Neutron dose and internal pollution diagnosis method and diagnostic apparatus

Abstract

The invention relates to the technical field of neutron dose and internal pollution diagnosis, and provides a neutron dose and internal pollution diagnosis method and diagnostic agent.The method comprises the following steps that gamma ray signals are obtained at the head top and the lung of a human body through a detection device, and gamma rays are converted into electric pulse signals through the detection device; a digital multi-channel signal processing device is used for conducting signal processing on the electric pulse signals, signal processing comprises digital filtering and peak value sampling, and energy spectrum data are obtained; analyzing the energy spectrum data, and respectively obtaining a Na-24 characteristic gamma photon peak value at the top of the head of the human body and a characteristic gamma photon peak value at the lung of the human body; and according to the Na-24 characteristic gamma photon peak value obtained at the top of the head, calculating the neutron exposure dose according to a neutron dose algorithm, and according to the characteristic gamma photon peak value obtained at the lung, determining the type and activity of the internal pollution nuclide. According to the invention, accurate calculation of neutron dose and identification and activity determination of internal pollution nuclide are realized, and dual functions of neutron exposure dose and internal pollution diagnosis can be provided at the same time.

Description

Technical Field

[0001] The present invention relates to the technical field of neutron dose and internal contamination diagnosis, and in particular to a neutron dose and internal contamination diagnosis method and a diagnostic instrument. Background Art

[0002] In the emergency response process of nuclear radiation accidents, it is crucial to quickly and accurately assess the neutron radiation dose received by personnel and the contamination of radionuclides in the body. When the human body is irradiated by neutrons, the sodium element in the body will be activated by neutrons to produce Na-24 and emit characteristic gamma rays. By measuring the gamma rays emitted by Na-24 above the head, the neutron dose received by the human body can be inferred. At the same time, when the human body is contaminated with radionuclides, these radionuclides are mainly enriched in the lungs in a short period of time. By measuring the gamma rays in the lungs, the type and activity of the internal contaminating nuclides can be quickly determined.

[0003] The existing technology usually adopts a separate measurement method, that is, using a neutron dose measuring instrument to measure the neutron exposure dose, and then using a whole-body counter to measure the activity of radionuclides in the body. This method has the following problems: first, it is necessary to equip multiple sets of different measurement equipment, which increases the equipment cost; second, it takes a long time to perform two measurements separately, which reduces the efficiency of emergency response. Due to the differences in measurement equipment and measurement time, the consistency and comparability of the measurement results are reduced. Summary of the invention

[0004] In view of this, the present invention proposes a neutron dose and internal contamination diagnosis method and diagnostic instrument, which solves the problem that the prior art requires two separate measurements, which is time-consuming and reduces the efficiency of emergency response, and the consistency and comparability of the measurement results are reduced due to differences in measuring equipment and measurement time.

[0005] The technical solution of the present invention is implemented as follows: On the one hand, the present invention provides a neutron dose and internal contamination diagnosis method, comprising the following steps:

[0006] Acquire gamma ray signals at the top of the head and lungs of the human body respectively through a detection device, and convert the gamma ray into an electrical pulse signal through the detection device;

[0007] Using a digital multi-channel signal processing device to process the electric pulse signal, the signal processing includes digital filtering and peak sampling, to obtain energy spectrum data;

[0008] Analyzing the energy spectrum data to obtain the characteristic gamma photon peak of Na-24 at the top of the human head and the characteristic gamma photon peak at the lungs;

[0009] Based on the Na-24 characteristic gamma photon peak value obtained at the top of the head, the neutron exposure dose is calculated according to the neutron dose algorithm, and the type and activity of the internal contamination nuclide are determined based on the characteristic gamma photon peak value obtained at the lungs.

[0010] On the basis of the above technical solution, preferably, the gamma ray signals are acquired at the top of the human head and the lungs respectively by the detection device, and the gamma ray is converted into an electrical pulse signal by the detection device, specifically including:

[0011] The detection device is placed at a preset distance from the top of the human head and the lungs to measure the gamma ray signal, and the measurement time of each measurement position does not exceed the preset time;

[0012] The detection device comprises a scintillator and a photomultiplier tube. The scintillator converts the gamma rays into scintillation photon pulse signals, and the photomultiplier tube converts the scintillation photon pulse signals into electrical pulse signals.

[0013] On the basis of the above technical solution, preferably, the electric pulse signal is processed by a digital multi-channel signal processing device, and the signal processing includes digital filtering and peak sampling to obtain energy spectrum data, specifically including:

[0014] Performing digital filtering processing on the electric pulse signal through an FPGA processing unit of a digital multi-channel signal processing device;

[0015] The FPGA processing unit includes a trigger module, a digital filter and a peak sampling module, the trigger module triggers signal acquisition, performs shaping filtering processing on the electrical pulse signal based on the digital filter, and extracts the pulse amplitude information of the electrical pulse signal through the peak sampling module;

[0016] Peak sampling and accumulation are performed on the pulse amplitude information to obtain a preset amount of energy spectrum data, wherein the energy spectrum data includes energy spectrum data obtained from the top of the head and energy spectrum data obtained from the lungs.

