Uncontrol measuring device and method for waste metal frame of nuclear power station

Abstract

The invention discloses a control-releasing measuring device for a waste metal frame of a nuclear power station. The control-releasing measuring device comprises a shielding box, a scintillator detection assembly arranged in the shielding box and a containing part used for containing the waste metal frame. The shielding box comprises a box body with an opening in the top end and a box cover used for sealing the opening of the box body; the scintillator detection assembly comprises four scintillator probes, the four scintillator probes define a rectangular frame structure, the scintillator detection assembly is arranged in the box body, and the containing part is located on the upper portion of the scintillator detection assembly. The control-releasing measuring device is small in size and convenient to carry and transport. The device is simple in structure and compact in structure, can be suitable for various experiment scenes including laboratory interior and field environment, can be easily carried no matter daily experiment is carried out in a laboratory or field measurement is carried out in the field, and does not occupy too much space. By means of the portable design, a user can use the instrument more flexibly, and needed measurement can be carried out anytime and anywhere.

Description

Technical Field

[0001] The invention relates to the technical field of radioactive scrap metal frame processing, and in particular to a control and measurement device and method for scrap metal frames of nuclear power plants. Background Art

[0002] my country's nuclear power plants produce a large number of scrapped air filters every year. Since they are only slightly contaminated with radioactivity and have a low level of radioactive contamination, most of their scrap metal frames can be separated from the radioactive waste supervision system through clean clearance operations according to clearance requirements and reused as general materials. At the same time, the storage of scrap metal can reduce the cost of radioactive waste management in nuclear power plants and improve the economic benefits of nuclear power plants. In the clearance process, it is crucial to measure the activity concentration of nuclides on scrap metal frames.

[0003] When measuring the activity concentration of nuclides on scrap metal frames, portable radiation measuring instruments (handheld radiation detectors or radiation measuring instruments) are used to measure the metal frames. However, for extremely low-level radioactive waste, such as Co-60 < 0.1Bq, it is usually difficult to detect, and there is a problem of insufficient sensitivity in detecting extremely low-level radioactive waste. HPGe gamma spectrometers are also used for measurement, which can detect extremely low-level radioactive waste, but sampling and transportation are required on site, and workers need to undergo thorough screening before leaving the factory after sampling (all countries have strict laws and regulations on the operation of nuclear facilities and the management of radioactive materials, requiring that all personnel and items leaving the controlled area must meet the requirements of radioactivity levels below the limit. Screening is one of the important links to ensure the compliance of nuclear power plants). It is troublesome and time-consuming. In addition, the sensitivity of these devices may vary due to different environments, resulting in differences in detection accuracy. Summary of the invention

[0004] One of the purposes of the present invention is to provide a control and measurement device for waste metal frames of nuclear power plants, so as to solve the problems of low sensitivity in detecting low-level radioactive waste and complicated measurement in existing nuclide measurement devices for waste metal frames.

[0005] The second purpose of the present invention is to provide a control and measurement method for waste metal frames of nuclear power plants to solve the problems of low sensitivity in detecting low-level radioactive waste and complicated measurement in existing nuclide measurement devices for waste metal frames.

[0006] To solve the above problems, one of the purposes of the present invention is achieved as follows:

[0007] The present invention discloses a control and measurement device for a waste metal frame of a nuclear power plant, comprising: a shielding box, a scintillator detection assembly arranged in the shielding box, and a receiving portion for placing the waste metal frame;

[0008] The shielding box includes a box body with an open top and a box cover for closing the open top of the box body; the scintillator detection assembly includes four scintillator probes, which together form a rectangular frame structure; the scintillator detection assembly is arranged in the box body, and the accommodating portion is located at the upper part of the scintillator detection assembly.

[0009] Wherein, each of the scintillator probes is in a strip or columnar structure.

[0010] Wherein, the ratio of the length, width and height of the rectangular frame structure of the scintillator detection assembly is between 5:6:1 and 7:8:1.

