized water enters the second multi-way valve 301 through the No. 206 electric valve 206, the No. 207 flowmeter 207, and the No. 601 electric valve 601, and enters the waste liquid injection injection module 50 through the second liquid inlet 30101, the first connection end 30106 of the second quantitative ring, the 0.1mL first quantitative ring 302, the second connection end 30103 of the second quantitative ring and the port of the second liquid outlet 30102 and the No. 303 electric valve 303. The injector 502 of the injection module 50 sucks the waste liquid discharged from the second liquid outlet 30102, and injects it into the reactor plant pit b through the No. 503 electric valve 503, the No. 105 electric valve 105 and the No. 106 electric valve 106 or the No. 107 electric valve 107 and the No. 108 electric valve 108, forming a flushing loop.

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

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

[0092] The demineralized water is used to flush the radioactive waste liquid or radionuclide contamination on the pipe wall that may remain in the flushing loop at a flow rate of 100 mL / min, and the flushing time is 5 minutes. The flushing liquid enters the injection module 50. The injector 502 sucks, injects and discharges the waste liquid generated by the above flushing to the reactor plant pit b.

[0093] Sampling, dilution and quantification of samples are performed. After the deionized water is used to flush the flushing circuit for 5 minutes at a flow rate of 100 mL / min, the staff remotely controls the electric isolation valve No. 206 electric valve 206 and No. 607 electric valve 607 through the electric valve controller 701 of the measurement system control platform 70, and opens the No. 205 electric valve 205. The No. 207 flow meter 207 transmits its measurement data to the first electric sampling pump 203 in real time by telex. The first electric sampling pump 203 automatically adjusts its sampling flow rate according to the measurement data fed back by the No. 207 flow meter 207, ensuring that the sampling flow rate of the first electric sampling pump 203 is stably controlled at 100 mL / min, and the flow rate of the sample entering the second liquid inlet end 30101 is stabilized 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, and the waste liquid passes through the No. 101 electric valve 101 and the No. 102 electric valve 102 or the No. 103 electric valve 103 and the No. 104 electric valve 104, the No. 202 electric valve 202, the first electric sampling pump 203, the sampling cooler 204, the No. 205 electric valve 205, the No. 207 flow meter 207, and the No. 601 electric valve 601 into the second multi-way valve 301, and enters the injection module 50 through the second liquid inlet end 30101, the second quantitative ring first connecting end 30106, the 0.1mL first quantitative ring 302, the second quantitative ring second connecting end 30103 and the second liquid outlet end 30102 and the No. 303 electric valve 303, forming a sampling loop.

[0095] At this time, the second multi-way valve 301 is in the injection mode, the injection flow rate is 100mL / min, and the injection duration is 5min, ensuring that the radioactive concentration of the liquid at each location in the injection loop is the same as the waste liquid in the reactor plant pit a. Thus, the injection of 0.1mL of the waste liquid in the reactor plant pit a into the 0.1mL second quantitative loop 302 of the second multi-way valve 301 is completed, and at this time, the radioactive concentration of the liquid in the 0.1mL second quantitative loop 302 is the same as the waste liquid in the reactor plant pit a.

[0096] After the second multi-way valve 301 runs in the injection mode for 5 minutes, the staff remotely controls the second multi-way valve 301 to adjust from the injection mode to the quantitative mode through the electric valve controller 701 of the measurement system control platform 70, and remotely closes the second electric sampling pump 311 through the sampling pump controller 702. In the quantitative mode, the deionized water enters the second multi-way valve 301 through the No. 304 electric valve 304 and the No. 305 flowmeter 305, and enters the collection container 308 through the second diluent injection end 30104 and the second quantitative ring second connection end 30103, the 0.1mL first quantitative ring 302, the second quantitative ring first connection end 30106 and the collection container connection end 30105, the No. 306 electric valve 306, and the second quick connector 307, forming a quantitative sampling waste liquid dilution circuit. Flow meter No. 305 feeds back its measurement data to electric valve No. 304 via telex. According to the measurement data of flow meter No. 305, electric valve No. 304 automatically adjusts the valve opening to ensure that the flow rate of desalted water entering the second diluent injection end 30104 is controlled at 100mL / 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 through the 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, and the electric valve controller 701 automatically closes the electric valve No. 304 304 and the electric valve No. 306 306 to ensure that the liquid level of the collection container 308 is 1000mL.