[0017] On the basis of the above technical solution, preferably, the energy spectrum data is analyzed to obtain the characteristic gamma photon peak of Na-24 at the top of the human head and the characteristic gamma photon peak at the lungs, respectively, specifically including:

[0018] The energy spectrum data obtained at the top of the head are analyzed. The energy value of the characteristic gamma photon of Na-24 is 1.36 MeV. In the energy spectrum measurement range of 80 keV to 3.0 MeV, the channel address of the peak of the characteristic gamma photon of Na-24 is located according to the average channel width of 2.8 keV, and the peak count is extracted;

[0019] The energy spectrum data obtained at the lungs are analyzed, and characteristic gamma photon peaks are located and extracted in all energy spectra based on characteristic gamma photon energy values ​​in a nuclide database, wherein the nuclide database includes characteristic gamma photon energy values ​​of fission and activation nuclides. The types of internal contamination nuclides are identified by comparing characteristic peaks in the energy spectrum data at the lungs with characteristic gamma photon energy values ​​in the nuclide database, and characteristic gamma photon peak counts corresponding to various internal contamination nuclides are extracted respectively.

[0020] On the basis of the above technical solution, preferably, the neutron exposure dose is calculated according to the neutron dose algorithm based on the Na-24 characteristic γ photon peak value obtained at the top of the head, and the type and activity of the internal contamination nuclide are determined according to the characteristic γ photon peak value obtained at the lungs, specifically including:

[0021] The neutron dose algorithm is used to calculate the peak value of Na-24 characteristic γ photons obtained at the top of the head to obtain the neutron exposure dose value;

[0022] The characteristic gamma photon peak value obtained at the lungs is compared and analyzed with the data in the nuclide database, which contains characteristic data of fission nuclides and activated nuclides. The type of internal contamination nuclides is determined by comparative analysis, and the activity of internal contamination nuclides is calculated based on the characteristic gamma photon peak counts.

[0023] On the basis of the above technical solution, preferably, the calculation formula of the neutron dose algorithm is:

[0024]

[0025] Where D is the neutron exposure dose value, F d is the neutron absorption conversion efficiency, λ is the decay parameter of Na-24, T is the time when the personnel is exposed to neutrons, t is the time interval from the end of exposure to the measurement, N(t) is the peak count of Na-24 characteristic γ photons obtained at time t, k1, k2, k3 are the first empirical correction coefficient, the second empirical correction coefficient, and the third empirical correction coefficient, respectively.

[0026] In a second aspect, the present invention further provides a neutron dose and internal contamination diagnostic instrument, the neutron dose and internal contamination diagnostic instrument comprising a detection device, a digital multi-channel signal processing device and a measuring cable, the digital multi-channel signal processing device being respectively connected to the detection device and the measuring cable signal;

[0027] The detection device is used to convert the gamma rays into electrical signals and perform preliminary amplification through a preamplifier;

[0028] The digital multi-channel signal processing device is used to receive and process the electrical signal from the detection device, perform signal analysis and energy spectrum processing through the digital spectrum analysis board, provide various required power supplies, and display the measurement results through the touch screen;

[0029] The measuring cable is used for signal transmission and power supply between the detection device and the digital multi-channel signal processing device through the signal output terminal, the preamplifier power supply terminal and the high-voltage input terminal.

[0030] On the basis of the above technical scheme, preferably, the detection device includes a LaBr3 scintillator, a photomultiplier tube, a charge-sensitive preamplifier, a voltage divider circuit and a shell. The shell is an aluminum cylindrical structure with a diameter of 8.5 cm and a thickness of 1.5 mm, and is provided with a vibration-proof and heat-insulating protective shell on the outside. The LaBr3 scintillator is a cylindrical structure with a diameter of 75 mm and a height of 75 mm, and is air-freely connected to the photomultiplier tube with an effective photosensitive area diameter of 70 mm through optical silicone oil.

[0031] On the basis of the above technical scheme, preferably, the digital multi-channel signal processing device adopts a double-layer circuit board structure, including an upper power supply board and a lower digital spectrum analysis board, the power supply board includes a DC-DC module and a high-voltage module for providing power; the digital spectrum analysis board includes an FPGA processing unit, a multi-channel analyzer and a memory for signal processing and energy spectrum analysis; and also includes a touch screen display module for displaying measurement results and performing human-computer interaction.

[0032] On the basis of the above technical solution, preferably, the measuring cable includes a signal output terminal, a preamplifier power supply terminal and a high-voltage input terminal, the signal output terminal is used to transmit the electrical signal output by the detection device, the preamplifier power supply terminal is used to power the preamplifier, and the high-voltage input terminal is used to provide high voltage for the photomultiplier tube.