[0011] Each of the scintillator probes includes a plastic scintillator, a light guide, and a photomultiplier tube. The plastic scintillator is in the shape of a strip or a column. The outer periphery of the light guide tapers axially. The large diameter end of the light guide is connected to one axial end of the plastic scintillator, and the small diameter end of the light guide is connected to the photomultiplier tube.

[0012] The outer peripheral side of the scintillator detection assembly is fitted with the inner side wall of the box body, and the bottom surface of the scintillator detection assembly is fitted with the bottom surface of the box body.

[0013] Among them, a shielding partition is arranged at a position of the box body on the inner peripheral side of the scintillator detection assembly, and the shielding partition is a horizontally arranged plate-like structure, and the shielding partition is fixedly connected to the bottom of the box body, and the top surface of the shielding partition is equal to or higher than the top surface of the scintillator detection assembly.

[0014] Wherein, a protective diaphragm is arranged between the accommodation portion and the scintillator detection assembly.

[0015] The second object of the present invention is achieved in this way:

[0016] A method for measuring and controlling a waste metal frame of a nuclear power plant according to the present invention adopts the above-mentioned device for measuring and controlling a waste metal frame of a nuclear power plant, and comprises the following steps:

[0017] a. Install the control and measurement device and connect it to the host computer control system;

[0018] b. Perform background measurement: close the box cover to the open end of the box body to seal the shielding box, perform background measurement, and record the background signal reference value after the measurement is completed;

[0019] c. Performing sample measurement: placing the sample to be measured in the containing portion of the shielding box, performing sample measurement, and after the sample measurement is completed, recording the sample measurement result;

[0020] d. Perform data analysis based on the measurement results: subtract the background signal reference value obtained in step b from the sample measurement results in step c to obtain a pure sample signal, and perform data analysis and processing on the pure sample signal.

[0021] Wherein, in step b, the specific process of performing background measurement is:

[0022] b1. Determine the measurement environment in the no-load state: the shielding box in the no-load state is buckled and sealed under no-load conditions, and other samples to be measured around the shielding box are removed;

[0023] b2. Setting measurement parameters, including measurement time, which is at least 2 minutes;

[0024] b3. Perform background measurement and record the background signal baseline value obtained.

[0025] Wherein, the box body and the box cover of the shielding box are both made of lead plates.

[0026] The beneficial effects of the present invention are:

[0027] When using the control and measurement device of the present invention to measure the nuclide activity concentration of the scrap metal frame of the nuclear power plant, the control and measurement device is connected to the host computer control system, and the scrap metal frame only needs to be loaded into the receiving part of the shielding box 1, the box cover is closed, and then the host computer control system starts the measurement command to perform the overall measurement of the scrap metal frame. Compared with the traditional measurement method, the control and measurement device of the present invention has the advantage of simple operation. In addition, during the measurement, there is no need to perform destructive sampling and sample preparation on the scrap metal frame to be measured, and the non-destructive overall detection of the scrap metal frame is realized. In addition, no other supplies or reagents need to be used during the measurement process, and no secondary waste will be generated, thereby reducing pollution. BRIEF DESCRIPTION OF THE DRAWINGS

[0028] The present invention will be further described in detail below in conjunction with the accompanying drawings and specific embodiments.

[0029] Figure 1 It is a schematic diagram of the structure of the present invention after removing the protective diaphragm;

[0030] Figure 2 It is a structural schematic diagram of a scintillator detection assembly;

[0031] Figure 3 for Figure 2 A top view of

[0032] Figure 4 Schematic diagram of the structure of the scintillator probe.

[0033] Description of Reference Numerals

[0034] 1. Shielding box; 11. Box body; 12. Box cover; 13. Receptacle; 14. Shielding partition; 2. Scintillator probe; 21. Plastic scintillator; 22. Light guide; 23. Photomultiplier tube; 3. Host computer control system; 4. Scrap metal frame. DETAILED DESCRIPTION

[0035] The technical scheme in the embodiment of the present invention is described clearly and completely below in conjunction with the accompanying drawings in the embodiment of the present invention. In the following description, many specific details are set forth to facilitate a full understanding of the present invention, but the present invention can also be implemented in other ways different from those described herein, and those skilled in the art can make similar generalizations without violating the connotation of the present invention, so the present invention is not limited by the specific embodiments disclosed below.