[0098] The flow rate of deionized water entering the second multi-way valve 301 is 100 mL / min, and the waste liquid with the same radioactive concentration as the waste liquid in the reactor plant pit a in the 0.1 mL second quantitative ring 302 is flushed to the collection container 308, and the flushing time is 10 minutes, so as to achieve the dilution of the 0.1 mL sample of the waste liquid in the reactor plant pit a. At this time, the radioactive concentration of the liquid in the collection container 308 is one ten-thousandth of the radioactive concentration of the waste liquid in the reactor plant pit a.

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

[0100] The No. 314 flowmeter 314 transmits its measurement data to the second electric sampling pump 311 in real time by telex. The second electric sampling pump 311 automatically adjusts its sampling flow rate according to the feedback measurement data of the No. 314 flowmeter 314, ensuring that the sampling flow rate of the second electric sampling pump 311 is stably controlled at 100mL / min and the injection flow rate of the first multi-way valve 320 is 100mL / min, and the injection duration is 5min, ensuring that the radioactivity concentration of the liquid at each location in the injection loop is the same as that of the liquid in the collection container 308. Thus, the injection of 1mL of liquid in the collection container 308 into the first multi-way valve 320 is completed. At this time, the radioactivity concentration of the liquid at each location in the injection loop is the same as that of the liquid in the collection container 308, which is one ten-thousandth of the radioactivity concentration of the waste liquid in the reactor plant pit a.

[0101] After completing the injection of 1 mL of the first quantitative loop 315 of the first multi-way valve 320, the staff remotely controls the closure of electric valve No. 603 and electric valve No. 316 through the electric valve controller 701 of the measurement system control platform 70, adjusts the first multi-way valve 320 from the injection mode to the quantitative mode, opens the electric valve No. 401, and at the same time, the sampling pump controller 702 remotely closes the second electric sampling pump 311.

[0102] like Figure 2As shown, the desalted water passes through the No. 317 electric valve 317, the No. 318 flowmeter 318, and enters the first multi-way valve 320, and enters the test liquid container 404 through the first diluent injection end 32004 and the second connection end 32003 of the first quantitative ring, the 1mL first quantitative ring 315, the first connection end 32006 of the first quantitative ring and the connection end 32005 of the test liquid container, the No. 604 electric valve 604 and the No. 401 electric valve 401, and the first quick connector 403. Flow meter No. 318 feeds back its measurement data to electric valve No. 317 via telex. Based on the measurement data of flow meter No. 318, electric valve No. 317 automatically adjusts the valve opening to ensure that desalted water enters the first multi-way valve 320 through electric valve No. 317 to flush 1 mL of liquid in the first quantitative ring 315 (the radioactivity concentration of the liquid is one ten-thousandth of the waste liquid in pit a of the reactor plant) into the tested liquid container 404 at a flow rate of 100 mL / min, and the flushing time is 1 min.

[0103] like Figure 3 As shown, the first liquid level monitoring module 405 transmits the liquid level data of the measured liquid container 404 to the monitoring module controller 703 of the measurement system control platform 70 in real time through the measuring cable. When the liquid level of the measured liquid container 404 reaches 100mL, the monitoring module controller 703 gives an alarm that the measured liquid container 404 reaches the measurement condition, and the alarm automatically closes after 30 seconds; at the same time, an alarm signal is sent to the electric valve controller 701, and the electric valve controller 701 automatically closes the No. 317 electric valve 317 and the No. 401 electric valve 401 to ensure that the liquid level of the measured liquid container 404 is 100mL. At this time, the radioactivity concentration of the liquid in the measured liquid container 404 is one millionth of the waste liquid in the reactor plant pit a.

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

[0105] The staff remotely controls the electric valve controller 701 of the measurement system control platform 70, and opens electric valve No. 316 316, electric valve No. 604 604, electric valve No. 603 603, electric valve No. 605 605, electric valve No. 317 317, electric valve No. 312 312, electric valve No. 602 602, electric valve No. 606 606 and electric valve No. 206 in sequence, and remotely opens the second electric sampling pump 311 through the sampling pump controller 702.