[0033] The neutron dose and internal contamination diagnosis method and diagnostic instrument of the present invention have the following beneficial effects compared with the prior art:

[0034] (1) Through the detection device and digital multi-channel signal processing device, the gamma-ray signals from the human head and lungs are acquired and processed, and then converted into electrical pulse signals for energy spectrum analysis, thereby realizing the accurate calculation of neutron dose and the identification and activity measurement of internal contamination nuclides. It can provide the dual functions of neutron exposure dose and internal contamination diagnosis at the same time, and improve the efficiency of detection;

[0035] (2) By accurately locating and extracting the characteristic γ photon peak of Na-24 and the characteristic peaks of other internal contamination nuclides within the preset energy spectrum measurement range and combining the preset high-precision average track width, high-precision identification and quantitative analysis of neutron dose and internal contamination nuclides are achieved, thereby improving the reliability of measurement;

[0036] (3) By introducing empirical correction coefficients to correct neutron dose calculations and taking into account decay parameters and time factors, the dose calculation results can be flexibly adjusted according to different measurement conditions and environmental factors, thus improving the adaptability of neutron dose assessment.

[0037] (4) By integrating the detection device, digital multi-channel signal processing device and measuring cable, efficient γ-ray signal conversion and processing are achieved. The detection device adopts LaBr3 scintillator and photomultiplier tube to ensure high sensitivity and accurate signal detection. The double-layer circuit board structure of the digital multi-channel signal processing device provides stable power supply and efficient signal processing capability, and realizes human-computer interaction and result display through the touch screen. The measuring cable ensures the reliable transmission of signals and power supply, which improves the measurement accuracy, stability and operation convenience of the equipment and is suitable for fast and accurate neutron dose and internal contamination diagnosis. BRIEF DESCRIPTION OF THE DRAWINGS

[0038] In order to more clearly illustrate the embodiments of the present invention or the technical solutions in the prior art, the drawings required for use in the embodiments or the description of the prior art will be briefly introduced below. Obviously, the drawings described below are only some embodiments of the present invention. For ordinary technicians in this field, other drawings can be obtained based on these drawings without paying creative work.

[0039] Figure 1 A flow chart of a neutron dose and internal contamination diagnosis method of the present invention;

[0040] Figure 2 A structural diagram of a neutron dose and internal contamination diagnostic instrument of the present invention;

[0041] Figure 3 A power supply circuit diagram of the power board of the present invention;

[0042] Figure 4 This is a diagram of the internal structure of the FPGA signal processing unit of the present invention. DETAILED DESCRIPTION

[0043] The following will be combined with 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 only 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.

[0044] See also Figure 1 The present invention provides a neutron dose and internal contamination diagnosis method, comprising the following steps:

[0045] Acquire gamma ray signals at the top of the head and lungs of the human body respectively through a detection device, and convert the gamma ray into an electrical pulse signal through the detection device;

[0046] Using a digital multi-channel signal processing device to process the electric pulse signal, the signal processing includes digital filtering and peak sampling, to obtain energy spectrum data;

[0047] Analyzing the energy spectrum data to obtain the characteristic gamma photon peak of Na-24 at the top of the human head and the characteristic gamma photon peak at the lungs;

[0048] Based on the Na-24 characteristic gamma photon peak value obtained at the top of the head, the neutron exposure dose is calculated according to the neutron dose algorithm, and the type and activity of the internal contamination nuclide are determined based on the characteristic gamma photon peak value obtained at the lungs.

[0049] Specifically, this embodiment acquires and processes the gamma-ray signals from the top of the human head and lungs through a detection device and a digital multi-channel signal processing device, converts them into electric pulse signals, and then performs energy spectrum analysis, thereby realizing accurate calculation of neutron dose and identification and activity measurement of internal contamination nuclides. It can simultaneously provide the dual functions of neutron exposure dose and internal contamination diagnosis, and improve the efficiency and accuracy of detection.

[0050] The method of acquiring gamma ray signals at the top of the human head and the lungs by means of a detection device and converting the gamma ray into an electrical pulse signal by means of the detection device specifically includes:

[0051] The detection device is placed at a preset distance from the top of the human head and the lungs to measure the gamma ray signal, and the measurement time of each measurement position does not exceed the preset time (20 minutes);

[0052] The detection device comprises a scintillator and a photomultiplier tube. The scintillator converts the gamma rays into scintillation photon pulse signals, and the photomultiplier tube converts the scintillation photon pulse signals into electrical pulse signals.

[0053] Specifically, the process of generating the electrical pulse signal in this embodiment includes:

[0054] There are two physical processes from the interaction between gamma rays and scintillator to generate photoelectrons to the output of electrical pulses at the anode of the photomultiplier tube. The first process is completed in the scintillator, converting the gamma ray signal into a scintillation photon pulse signal; the second process is completed by the photomultiplier tube, converting the scintillation photon pulse into an electrical pulse.