[0036] Embodiment 1:

[0037] like Figure 1-Figure 3 As shown, a control and measurement device for a waste metal frame 4 of a nuclear power plant of the present invention comprises: a shielding box 1, a scintillator detection assembly arranged in the shielding box 1, and a receiving portion 13 for placing the waste metal frame 4.

[0038] The shielding box 1 includes a box body 11 with an open top, and a box cover 12 for closing the open portion of the box body 11. The scintillator detection assembly includes four scintillator probes 2, which together form a rectangular frame structure. The scintillator detection assembly is arranged in the box body 11, and the receiving portion 13 is located at the upper part of the scintillator detection assembly.

[0039] When the control and measurement device of the present invention is used to measure the nuclide activity concentration of the scrap metal frame 4 of the nuclear power plant, the control and measurement device is connected to the host computer control system, and the scrap metal frame 4 only needs to be loaded into the receiving portion 13 of the shielding box 1, the box cover 12 is closed, and then the host computer control system starts the measurement command to perform the overall measurement of the scrap metal frame 4. Compared with the traditional measurement method, the control and measurement device of the present invention has the advantage of simple operation. In addition, during the measurement, there is no need to perform destructive sampling and sample preparation on the scrap metal frame 4 to be measured, and the non-destructive overall detection of the scrap metal frame 4 is realized. In addition, no other supplies or reagents need to be used during the measurement process, and no secondary waste is generated, thereby reducing pollution.

[0040] like Figure 4 As shown, each scintillator probe 2 is in a strip or columnar structure. The strip or columnar structure design can cover a larger area, help increase the detection volume, improve the capture capability of low-activity radionuclides, and thus improve the detection efficiency of gamma rays.

[0041] Each scintillator probe 2 includes a plastic scintillator 21, a light guide 22, and a photomultiplier tube 23. The plastic scintillator 21 is in the shape of a strip or a column. The outer periphery of the light guide 22 gradually shrinks along the axial direction. The large diameter end of the light guide 22 is connected to one end of the plastic scintillator 21 in the axial direction, and the small diameter end of the light guide 22 is connected to the photomultiplier tube 23. Among them, the four plastic scintillators 21 enclose a rectangular frame structure of the scintillator detection assembly, and the length, width and height ratio of the rectangular frame structure of the scintillator detection assembly is between 5:6:1 and 7:8:1. This length, width and height ratio design can realize efficient detection of gamma rays, optimize the detection efficiency, and facilitate reasonable arrangement in a multi-probe array (such as the four scintillator probes 2 in the present invention) to optimize the overall detection sensitivity and spatial coverage, so as to take into account the sensitivity and detection efficiency of low-level radioactive waste, while ensuring the compactness and operability of the equipment. The volume of the plastic scintillator 21 is equal to or greater than 0.5L. In the present embodiment, the volume of the plastic scintillator 21 is equal to 0.5L. Using four plastic scintillators 21 with a volume of 0.5L is equivalent to expanding the total detection volume to 2L. The larger detection volume significantly improves the probability of interaction of the detector with gamma rays, and can capture more gamma rays from low-intensity radionuclides, thereby improving the detection efficiency.

[0042] In the present embodiment, the material of the plastic scintillator 21 is polystyrene, and a fluorescent agent (such as p-Terphenyl or POPOP) is added to the polystyrene to improve the fluorescence efficiency and wavelength matching. In order to prevent the inconsistency of the internal light output from affecting the detection result, the fluorescent agent is evenly dispersed in the polystyrene. The surface of the plastic scintillator 21 is polished and wrapped with a high-reflection film. The high-reflection film is made of a high-reflection material to maximize the light collection efficiency. In the present embodiment, the high-reflection film is made of aluminum foil. In other embodiments, the high-reflection film can be made of a white reflective film. The plastic scintillator 21 is made of polystyrene, which has good light transmittance, can efficiently transmit the light signal generated by gamma-ray excitation, is easy to process into the desired shape, is light in weight, and is convenient for on-site deployment. Figure 4 As shown, in the present embodiment, each plastic scintillator 21 is in a strip or columnar structure, and each plastic scintillator 21 is in a rectangular long columnar structure, that is, the cross section of the scintillator probe 2 perpendicular to the axial direction is rectangular. This structural design has certain advantages: it helps to increase the detection volume, thereby improving the detection efficiency of gamma rays; it can cover a larger area and improve the capture capability of low-activity radionuclides; it is convenient for Monte Carlo simulation and optimization design of detection efficiency, and can enhance the detection range in a certain direction while optimizing space utilization.

