A detection efficiency correction method for nuclear material inventory analysis and a detector

Abstract

The present application belongs to the technical field of nuclear material retention amount measurement and analysis, and particularly relates to a detection efficiency correction method for nuclear material retention amount analysis and a detector, wherein the method comprises the following steps: step S1, establishing a mathematical model suitable for a detection object; step S2, obtaining an environmental background and a gamma energy spectrum of the detection object; step S3, analyzing a full-energy peak area of characteristic gamma rays in the gamma energy spectrum; step S4, deducting the contribution of the environmental background to obtain a count rate of characteristic gamma rays of the nuclear material retention amount; step S5, determining intrinsic detection efficiency; step S6, determining geometric detection efficiency; step S7, performing decay absorption efficiency correction to obtain a decay correction parameter; and step S8, introducing the decay correction parameter to calculate the nuclear material retention amount.

Description

Technical Field

[0001] This invention belongs to the field of nuclear material retention measurement and analysis technology, specifically relating to a detection efficiency correction method and detector for nuclear material retention analysis. Background Technology

[0002] During the operation of nuclear facilities, nuclear materials are extensively deposited in pipelines and various process equipment. After the facility ceases operation, it is difficult to remove all the nuclear materials deposited in the process equipment, connecting pipelines, and adjacent work areas, resulting in residual material. Its formation mechanism is complex, related to the chemical properties of the nuclear materials, such as chemical composition and isotopic composition, as well as the nature, purpose, size, structural materials, and surface roughness of the process equipment. It is also related to the temperature, pressure, and flow rate parameters as the nuclear materials pass through the equipment or pipelines. Measurement and analysis of residual nuclear materials is essential for the nuclear material balance and control of the facility, reflecting the operational status and radiation safety status of the nuclear facility. The most commonly used and effective technique for measuring residual nuclear materials is gamma spectroscopy (GDS). GDS measures the residual material... 235 The characteristic 186keV gamma rays emitted by U, 239 The intensity of the characteristic gamma rays emitted by Pu at 129keV, 375keV, or 413keV is used to calculate the amount of nuclear material retained after detection efficiency correction. The measurement uncertainty is closely related to the object being detected. Summary of the Invention

[0003] The purpose of this invention is to provide a method that can customize a model based on the characteristics of the object to be measured and analyzed, obtain the characteristic gamma-ray detection efficiency matching the object after computation, and derive accurate quantitative analysis results of nuclear materials based on this efficiency. This approach facilitates the rationalization and precision of retention quantity measurement and analysis. For different objects, regardless of whether they are in production or operation, the retention quantity can be accurately quantified by customizing a model and combining it with the characteristics of their gamma-ray energy spectrum.

[0004] To achieve the above objectives, the technical solution adopted in this invention is a detection efficiency correction method for nuclear material retention analysis, comprising the following steps:

[0005] Step S1: Establish a mathematical model adapted to the detection object. The mathematical model includes a virtual first radiation source model for the first radiation source, a virtual detector model for the detector, and a virtual detection object model for the detection object.

[0006] Step S2: Use the detector to obtain the background gamma spectrum of the environment, and use the detector to obtain the gamma spectrum of the nuclear material retention in the target object.

[0007] Step S3: Analyze the full-energy peak area of ​​the characteristic γ-rays in the γ-ray spectrum of the background environment. The characteristic γ-rays refer to the γ-rays of specific energies emitted by nuclides in the nuclear material retention of the detection object.

[0008] Step S4: After deducting the environmental background, analyze the full-energy peak area of ​​the characteristic γ-rays in the γ-ray spectrum of the nuclear material retention amount to obtain the net count rate of the characteristic γ-rays in the nuclear material retention amount.

[0009] Step S5: Using the first radiation source model, the detector model, and the detection object model, determine the intrinsic detection efficiency of the detector and establish the intrinsic detection efficiency function of the characteristic γ-ray;

[0010] Step S6: Using the first radioactive source model, the detector model, and the detection object model, determine the geometric detection efficiency of the source and detector at their respective positions in space, and establish a geometric detection efficiency function related to the geometric structure of the characteristic γ-ray and the amount of nuclear material retained.

[0011] Step S7: Based on the physical parameters of the object being detected, introduce attenuation correction parameters for the characteristic γ-rays of the nuclear material retention amount; the physical parameters of the object being detected include material, shape, and size;

[0012] Step S8: The net count rate is corrected for attenuation absorption efficiency, and further corrected using the intrinsic detection efficiency and the geometric detection efficiency to calculate the amount of nuclear material retained.

[0013] Furthermore, in step S1, the mathematical model is established using a Monte Carlo simulation tool, and the object of detection is an actual existing pipeline structure.

[0014] further,

[0015] In step S5, through several experiments, the distance and angle between the detector model, the first radiation source model, and the detection target model are adjusted in each experiment. Based on the experimental results, the intrinsic detection efficiency function of the characteristic γ-rays of the detector model under the source-detector distance condition of the first radiation source model is obtained, and the formula is derived:

[0016]

[0017] In the formula,

[0018] ε intrinsic The intrinsic detection efficiency is mentioned above;

[0019] E is the energy of the characteristic γ-ray, expressed in keV;

[0020] x is the source-detection distance of the first radioactive source, in cm;

[0021] a i These are the mathematical fitting parameters.

[0022] Furthermore, in step S6, using the detector model and the first radioactive source model, and based on the geometric structure of the nuclear material retention, a geometric detection efficiency function for the characteristic gamma rays is established, yielding the formula:

[0023]

[0024] In the formula,

[0025] ε geometry For geometric detection efficiency;

[0026] n is the quantity obtained by discretizing the geometric distribution of nuclear material retention;

[0027] x represents the detection distance of the first radioactive source, in cm;

[0028] b i These are the mathematical fitting coefficients.