[0055] The positions of the electrons emitted by the photocathode are different, resulting in different photoelectron transit paths, so the transit time is discrete and obeys the Gaussian distribution. Therefore, for an electron converted from a photon, after being multiplied in the photomultiplier tube, the final current formed on the anode is:

[0056]

[0057] Among them, i a is the final current formed on the anode, M is the electron gain on the photomultiplier tube, e is the electron, τ is the transit time, σ is the standard deviation of the transit time, which is used to describe the discrete degree of the photoelectron transit time, t is the current time point, and t' is the delay time from the photoelectron's departure to its arrival at the anode;

[0058] Assume that the total number of electrons emitted by the photocathode after a gamma radiation photon interacts with the scintillator is n k , then the number of electrons emitted by the photocathode from time t' to time dt' is dn = n k dt', where:

[0059]

[0060] Among them, n k is the total number of electrons emitted by the photocathode, N k It is the number of scintillation fluorescence that can interact with the photocathode after a γ radiation photon interacts with the crystal;

[0061] Then these photoelectrons are delayed by time t' compared to the incident time of the γ photon (t = 0), and the current pulse formed after reaching the anode of the photomultiplier tube is:

[0062]

[0063] Among them, i a is the final current formed on the anode, n k is the total number of electrons emitted by the photocathode;

[0064] That is, after the entire photoelectron pulse is multiplied at each energy level of the photomultiplier tube, the current pulse finally generated at the anode is:

[0065] i=i a *N k ;

[0066] Where i is the current generated by a γ radiation photon in the photomultiplier tube, i a is the final current formed on the anode.

[0067] This embodiment measures the gamma-ray signal by placing the detection device within a preset distance from the top of the human head and lungs, thereby ensuring the accuracy and effectiveness of the measurement. The detection device includes a scintillator and a photomultiplier tube, which can convert the gamma-ray into a scintillation photon pulse signal, and then convert it into an electrical pulse signal through the photomultiplier tube, thereby improving the efficiency and accuracy of the signal conversion. In addition, the measurement time does not exceed the preset time (20 minutes), which helps to improve the detection efficiency while ensuring the measurement accuracy and reduce the impact on the person being measured. This embodiment realizes efficient and accurate gamma-ray signal acquisition and conversion.

[0068] The electric pulse signal is processed by a digital multi-channel signal processing device, wherein the signal processing includes digital filtering and peak sampling to obtain energy spectrum data, specifically including:

[0069] Performing digital filtering processing on the electric pulse signal through an FPGA processing unit of a digital multi-channel signal processing device;

[0070] The FPGA processing unit includes a trigger module, a digital filter and a peak sampling module, the trigger module triggers signal acquisition, performs shaping filtering processing on the electrical pulse signal based on the digital filter, and extracts the pulse amplitude information of the electrical pulse signal through the peak sampling module;

[0071] The pulse amplitude information is peak sampled and accumulated to obtain a preset number (1024 channels) of energy spectrum data, wherein the energy spectrum data includes energy spectrum data obtained from the top of the head and energy spectrum data obtained from the lungs.

[0072] In a specific embodiment, the measurement range of the energy spectrum data is 80keV to 3.0MeV, the average channel width is 2.8keV, a 16-bit ADC analog-to-digital conversion chip is used for data acquisition, the conversion rate is greater than 10MHz, and the dead time is less than 1μs.

[0073] Specifically, this embodiment uses the FPGA processing unit of the digital multi-channel signal processing device to perform digital filtering on the electric pulse signal. The FPGA processing unit includes a trigger module, a digital filter and a peak sampling module. The trigger module is used to collect signals, the digital filter performs integer filtering on the electric pulse signal, and the peak sampling module extracts the pulse amplitude information, and performs peak sampling and accumulation on it, thereby obtaining a preset number (1024 channels) of energy spectrum data.

[0074] This embodiment can effectively remove noise and improve the signal-to-noise ratio of the signal through the integer filtering processing of the digital filter, thereby improving the accuracy of the energy spectrum data. The efficient processing capability of the FPGA processing unit enables rapid signal acquisition and processing, ensuring that high-quality energy spectrum data is obtained in a short time. Through the precise sampling and accumulation of the peak sampling module, high-resolution energy spectrum data can be obtained, which is crucial for accurately identifying the characteristic γ photon peak and identifying the internal contamination nuclides. The parallel processing capability of the FPGA allows real-time processing of large amounts of data, which is suitable for neutron dose and internal contamination diagnosis applications that require rapid response.

[0075] The energy spectrum data is analyzed to obtain the characteristic gamma photon peak of Na-24 at the top of the human head and the characteristic gamma photon peak at the lungs, specifically including:

[0076] The energy spectrum data obtained at the top of the head are analyzed. The energy value of the characteristic gamma photon of Na-24 is 1.36 MeV. In the energy spectrum measurement range of 80 keV to 3.0 MeV, the channel address of the peak of the characteristic gamma photon of Na-24 is located according to the average channel width of 2.8 keV, and the peak count is extracted;

[0077] The energy spectrum data obtained at the lungs are analyzed, and characteristic gamma photon peaks are located and extracted in all energy spectra based on characteristic gamma photon energy values ​​in a nuclide database, wherein the nuclide database includes characteristic gamma photon energy values ​​of fission and activation nuclides. The types of internal contamination nuclides are identified by comparing characteristic peaks in the energy spectrum data at the lungs with characteristic gamma photon energy values ​​in the nuclide database, and characteristic gamma photon peak counts corresponding to various internal contamination nuclides are extracted respectively.