[0043] The outer periphery of the light guide portion 22 tapers in the axial direction, the large diameter end of the light guide portion 22 is connected to one end of the plastic scintillator 21 in the axial direction, and the small diameter end of the light guide portion 22 is connected to the photomultiplier tube 23. The light guide portion 22 is a conical transition structure, which can play a role in guiding light, and converge the light signal generated in the long columnar plastic scintillator 21 to the receiving surface of the photomultiplier tube 23, thereby reducing light loss. The light guide portion 22 is a conical transition structure, which can reduce the reflection and scattering of light at the connection point compared to the design of direct planar connection, thereby improving the light collection efficiency.

[0044] The end surface of the light guide portion 22 connected to the photomultiplier tube 23 is circular, which facilitates efficient optical coupling with the photomultiplier tube 23, can better match the receiving window of the photomultiplier tube 23, provide a firm mechanical connection, and reduce manufacturing complexity.

[0045] like Figure 1-Figure 3 As shown, in order to enhance the shielding effect of the shielding box 1, the box body 11 and the box cover 12 of the shielding box 1 are made of lead plate, and the thickness of the box body 11 and the box cover 12 is equal to or greater than 5mm. In this embodiment, the thickness of the box body 11 and the box cover 12 is equal to 5mm. The outer peripheral side of the scintillator detection assembly is arranged in affixed with the inner wall of the box body 11, and the bottom surface of the scintillator detection assembly is arranged in affixed with the bottom surface of the box body 11. On the one hand, the arrangement of the scintillator detection assembly can be reasonably planned, and on the other hand, other substances can be avoided as much as possible from interfering with the measurement of the scintillator detection assembly. The four scintillator probes 2 are arranged around the box body of the shielding box 1, which can achieve multi-directional coverage and reduce the signal loss caused by absorption or scattering of rays. Each scintillator probe 2 receives gamma rays from different directions, which effectively reduces the geometric efficiency loss in the measurement. In the measurement environment, the plastic scintillator 21 is placed in the shielding box 1, which can effectively shield the interference of environmental background radiation, thereby reducing the influence of false signals. Under conditions of low background radiation, the statistical significance of weak signals (Signal-to-Noise Ratio, SNR) is improved, thereby improving the sensitivity and lower limit of detection.

[0046] Among them, a shielding baffle 14 is provided at a position of the box body 11 on the inner peripheral side of the scintillator detection assembly. The shielding baffle 14 is a horizontally arranged plate-like structure. The shielding baffle 14 is fixedly connected to the bottom of the box body 11, and the top surface of the shielding baffle 14 is equal to or higher than the top surface of the scintillator detection assembly. In this embodiment, the shielding baffle 14 is a horizontally arranged rectangular plate, and the shielding baffle 14 is integrally connected to the bottom of the box body 11. The top surface of the shielding baffle 14 is equal to or slightly higher than the top surface of the scintillator detection assembly. The shielding baffle 14 is made of lead plate, which can prevent other substances from interfering with the measurement of the scintillator detection assembly. A protective membrane (not shown) is provided between the receiving portion 13 and the scintillator detection assembly. The protective membrane is laid between the bottom of the receiving portion 13 and the top of the scintillator detection assembly. The protective membrane can be an annular sheet structure, which is only provided on the top surface of the scintillator detection assembly, or a rectangular sheet structure, which is laid on the top surface of the shielding partition 14. The four sides of the protective membrane extend horizontally from the top surface of the shielding partition 14 to the inner side wall of the corresponding side of the box body 11, which can prevent the scintillator detection assembly from directly contacting the scrap metal frame to be tested. After the two are in direct contact, the scintillator detection assembly is easily contaminated, affecting the measurement accuracy of other samples to be tested, and reducing the wear of the scintillator detection assembly. In this embodiment, the receiving portion 13 is the portion of the box body 11 located above the scintillator detection assembly.