[0106] The No. 314 flowmeter 314 transmits its measurement data to the second electric sampling pump 311 in real time by telex. The second electric sampling pump 311 automatically adjusts its sampling flow rate according to the feedback measurement data of the No. 314 flowmeter 314, ensuring that the sampling flow rate of the second electric sampling pump 311 is stably controlled at 100mL / min. The No. 318 flowmeter 318 feeds back its measurement data to the No. 317 electric valve 317 by telex. According to the measurement data of the No. 318 flowmeter 318, the No. 317 electric valve 317 automatically adjusts the valve opening to ensure that the flow rate of the desalted water entering the first diluent injection end 32004 is controlled at 100mL / min.

[0107] Flushing of the sampling loop. The deionized water enters the collection container 308 through the No. 206 electric valve 206, the No. 207 flowmeter 207, the No. 602 electric valve 602 and the No. 606 electric valve 606, and the second quick connector 307. The liquid in the collection container 308 enters the first multi-way valve 320 through the third quick connector 310, the second electric sampling pump 311, the No. 312 electric valve 312, and the No. 314 flowmeter 314, and enters the injection module 50 through the first liquid inlet end 32001, the first connection end 32006 of the first quantitative ring, the 1mL first quantitative ring 315, the second connection end 32003 of the first quantitative ring, the first liquid outlet end 32002, and the No. 316 electric valve 316. The injector 502 of the injection module 50 sucks the waste liquid discharged from the first liquid outlet 32002, and injects it into the reactor building pit b through the No. 503 electric valve 503, the No. 105 electric valve 105 and the No. 106 electric valve 106 or the No. 107 electric valve 107 and the No. 108 electric valve 108, forming a flushing loop.

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

[0109] The demineralized water is used to flush the radioactive waste liquid or radionuclide contamination on the pipe wall that may remain in the flushing loop at a flow rate of 100 mL / min, and the flushing time is 5 minutes. The flushing liquid enters the injection module 50. The injector 502 sucks, injects and discharges the waste liquid generated by the above flushing to the reactor plant pit b.

[0110] Sampling and quantification of the sample. After the deionized water is flushed into the flushing loop at a flow rate of 100 mL / min for 5 minutes, the staff remotely controls the closing of the No. 206 electric valve 206 and the No. 607 electric valve 607 through the electric valve controller 701 of the measurement system control platform 70, and opens the electric isolation valve No. 205 electric valve 205. The No. 207 flow meter 207 transmits its measurement data to the first electric sampling pump 203 in real time by telex mode. The first electric sampling pump 203 automatically adjusts its sampling flow rate according to the measurement data fed back by the No. 207 flow meter 207, ensuring that the sampling flow rate of the first electric sampling pump 203 is stably controlled at 100 mL / min, ensuring that the flow rate of the sample entering the collection container 308 and the flow rate of the sample entering the first liquid inlet end 32001 are 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, and the waste liquid passes through electric valve No. 101 101 and electric valve No. 102 102 or electric valve No. 103 103 and electric valve No. 104 104, electric valve No. 202 202, the first electric sampling pump 203, the sampling cooler 204, electric valve No. 205 205, flow meter No. 207, electric valve No. 602 602, electric valve No. 606 606 and the second quick connector 307 into the collection container 308. The liquid in the collection container 308 enters the first multi-way valve 320 through the third quick connector 310, the second electric sampling pump 311, the 312 electric valve 312, and the 314 flowmeter 314, and enters the injection module 50 through the first liquid inlet 32001, the first connection end 32006 of the first quantitative ring, the 1mL first quantitative ring 315, the second connection end 32003 of the first quantitative ring, the first liquid outlet 32002, and the 316 electric valve 316, forming a sampling loop. At this time, the first multi-way valve 320 is in the sampling mode, the sampling flow rate is 100mL / min, and the sampling duration is 5min, ensuring that the radioactive concentration of the liquid at each location in the sampling loop is the same as the waste liquid in the reactor plant pit a. Thus, the sampling of 1mL of the waste liquid in the reactor plant pit a in the 1mL first quantitative ring 315 is completed, and the radioactive concentration of the liquid in the 1mL first quantitative ring 315 is the same as the waste liquid in the reactor plant pit a.