[0078] Specifically, this embodiment can accurately identify the Na-24 characteristic γ photon peak through energy positioning and channel width setting, ensure the accuracy and reliability of measurement, use a 16-bit ADC analog-to-digital conversion chip for data acquisition, the conversion rate is greater than 10MHz, and the dead time is less than 1μs, which ensures the high resolution and high precision of the energy spectrum data, helps to accurately locate the characteristic peak, and extracts the Na-24 characteristic γ photon peak count to provide the necessary data basis for neutron dose calculation, ensuring the accuracy of neutron dose assessment. While identifying the Na-24 characteristic peak, it also provides a reference for the identification of internal contamination nuclides, and can effectively distinguish the characteristic peaks of different nuclides.

[0079] The neutron exposure dose is calculated according to the neutron dose algorithm based on the Na-24 characteristic gamma photon peak value obtained at the top of the head, and the type and activity of the internal contamination nuclide are determined according to the characteristic gamma photon peak value obtained at the lungs, specifically including:

[0080] The neutron dose algorithm is used to calculate the peak value of Na-24 characteristic γ photons obtained at the top of the head to obtain the neutron exposure dose value;

[0081] The characteristic gamma photon peak value obtained at the lungs is compared and analyzed with the data in the nuclide database, which contains characteristic data of fission nuclides and activated nuclides. The type of internal contamination nuclides is determined by comparative analysis, and the activity of internal contamination nuclides is calculated based on the characteristic gamma photon peak counts.

[0082] In a specific embodiment, the diagnostic lower limit of the activity of the internal contamination nuclide is no more than 1×10 5 Bq.

[0083] Specifically, this embodiment evaluates the neutron exposure dose by accurately calculating the peak value of the characteristic gamma photon of Na-24, and utilizes the characteristic gamma photon energy value of Na-24 to ensure the accuracy of dose calculation. This embodiment uses multiple empirical correction coefficients, which can be adjusted according to actual measurement conditions, thereby improving the reliability and applicability of the calculation results.

[0084] Through the optimized neutron dose algorithm, the dose calculation can be completed in a relatively short time, which is suitable for occasions where neutron radiation exposure needs to be quickly assessed. This method is not only suitable for laboratory environments, but can also be used to quickly diagnose neutron radiation doses on site, providing a wide range of application possibilities.

[0085] The calculation formula of the neutron dose algorithm is:

[0086]

[0087] Where D is the neutron exposure dose value, F d is the neutron absorption conversion efficiency, λ is the decay parameter of Na-24, T is the time when the personnel is exposed to neutrons, t is the time interval from the end of exposure to the measurement, N(t) is the peak count of Na-24 characteristic γ photons obtained at time t, k1, k2, k3 are the first empirical correction coefficient, the second empirical correction coefficient, and the third empirical correction coefficient, respectively.

[0088] Specifically, this embodiment introduces an empirical correction coefficient to adjust the neutron dose calculation according to different measurement conditions and environmental factors, thereby improving the accuracy of the calculation results and allowing high-precision dose assessment to be maintained in different operating environments. The use of the correction coefficient enables the method to adapt to different measurement scenarios and conditions, providing a wider range of application possibilities. Whether in a laboratory environment or in field measurements, a reliable dose assessment can be obtained by adjusting the correction coefficient.

[0089] In actual measurements, there may be systematic errors caused by device characteristics or environmental changes. These errors can be effectively compensated by empirical correction coefficients to ensure the stability and consistency of the measurement results. According to specific measurement requirements and conditions, users can adjust the correction coefficients to meet the requirements of specific applications, making this embodiment more flexible and efficient in actual applications.

[0090] See also Figure 2 The present invention also provides a neutron dose and internal contamination diagnostic instrument, the neutron dose and internal contamination diagnostic instrument comprises a detection device, a digital multi-channel signal processing device and a measuring cable, the digital multi-channel signal processing device is respectively connected to the detection device and the measuring cable signal;

[0091] The detection device is used to convert the gamma rays into electrical signals and perform preliminary amplification through a preamplifier;

[0092] The digital multi-channel signal processing device is used to receive and process the electrical signal from the detection device, perform signal analysis and energy spectrum processing through the digital spectrum analysis board, provide various required power supplies, and display the measurement results through the touch screen;

[0093] The measuring cable is used for signal transmission and power supply between the detection device and the digital multi-channel signal processing device through the signal output terminal, the preamplifier power supply terminal and the high-voltage input terminal.