[0047] Embodiment 2

[0048] like Figure 1-Figure 4 As shown, a method for controlling and measuring a waste metal frame 4 of a nuclear power plant according to the present invention adopts the device for controlling and measuring a waste metal frame 4 of a nuclear power plant according to the first embodiment, and comprises the following steps:

[0049] a. Install the control and measurement device, and connect the control and measurement device to the host computer control system 3. In this embodiment, the host computer in the host computer control system 3 can be a notebook computer.

[0050] b. Perform background measurement: Cover the box cover 12 to the open end of the box body 11 to seal the shielding box 1, perform background measurement, and record the background signal reference value after the measurement. One or more background measurements can be performed before performing sample measurement.

[0051] c. Perform sample measurement: Place the sample to be measured in the receiving portion 13 of the shielding box 1, perform sample measurement, and after the sample measurement is completed, record the sample measurement result. The same sample needs to be measured at least three times to improve the accuracy of the measurement.

[0052] d. Perform data analysis based on the measurement results: subtract the background signal reference value obtained in step b from the sample measurement results in step c to obtain a pure sample signal, and perform data analysis and processing on the pure sample signal.

[0053] Wherein, in step b, the specific process of performing background measurement is:

[0054] b1. Determine the measurement environment in the no-load state: The shielding box 1 in the no-load state is sealed and other samples to be tested are removed from the shielding box 1. That is, determine a measurement environment in the no-load state, that is, a state where no objects to be tested (such as radioactive scrap metal frames) are placed near the control and measurement device. This environment should be free of external interference as much as possible to ensure that only background radiation is measured.

[0055] b2. Setting measurement parameters, including measurement time, which is at least 2 minutes. In the present invention, the measurement time is at least 2 minutes, which can ensure that enough statistical data can be obtained.

[0056] b3. Perform background measurement and record the background signal baseline value obtained. The data recorded during the background measurement is usually the spectrum of gamma rays or the total count rate, based on which the radiation level of the environmental background can be calculated.

[0057] The box body 11 and the box cover 12 of the shielding box 1 are both made of lead plates.

[0058] The measuring method of the present invention can extract weaker gamma-ray signals from noise through spectrum analysis, background subtraction and signal enhancement algorithms, thereby further reducing the detection limit.

[0059] The patent of this invention has the following advantages:

[0060] (i) Simple operation and fixed process: Compared with the traditional measurement method, the measurement method of the present invention is simple to operate. It only needs to put the waste metal frame 4 of the nuclear power plant into the shielding box 1, connect the host computer control system 3 and click to start measurement to perform the overall measurement.

[0061] (ii) Non-destructive measurement: The measurement method of the present invention uses four plastic scintillators 21 with a volume of 0.5 L to directly measure the gamma rays emitted by the radioactive waste. There is no need to perform destructive sampling and sample preparation on the radioactive waste to be measured. There is no need to use other supplies or reagents during the measurement process, and no secondary waste is generated.

[0062] (III) Four 0.5L plastic scintillators 21 improve the detection limit: This method uses four scintillator probes 2 to measure the radioactive nuclides of the scrap metal frame 4 in the shielding box 1. With the existence of the shielding box 1 and the larger detection volume, the detection efficiency under the same conditions is significantly increased. In the analysis of the measurement results, multiple detections of the same material improve the accuracy of the measurement.

[0063] (IV) Reducing measurement time by subtracting background: Before the formal measurement, one or more background measurements are performed to obtain a baseline value of the background signal. After the sample measurement is completed, the background signal is subtracted from the measurement result to obtain the pure sample signal, thereby significantly reducing the measurement time.

[0064] (V) The control and measurement device of the present invention is small and easy to carry and transport. It has a compact structure and can be applied to various experimental scenarios, including laboratory and field environments. Whether it is for daily experiments in the laboratory or field measurements in the field, the measurement device can be easily carried and does not take up too much space. This portable design enables users to use the measurement device more flexibly and perform the required measurements anytime and anywhere.