[0112] After the 1mL second quantitative loop 315 of the first multi-way valve 320 is injected, the staff remotely controls the No. 603 electric valve 603 to close through the electric valve controller 701 of the measurement system control platform 70, adjusts the first multi-way valve 320 from the injection mode to the quantitative mode, and simultaneously opens the No. 401 electric valve 401. The deionized water enters the first multi-way valve 320 through the No. 317 electric valve 317, the No. 318 flowmeter 318, and enters the first multi-way valve 320, and enters the test liquid container 404 through the first diluent injection end 32004 and the second connection end 32003 of the first quantitative loop, the 1mL first quantitative loop 315, the first connection end 32006 of the first quantitative loop and the connection end 32005 of the test liquid container, the No. 604 electric valve 604 and the No. 401 electric valve 401, and the first quick connector 403. Flow meter No. 318 feeds back its measurement data to electric valve No. 317 via telex. Based on the measurement data of flow meter No. 318, electric valve No. 317 automatically adjusts the valve opening to ensure that desalted water enters the first multi-way valve 320 through electric valve No. 317 to flush 1 mL of liquid in the first quantitative ring 315 (the radioactive concentration of the liquid is the same as the waste liquid in pit a of the reactor plant) into the tested liquid container 404 at a stable flow rate of 100 mL / min, and the flushing time is 1 min.

[0113] The first liquid level monitoring module 405 transmits the liquid level data of the measured liquid container 404 to the monitoring module controller 703 of the measurement system control platform 70 in real time through the measuring cable. When the liquid level of the measured liquid container 404 reaches 100mL, the monitoring module controller 703 gives an alarm that the measured liquid container 404 reaches the measurement condition, and the alarm automatically closes after 30 seconds; at the same time, an alarm signal is sent to the electric valve controller 701, and the electric valve controller 701 automatically closes the No. 317 electric valve 317 and the No. 401 electric valve 401 to ensure that the liquid level of the measured liquid container 404 is 100mL. At this time, the radioactivity concentration of the liquid in the measured liquid container 404 is one percent of the waste liquid in the reactor plant pit a.

[0114] when This is sampling mode 3. After the measured data of the γ dose rate meter 201 is stable, Figure 4 The measurement data of the γ dose rate meter 201 is shown as When , sampling is performed according to mode 3.

[0115] The staff remotely controls the electric valve controller 701 of the measurement system control platform 70 to open the electric valve No. 605, the electric valve No. 608, the electric valve No. 602, the electric valve No. 602 and the electric valve No. 206 in sequence.

[0116] The deionized water enters the injection module 50 through the No. 206 electric valve 206, the No. 207 flow meter 207, the No. 602 electric valve 602, the No. 608 electric valve 608 and the No. 605 electric valve 605, forming a flushing loop.

[0117] The deionized water is used to flush the radioactive waste liquid or radionuclide contamination on the pipe wall that may remain in the flushing loop at a flow rate of 100 mL / min, and the flushing time is 5 minutes. The flushing liquid enters the injection module 50. The ejector 502 of the injection module 50 sucks the waste liquid from the outlet of the No. 605 electric valve 605, and injects it into the reactor building pit b through the No. 503 electric valve 503, the No. 105 electric valve 105 and the No. 106 electric valve 106 or the No. 107 electric valve 107 and the No. 108 electric valve 108, forming a flushing loop.

[0118] Take samples. After the deionized water is used to flush the flushing circuit at a flow rate of 100 mL / min for 5 minutes, the staff remotely controls the closing of the No. 206 electric valve 206 and the No. 607 electric valve 607 through the electric valve controller 701 of the measurement system control platform 70, and opens the No. 205 electric valve 205. The No. 207 flow meter 207 transmits its measurement data to the first electric sampling pump 203 in real time by telex. The first electric sampling pump 203 automatically adjusts its sampling flow rate according to the measurement data fed back by the No. 207 flow meter 207, ensuring that the sampling flow rate of the first electric sampling pump 203 is stably controlled at 100 mL / min.

[0119] The first electric sampling pump 203 extracts waste liquid from the reactor plant pit a, and the waste liquid passes through the No. 101 electric valve 101 and the No. 102 electric valve 102 or the No. 103 electric valve 103 and the No. 104 electric valve 104, the No. 202 electric valve 202, the first electric sampling pump 203, the sampling cooler 204, the No. 205 electric valve 205, the No. 207 flow meter 207, the No. 602 electric valve 602, the No. 608 electric valve 608 and the No. 605 electric valve 605 into the injection module 50. The whole process lasts for 5 minutes, and the radioactivity concentration of the sample in the whole loop is the same as the waste liquid in the reactor plant pit a. The staff remotely controls the opening of the No. 402 electric valve 402 and the closing of the No. 608 electric valve 608 through the electric valve controller 701 of the measurement system control platform 70.