[0094] The detection device includes a LaBr3 scintillator, a photomultiplier tube, a charge-sensitive preamplifier, a voltage divider circuit and a shell. The shell is an aluminum cylindrical structure with a diameter of 8.5 cm and a thickness of 1.5 mm, and is provided with a vibration-proof and heat-insulating protective shell on the outside. The LaBr3 scintillator is a cylindrical structure with a diameter of 75 mm and a height of 75 mm, and is air-freely connected to the photomultiplier tube with an effective photosensitive area diameter of 70 mm through optical silicone oil.

[0095] The digital multi-channel signal processing device adopts a double-layer circuit board structure, including an upper power board and a lower digital spectrum analysis board. The power board includes a DC-DC module and a high-voltage module for providing power; the digital spectrum analysis board includes an FPGA processing unit, a multi-channel analyzer and a memory for signal processing and energy spectrum analysis; and also includes a touch screen display module for displaying measurement results and performing human-computer interaction.

[0096] See also Figure 3 , is a power supply circuit diagram of the power supply board of this embodiment. The power supply board converts the 14.8V voltage output by the lithium battery into 1 +5V voltage and 1 ±12V voltage through a DC-DC module inside the digital multi-channel signal processing device, which is used to power the FPGA signal processing unit and the detection unit. For details, see Figure 3Among them, +5V is mainly used to power the digital circuit, while ±12V is mainly used to power the analog circuit. The two are connected through a single-point grounding method to reduce the impact of noise in the digital circuit on the analog circuit (since a high-speed ADC is used in the circuit, the digital circuit cannot be completely isolated from the analog circuit, and a single-point grounding method is used for connection).

[0097] See also Figure 4 , is the internal structure diagram of the FPGA signal processing unit of this embodiment. Since the sampling frequency of the high-speed A / D is fast, the amount of data output is huge, up to tens of megabytes per second. The general processor cannot process such huge data in real time, so an FPGA signal processing unit is used to process the output data of the A / D. The FPGA signal processing unit integrates the internal hardware gate circuit into a digital filter, performs integer filtering on the output signal of the A / D and extracts the pulse amplitude information of the electrical pulse signal, which is stored in the internal memory of the FPGA signal processing unit. Since the number of pulse amplitude information is only related to the number of pulses and has nothing to do with the frequency of pulse sampling, the amount of data output by the FPGA signal processing unit is effectively controlled. The interior of the FPGA signal processing unit mainly includes a trigger module, a digital filter, a peak sampling module, a memory, and an external bus interface.

[0098] The internal memory of the FPGA signal processing unit is used to store the energy spectrum data accumulated by peak sampling. The memory is usually implemented using FIFO or dual-port RAM. The external processor (ARM or DSP) can access the internal memory of the FPGA signal processing unit through the external bus interface of the FPGA signal processing unit to read the energy spectrum data and perform processing and analysis. In addition, the external processor can also write control parameters into the FPGA signal processing unit through the external bus interface to dynamically change the energy spectrum acquisition parameters (such as trapezoidal rise / fall time, decay time, etc.).

[0099] The acquisition of the energy spectrum is mainly completed by the spectrum analysis board in the signal processing unit. The energy spectrum acquisition accuracy is controlled, and the energy spectrum in this embodiment is 1024 channels. When the energy spectrum measurement range is 80keV~3.0MeV, the average channel width is about 2.8keV. When the pulse input amplitude range of the selected multi-channel analyzer ADC analog-to-digital conversion chip is 0V~4V, the amplitude corresponding to the average channel width is about 4mV, which has high requirements on the noise suppression of the analog signal processing circuit and the sampling performance of the analog-to-digital conversion circuit. In order to ensure the accuracy of energy spectrum acquisition, in this embodiment, on the one hand, when processing analog signals, including the amplification circuit, the filter shaping circuit, the baseline recovery circuit and the peak holding circuit design, try to avoid signal interference and improve the signal-to-noise ratio; on the other hand, when converting analog to digital, a 16-bit wide ADC analog-to-digital conversion chip is selected to reserve sufficient sampling accuracy to ensure the stability of the sampling value.

[0100] In order to ensure that the random radiation pulse overlap is as small as possible, the dead time of the multi-channel analyzer is required to be less than 1μs, that is, the peak holding discharge time before ADC conversion and the ADC analog-to-digital conversion time are less than 1μs. It is mainly solved from two aspects: one is the width control of the electric pulse signal. When designing the amplification circuit, filter shaping circuit, baseline recovery circuit and peak holding circuit, the time constant of each signal processing circuit is designed to be about 1us; the second is the ADC analog-to-digital conversion rate control. A high-speed ADC analog-to-digital conversion chip is selected, and the time to complete an analog-to-digital conversion is less than 1μs. Generally, an ADC chip with a conversion rate of more than 10MHz is selected.