[0065] In addition, the control and measurement device and control and measurement method of the present invention have been verified in actual scenarios including Hainan Nuclear Power and Jiangsu Nuclear Power. The test results show that they can accurately detect extremely low radioactive contamination with Co-60 concentration less than 0.1Bq. Compared with traditional scintillation detector equipment, the sensitivity is significantly improved.

[0066] Obviously, the described embodiments are only some embodiments of the present invention, not all 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.

Claims

1. A control and measurement device for waste metal frames in nuclear power plants, characterized in that: include: A shielding box, a scintillator detection assembly disposed in the shielding box, and a receiving portion for placing a scrap metal frame; The shielding box includes a box body with an open top and a box cover for closing the open top of the box body; the scintillator detection assembly includes four scintillator probes, which together form a rectangular frame structure; the scintillator detection assembly is arranged in the box body, and the accommodating portion is located at the upper part of the scintillator detection assembly.

2. A control and measurement device for waste metal frames of nuclear power plants according to claim 1, characterized in that: Each of the scintillator probes has a strip or column structure.

3. The control and measurement device for waste metal frames of nuclear power plants according to claim 1, characterized in that: The ratio of the length, width and height of the rectangular frame structure of the scintillator detection assembly is between 5:6:1 and 7:8:

1.

4. The control and measurement device for waste metal frames of nuclear power plants according to claim 2, characterized in that: Each of the scintillator probes comprises a plastic scintillator, a light guide, and a photomultiplier tube. The plastic scintillator is in a strip or column shape. The outer periphery of the light guide tapers axially. The large diameter end of the light guide is connected to one axial end of the plastic scintillator, and the small diameter end of the light guide is connected to the photomultiplier tube.

5. The control and measurement device for waste metal frames of nuclear power plants according to claim 1, characterized in that: The outer peripheral side of the scintillator detection assembly is arranged in contact with the inner side wall of the box body, and the bottom surface of the scintillator detection assembly is arranged in contact with the bottom surface of the box body.

6. The control and measurement device for waste metal frames of nuclear power plants according to claim 1, characterized in that: A shielding partition is arranged at a position of the box body on the inner peripheral side of the scintillator detection assembly. The shielding partition is a horizontally arranged plate-like structure. The shielding partition is fixedly connected to the bottom of the box body. The top surface of the shielding partition is equal to or higher than the top surface of the scintillator detection assembly.

7. The control and measurement device for waste metal frames of nuclear power plants according to claim 1, characterized in that: A protective diaphragm is arranged between the accommodation portion and the scintillator detection assembly.

8. A method for monitoring and measuring waste metal frames in nuclear power plants, characterized in that: The control and measurement device for a waste metal frame of a nuclear power plant according to any one of claims 1 to 7 comprises the following steps: a. Install the control and measurement device and connect it to the host computer control system; b. Perform background measurement: close the box cover to the open end of the box body to seal the shielding box, perform background measurement, and record the background signal reference value after the measurement is completed; c. Performing sample measurement: placing the sample to be measured in the containing portion of the shielding box, performing sample measurement, and after the sample measurement is completed, recording the sample measurement result; d. Perform data analysis based on the measurement results: subtract the background signal reference value obtained in step b from the sample measurement results in step c to obtain a pure sample signal, and perform data analysis and processing on the pure sample signal.

9. A method for monitoring and measuring waste metal frames in nuclear power plants according to claim 8, characterized in that: In step b, the specific process of performing background measurement is: b1. Determine the measurement environment in the no-load state: the shielding box in the no-load state is buckled and sealed under no-load conditions, and other samples to be measured around the shielding box are removed; b2. Setting measurement parameters, including measurement time, which is at least 2 minutes; b3. Perform background measurement and record the background signal baseline value obtained.

10. A method for monitoring and measuring waste metal frames in nuclear power plants according to claim 8, characterized in that: The box body and the box cover of the shielding box are both made of lead plates.

Images