[0120] The first electric sampling pump 203 extracts waste liquid from the reactor building pit a, and the waste liquid passes through electric valve No. 101 101 and electric valve No. 102 102 or electric valve No. 103 103 and electric valve No. 104 104, electric valve No. 202 202, the first electric sampling pump 203, the sampling cooler 204, electric valve No. 205 205, flow meter No. 207 207, electric valve No. 602 602 and electric valve No. 402 402, 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 the reactor building pit a.

[0121] The first liquid level monitoring module 405 transmits the liquid level data of the measured liquid container 404 to the monitoring module controller 703 of the measurement system control platform 70 in real time through the measuring cable. When the liquid level of the measured liquid container 404 reaches 100mL, the monitoring module controller 703 gives an alarm that the measured liquid container 404 reaches the measurement condition, and the alarm automatically closes after 30 seconds; at the same time, an alarm signal is sent to the electric valve controller 701, and the electric valve controller 701 automatically closes the No. 317 electric valve 317 and the No. 401 electric valve 401 to ensure that the liquid level of the measured liquid container 404 is 100mL. At this time, the radioactivity concentration of the liquid in the measured liquid container 404 is the same as the waste liquid in the reactor plant pit a.

[0122] After the alarm of the measured liquid container 404 reaching the measurement condition is triggered, the staff checks whether the temperature of the high-purity germanium detector 406 reaches the stable working condition through the high-purity germanium spectrometer controller 707 of the measurement system control platform 70. If the working condition is not reached, the refrigerator 407 continues to cool the high-purity germanium detector 406 until its temperature reaches the low temperature condition of -175°C required for normal operation.

[0123] If the temperature of the high-purity germanium detector 406 has reached the low temperature condition for its normal operation, the staff will start the gamma spectrum measurement of the sample in the test liquid container 404 through the gamma spectrum analysis software 706 configured on the measurement system control platform 70. The high-purity germanium detector 406 will convert the measured gamma rays into corresponding voltage pulse amplitudes according to the energy size, and transmit the voltage signal to the gamma spectrum measurement system main amplifier 704 of the measurement system control platform 70 through the measurement cable to further amplify it into a voltage signal of 0 to 5V. The amplified voltage signal is processed into a digital signal by the multi-channel analyzer 705 to achieve analog-to-digital conversion. The multi-channel analyzer 705 transmits the digital signal to the gamma spectrum analysis software 706 to execute the collection of gamma spectrum data during the gamma spectrum measurement of the sample in the test liquid container 404. It is generally recommended to set the data collection time to 3600 seconds or 5000 seconds (the data collection time is adjustable). After the data collection is completed, the gamma spectrum analysis software 706 automatically completes the analysis of the gamma spectrum data and outputs the type i of the nuclide and the radioactivity information A of the sample in the test liquid container 404.i .

[0124] The radioactive concentration of the waste liquid in the reactor plant pit a is calculated. According to the corresponding sampling mode and in combination with the γ spectrum analysis software 706, the radionuclide type and radioactivity information A of the sample in the test liquid container 404 are automatically analyzed. i , the concentration of each radionuclide in the waste liquid in pit a of the reactor plant was calculated.

[0125] For sampling mode 1, that is, According to sampling mode 1, the radioactivity concentration of the liquid in the test liquid container 404 is one millionth of the waste liquid in the reactor plant pit a. Thus, the relationship between the concentration of each radioactive nuclide i in the waste liquid in the reactor plant pit a and the radioactivity concentration of each nuclide i in the liquid in the test liquid container 404 can be established:

[0126]

[0127] C i =10 10 ×A i (Bq / m 3 )

[0128] For sampling mode 2: that is, when 5μSv / h<·D≤5mSv / h. According to sampling mode 2, the radioactivity concentration of the liquid in the test liquid container 404 is one percent of the waste liquid in the reactor plant pit a. Thus, the relationship between the concentration of each radioactive nuclide i in the waste liquid in the reactor plant pit a and the radioactivity concentration of each nuclide i in the liquid in the test liquid container 404 can be established:

[0129]

[0130] C i =10 6 ×A i (Bq / m 3 )

[0131] For sampling mode 3, that is, According to sampling mode 3, the radioactivity concentration of the liquid in the test liquid container 404 is the same as the radioactivity concentration of the waste liquid in the reactor plant pit a. The concentration of each radionuclide i in the waste liquid in the reactor plant pit a is:

[0132]

[0133] C i =10 4 ×A i (Bq / m 3 )

[0134] Among them, A iThe radionuclide type i and its radioactivity are measured for the sample in the test liquid container 404, in Bq; C i is the radioactivity concentration of each radionuclide i in the wastewater in the reactor plant pit a, in Bq / m 3 ; Bq / mL and Bq / m 3 Both are units of radioactivity concentration.