[0101] Energy spectrum linearity control should try to improve the linearity of energy spectrum acquisition while ensuring the accuracy and speed of energy spectrum acquisition. This is mainly achieved in each analog signal processing circuit. The principle is to ensure that the pulse corresponding to the highest energy ray incident does not appear overload saturation in each signal processing channel, and in the pulse sampling and holding circuit, the leakage current of the capacitor is kept as small as possible, and the discharge speed of the discharge circuit is as fast as possible.

[0102] The measuring cable comprises a signal output terminal, a preamplifier power supply terminal and a high voltage input terminal. The signal output terminal is used to transmit the electrical signal output by the detection device, the preamplifier power supply terminal is used to power the preamplifier, and the high voltage input terminal is used to provide high voltage for the photomultiplier tube.

[0103] The basic performance requirements of a neutron dose and internal contamination diagnostic instrument of this embodiment include:

[0104] 1) The measurement object is gamma rays;

[0105] 2) Energy resolution is ≤3.0% (for 662keV energy of Cs-137);

[0106] 3) The energy range is 80keV~3MeV (energy spectrum range);

[0107] 4) Detection lower limit: neutron exposure dose diagnosis lower limit ≤ 0.1Gy, internal contamination diagnosis lower limit ≤ 1×10 5 Bq, diagnosis speed ≤ 20 min / person;

[0108] 5) Data throughput rate: not less than 100kcps;

[0109] 6) Integral nonlinearity: ≤±0.025%;

[0110] 7) Differential nonlinearity: ≤±1%;

[0111] 8) Continuous working time: not less than 72h;

[0112] 9) Multi-channel spectrometer: weight not exceeding 4kg (excluding batteries), size ≤ 250mm×200mm×100mm;

[0113] 10) Operation mode: touch screen operation;

[0114] 11) Collimator shielding performance: 5mmPb equivalent (for 662keV energy of Cs-137);

[0115] 12) Data management: automatic storage and export through USB interface;

[0116] 13) It has a basic nuclear database (including fission, activated nuclides, etc.), can intuitively characterize the neutron dose value and internal contamination degree of radiation-damaged patients, and provide software upgrade services.

[0117] The working principle of this embodiment is:

[0118] After a person is irradiated by neutrons, the Na in the body is activated to produce Na-24 and emit gamma rays. Based on the existing measurement technology, the neutron exposure dose of the injured person is calculated by measuring the Na-24 at the top of the head of the injured person. After a person is contaminated by radioactive nuclides, the radioactive nuclides are mainly concentrated in the lungs in a short period of time. By measuring the gamma rays in the lungs, the type and activity of the gamma nuclides contaminating the person can be quickly determined.

[0119] The above description is only a preferred embodiment of the present invention and is not intended to limit the present invention. Any modifications, equivalent substitutions, improvements, etc. made within the spirit and principles of the present invention should be included in the protection scope of the present invention.

Claims

1. A method for neutron dose and internal contamination diagnosis, characterized in that: The following steps are involved: Acquire gamma ray signals at the top of the head and lungs of the human body respectively through a detection device, and convert the gamma ray into an electrical pulse signal through the detection device; Using a digital multi-channel signal processing device to process the electric pulse signal, the signal processing includes digital filtering and peak sampling, to obtain energy spectrum data; Analyzing the energy spectrum data to obtain the characteristic gamma photon peak of Na-24 at the top of the human head and the characteristic gamma photon peak at the lungs; Based on the Na-24 characteristic gamma photon peak value obtained at the top of the head, the neutron exposure dose is calculated according to the neutron dose algorithm, and the type and activity of the internal contamination nuclide are determined based on the characteristic gamma photon peak value obtained at the lungs.

2. A neutron dose and internal contamination diagnosis method as claimed in claim 1, characterized in that: The method of acquiring gamma ray signals at the top of the human head and the lungs by means of a detection device and converting the gamma ray into an electrical pulse signal by means of the detection device specifically includes: The detection device is placed at a preset distance from the top of the human head and the lungs to measure the gamma ray signal, and the measurement time of each measurement position does not exceed the preset time; The detection device comprises a scintillator and a photomultiplier tube. The scintillator converts the gamma rays into scintillation photon pulse signals, and the photomultiplier tube converts the scintillation photon pulse signals into electrical pulse signals.

3. A neutron dose and internal contamination diagnosis method as claimed in claim 1, characterized in that: The electric pulse signal is processed by a digital multi-channel signal processing device, wherein the signal processing includes digital filtering and peak sampling to obtain energy spectrum data, specifically including: Performing digital filtering processing on the electric pulse signal through an FPGA processing unit of a digital multi-channel signal processing device; The FPGA processing unit includes a trigger module, a digital filter and a peak sampling module, the trigger module triggers signal acquisition, performs shaping filtering processing on the electrical pulse signal based on the digital filter, and extracts the pulse amplitude information of the electrical pulse signal through the peak sampling module; Peak sampling and accumulation are performed on the pulse amplitude information to obtain a preset amount of energy spectrum data, wherein the energy spectrum data includes energy spectrum data obtained from the top of the head and energy spectrum data obtained from the lungs.