[0135] The above is only a specific embodiment of the present invention, but the protection scope of the present invention is not limited thereto. Any technician familiar with the technical field can easily think of various equivalent modifications or replacements within the technical scope disclosed by the present invention, and these modifications or replacements should be included in the protection scope of the present invention. Therefore, the protection scope of the present invention shall be based on the protection scope of the claims.

Claims

1. A reactor building radioactive waste liquid measuring device, characterized in that: include: The measuring module comprises 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 arranged on the peripheral side of the test liquid container, and the radioactivity measurement and control module is arranged 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; The first quantitative sampling module is connected to the test liquid container and is used for quantitatively injecting radioactive waste liquid into the test liquid container.

2. The reactor building radioactive waste liquid measuring device according to claim 1, characterized in that: The measurement module includes a first shielding shell, the test liquid container is arranged in the first shielding shell, the radioactivity measurement and control module includes a high-purity germanium detector, a refrigerator, and a temperature and humidity controller, the high-purity germanium detector is arranged below the test liquid container, the refrigerator is connected to the high-purity germanium detector, and the temperature and humidity controller is arranged on the surrounding side of the high-purity germanium detector, wherein the first shielding shell is used to shield the environment and cosmic gamma rays, and the high-purity germanium detector is used to convert the gamma rays of the radioactive waste liquid into electrical signals.

3. The reactor building radioactive waste liquid measuring device according to claim 2, characterized in that: The radioactivity measurement and control module also includes a signal conversion module, which is electrically connected to the high-purity germanium detector and is used to convert the electrical signal into a digital signal.

4. The reactor building radioactive waste liquid measuring device according to any one of claims 1 to 3, characterized in that: The first quantitative injection module includes a first quantitative ring and a first multi-way valve, the first multi-way valve includes a first liquid inlet, a first liquid outlet, a first connection end of the first quantitative ring, a second connection end of the first quantitative ring, a first diluent injection end and a test liquid container connection end, and the first quantitative ring is used for quantitative sampling of radioactive waste liquid; wherein, when in the injection mode, the first liquid inlet, the first connection end of the first quantitative ring, the first quantitative ring, the first connection end of the first quantitative ring, and the first liquid outlet are connected in sequence; and / or, when in the quantitative mode, the first diluent injection end, the second connection end of the first quantitative ring, the first quantitative ring, the first connection end of the first quantitative ring, and the test liquid container connection end are connected in series in sequence.

5. A reactor building radioactive liquid sampling and measurement system, characterized in that: include: A radioactive liquid waste measuring unit, comprising a reactor building radioactive liquid waste measuring device as claimed in any one of claims 1 to 4; A liquid inlet unit, the liquid inlet unit is used to allow the radioactive waste liquid in the reactor building pit to pass through the containment shell and enter the sampling and measuring system; A liquid discharge unit connected to the radioactive waste liquid measurement unit, the liquid discharge unit being used to discharge the liquid in the sampling and measurement system through the containment into a reactor building pit; a radioactive waste liquid sampling unit connected to the liquid inlet unit and the radioactive waste liquid measuring unit, the radioactive waste liquid sampling unit being used to sample the radioactive waste liquid; A control unit is electrically connected to the radioactivity measurement unit, the radioactive waste liquid sampling unit, the liquid inlet unit and the liquid discharge unit, and the control unit is used to control the operation of the sampling and measurement system.

6. The reactor building radioactive liquid sampling and measuring system according to claim 5, characterized in that: The radioactive waste liquid sampling unit includes a second quantitative sampling module and a sampling module, the sampling module includes a collecting container, a second shielding cover and a second liquid level monitoring module, the collecting container is arranged in the second shielding cover, the second liquid level monitoring module is arranged on the surrounding side of the collecting container, the second quantitative sampling module is arranged outside the second shielding cover, the second quantitative sampling module is connected to the input end of the collecting container, and the output end of the collecting container is connected to the first quantitative sampling module, wherein the second shielding cover is used to shield the gamma rays released by the radioactive waste liquid in the collecting container.

7. The reactor building radioactive liquid sampling and measuring system according to claim 6, characterized in that: The second quantitative injection module includes a second quantitative ring and a second multi-way valve, the second multi-way valve includes a second liquid inlet, a second liquid outlet, a first connection end of the second quantitative ring, a second connection end of the second quantitative ring, a second diluent injection end and a collection container connection end, and the second quantitative ring is used for quantitative sampling of radioactive waste liquid; wherein, when in the injection mode, the second liquid inlet, the first connection end of the second quantitative ring, the second quantitative ring, the second connection end of the second quantitative ring, and the second liquid outlet are connected in sequence; and / or, when in the quantitative mode, the second diluent injection end, the second connection end of the second quantitative ring, the second quantitative ring, the first connection end of the second quantitative ring, and the collection container connection end are connected in series in sequence.

8. The reactor building radioactive liquid sampling and measuring system according to claim 7, characterized in that: The drainage unit includes an injection module and a drainage module, the output end of the injection module is connected to the drainage module, and the input end of the injection module is connected to the first quantitative injection module and the second quantitative injection module, wherein the drainage module is used to discharge the liquid in the sampling system through the containment into the reactor building pit, and the injection module is used to pressurize the liquid to be discharged from the sampling system.

9. The reactor building radioactive liquid sampling and measuring system according to claim 7, characterized in that: The liquid inlet unit includes a liquid inlet module and a flow control module, the input end of the flow control module is connected to the liquid inlet module, and the output end of the flow control module is connected to the input end of the radioactive waste liquid measuring unit and the second liquid inlet end, wherein the liquid inlet module 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 is used to control the flow rate of the liquid in the sampling and measurement system.

10. The reactor building radioactive liquid sampling and measuring system according to any one of claims 5 to 9, characterized in that: It also includes a flushing module, which is connected to the radioactive waste liquid sampling unit, the radioactive waste liquid measuring unit and the drainage unit, and is used to flush the sampling and measuring system.

11. A method for sampling and measuring radioactive waste liquid in a reactor building, characterized in that: A reactor building radioactive liquid sampling and measurement system applied to any one of claims 5 to 10, comprising: Opening the liquid inlet unit and the liquid discharge unit to allow the radioactive waste liquid in the reactor building pit to enter the reactor building radioactive liquid sampling and measuring system; Obtaining a gamma dose caused by the radioactive waste liquid in the reactor building pit by means of a gamma dose rate meter, and selecting a sampling mode according to the gamma 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 measurement unit to measure the nuclide types and radioactivity concentrations of the radioactive waste liquid according to the sampling mode; The liquid inlet unit and the liquid discharge unit are closed, and sample liquid is taken from the radioactive waste liquid sampling unit.

12. The method for sampling and measuring radioactive waste liquid from a reactor building according to claim 11, characterized in that: The step of starting the radioactive waste liquid sampling unit and sampling 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, a first sampling mode is used, and the radioactive waste liquid sampling unit dilutes and 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 the second preset value, a second sampling mode is used, and the radioactive waste liquid sampling unit dilutes and 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, a third sampling mode is used, and the radioactive waste liquid sampling unit dilutes and samples the radioactive waste liquid according to a third preset dilution ratio of the third sampling mode.

13. The method for sampling and measuring radioactive waste liquid from a reactor building according to claim 12, characterized in that: The step of starting the radioactive waste liquid measurement unit and measuring the type of nuclides and the radioactivity concentration of the radioactive waste liquid according to the sampling mode comprises: Inject radioactive waste liquid into the test liquid container, and measure the type of nuclides and radioactivity information of the radioactive waste liquid through the radioactivity measurement and control module; When the gamma dose is greater than a first preset value, the concentration of each radioactive nuclide in the waste liquid in the reactor plant pit is calculated according to the first preset formula of the first sampling mode and the radioactivity information of each nuclide; When the gamma 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 plant pit is calculated according to the second preset formula of the second sampling mode and the radioactivity information of each radionuclide; When the gamma dose is less than the second preset value, the concentration of each radioactive nuclide in the waste liquid in the reactor plant pit is calculated according to the third preset formula of the third sampling mode and the radioactivity information of each nuclide.

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