4. A neutron dose and internal contamination diagnosis method as claimed in claim 1, characterized in that: The energy spectrum data is analyzed to obtain the characteristic gamma photon peak of Na-24 at the top of the human head and the characteristic gamma photon peak at the lungs, specifically including: The energy spectrum data obtained at the top of the head are analyzed. The energy value of the characteristic gamma photon of Na-24 is 1.36 MeV. In the energy spectrum measurement range of 80 keV to 3.0 MeV, the channel address of the peak of the characteristic gamma photon of Na-24 is located according to the average channel width of 2.8 keV, and the peak count is extracted; The energy spectrum data obtained at the lungs are analyzed, and characteristic gamma photon peaks are located and extracted in all energy spectra based on characteristic gamma photon energy values ​​in a nuclide database, wherein the nuclide database includes characteristic gamma photon energy values ​​of fission and activation nuclides. The types of internal contamination nuclides are identified by comparing characteristic peaks in the energy spectrum data at the lungs with characteristic gamma photon energy values ​​in the nuclide database, and characteristic gamma photon peak counts corresponding to various internal contamination nuclides are extracted respectively.

5. A neutron dose and internal contamination diagnosis method as claimed in claim 1, characterized in that: The neutron exposure dose is calculated according to the neutron dose algorithm based on the Na-24 characteristic gamma photon peak value obtained at the top of the head, and the type and activity of the internal contamination nuclide are determined according to the characteristic gamma photon peak value obtained at the lungs, specifically including: The neutron dose algorithm is used to calculate the peak value of Na-24 characteristic γ photons obtained at the top of the head to obtain the neutron exposure dose value; The characteristic gamma photon peak value obtained at the lungs is compared and analyzed with the data in the nuclide database, which contains characteristic data of fission nuclides and activated nuclides. The type of internal contamination nuclides is determined by comparative analysis, and the activity of internal contamination nuclides is calculated based on the characteristic gamma photon peak counts.

6. A neutron dose and internal contamination diagnosis method as claimed in claim 5, characterized in that: The calculation formula of the neutron dose algorithm is: Where D is the neutron exposure dose value, F d is the neutron absorption conversion efficiency, λ is the decay parameter of Na-24, T is the time when the personnel is exposed to neutrons, t is the time interval from the end of exposure to the measurement, N(t) is the peak count of Na-24 characteristic γ photons obtained at time t, k1, k2, k3 are the first empirical correction coefficient, the second empirical correction coefficient, and the third empirical correction coefficient, respectively.

7. A neutron dose and internal contamination diagnostic instrument, which performs a neutron dose and internal contamination diagnostic method as claimed in any one of claims 1 to 6, characterized in that: The neutron dose and internal contamination diagnostic instrument comprises a detection device, a digital multi-channel signal processing device and a measuring cable, wherein the digital multi-channel signal processing device is respectively connected to the detection device and the measuring cable signal; The detection device is used to convert the gamma rays into electrical signals and perform preliminary amplification through a preamplifier; The digital multi-channel signal processing device is used to receive and process the electrical signal from the detection device, perform signal analysis and energy spectrum processing through the digital spectrum analysis board, provide various required power supplies, and display the measurement results through the touch screen; The measuring cable is used for signal transmission and power supply between the detection device and the digital multi-channel signal processing device through the signal output terminal, the preamplifier power supply terminal and the high-voltage input terminal.

8. A neutron dose and internal contamination diagnostic instrument as claimed in claim 7, characterized in that: The detection device includes a LaBr3 scintillator, a photomultiplier tube, a charge-sensitive preamplifier, a voltage divider circuit and a shell. The shell is an aluminum cylindrical structure with a diameter of 8.5 cm and a thickness of 1.5 mm, and is provided with a vibration-proof and heat-insulating protective shell on the outside. The LaBr3 scintillator is a cylindrical structure with a diameter of 75 mm and a height of 75 mm, and is air-freely connected to the photomultiplier tube with an effective photosensitive area diameter of 70 mm through optical silicone oil.

9. A neutron dose and internal contamination diagnostic instrument as claimed in claim 7, characterized in that: The digital multi-channel signal processing device adopts a double-layer circuit board structure, including an upper power board and a lower digital spectrum analysis board. The power board includes a DC-DC module and a high-voltage module for providing power; the digital spectrum analysis board includes an FPGA processing unit, a multi-channel analyzer and a memory for signal processing and energy spectrum analysis; and also includes a touch screen display module for displaying measurement results and performing human-computer interaction.

10. A neutron dose and internal contamination diagnostic instrument as claimed in claim 7, characterized in that: The measuring cable comprises a signal output terminal, a preamplifier power supply terminal and a high voltage input terminal. The signal output terminal is used to transmit the electrical signal output by the detection device, the preamplifier power supply terminal is used to power the preamplifier, and the high voltage input terminal is used to provide high voltage for the photomultiplier tube.

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