[0029] further,

[0030] In step S7, the attenuation correction parameter is obtained by actual measurement of the probe object or by simulation measurement using a simulation model of the probe object; the attenuation correction parameter is denoted as K, as shown in the formula:

[0031]

[0032] In the formula,

[0033] ε total The detection efficiency of representative gamma rays after attenuation and absorption.

[0034] ε geometry This represents the geometric detection efficiency obtained in step S6.

[0035] further,

[0036] In step S8, the detection efficiency of the nuclear material retention under the actual detection range of the radioactive source is obtained by multiplying the attenuation correction parameters by the formula, as shown in the formula:

[0037]

[0038] This invention also discloses a detector for a detection efficiency correction method for nuclear material retention analysis as described above. The detector includes a detector crystal disposed within a lead collimator. A cladding shell surrounds the detector crystal, and a reflective layer is disposed between the cladding shell and the detector crystal. The top of the detector crystal is near the opening of the lead collimator, and an optical glass is disposed at the bottom of the detector crystal.

[0039] Furthermore, it also includes a stainless steel outer shell disposed around the lead collimator, and a polytetrafluoroethylene material is disposed between the lead collimator and the outer shell.

[0040] Furthermore, the detector crystal is a lanthanum bromide crystal, the cladding material is aluminum, and the reflective layer material is MgO.

[0041] The beneficial effects of this invention are as follows:

[0042] 1. The model customization of the retention quantity object has been realized.

[0043] 2. In the process of eigendetection efficiency correction of the detector, the intrinsic detection efficiency function of the detector for the characteristic γ-rays of nuclear materials was established under different source distances.

[0044] 3. During the geometric detection efficiency correction process, a geometric detection efficiency function relationship was established for γ-rays with different energy characteristics and nuclear material retention amounts.

[0045] 4. In the process of attenuation absorption efficiency correction, attenuation correction parameters for characteristic γ-rays of nuclear material retention in different substances were introduced.

[0046] 5. The detection efficiency correction method provided by this invention is applicable to various objects of retention analysis.

[0047] 6. This method can also be applied to the screening and quantitative analysis of other gamma radionuclides besides nuclear materials.

[0048] 7. This invention features the speed and convenience of NDA technology, and can simultaneously analyze the content of multiple elements and their isotopes, providing data for nuclear material closed-loop balance calculations, safe operation of nuclear facilities, production efficiency evaluation, and decommissioning management. Attached Figure Description

[0049] Figure 1 This is a schematic diagram of the detector described in a specific embodiment of the present invention;

[0050] Figure 2 This is a schematic diagram showing the positional relationship between the detector and the radiation source (first radiation source or second radiation source) in a specific embodiment of the present invention;

[0051] Figure 3This is a schematic diagram of the detector, radiation source, and detection object (pipe) described in a specific embodiment of the present invention during a detection test (the detector, radiation source, and detection object can be virtual models or physical objects);

[0052] Figure 4 This is a schematic diagram illustrating the principle of the nuclear material retention quantity detection efficiency correction method (γ-ray energy spectroscopy method for measuring retention quantity) provided by the present invention.

[0053] In the diagram: 1-Stainless steel casing, 2-Lead collimator, 3-PTFE material, 4-Detector crystal, 5-Shell, 6-Reflective layer, 7-Optical glass, 8-Radiation source, 9-Pipe (detection target). Detailed Implementation

[0054] The present invention will be further described below with reference to the accompanying drawings and embodiments.

[0055] The principle of nuclear material retention detection is based on the fact that the nuclear material retention is radioactive. According to the general principle of radioactive decay, the radioactivity A of the retention is related to its mass M, as shown in the following equation (1):

[0056]

[0057] In the formula, λ is the decay constant of the radionuclide, M is the mass of the radioactive material, M0 is the molar mass of the radionuclide, and N... A is Avogadro's constant.

[0058] After obtaining the γ-ray spectrum of the target object by measurement, the full-energy peak area of ​​the characteristic γ-rays of the nuclide of interest is obtained by energy spectrum analysis. The count rate n0 is obtained by dividing the energy spectrum acquisition time. n0 is related to factors such as the activity of the radioactive source, detector efficiency, geometric conditions, γ-ray branching ratio and absorption of shielding material, as shown in the following equation (2):

[0059] n0=ε1·ε2·ε3·Br·A (2)

[0060] In the formula, ε1 is the intrinsic detection efficiency of the detector, ε2 is the geometric distribution efficiency, ε3 is the shielding absorption efficiency, and Br is the branching ratio of the characteristic γ-rays.

[0061] This invention provides a method for correcting the detection efficiency of nuclear material retention analysis, comprising the following steps:

[0062] Step S1: Establish a mathematical model adapted to the structure of the target object. The mathematical model includes a virtual first radioactive source model for the first radioactive source, a virtual detector model for the detector, and a virtual target object model for the target object; (see detector for details). Figure 1The positional relationship between the first radiation source and the detector is shown in [reference needed]. Figure 2 )

[0063] Step S2: Use a detector to obtain the amount of nuclear material remaining in the target and the gamma spectrum of the background environment, and analyze the full-energy peak area of ​​characteristic gamma rays in the background gamma spectrum. Characteristic gamma rays refer to gamma rays of specific energy emitted by nuclides in the amount of nuclear material remaining.

[0064] Step S3: Analyze the full-energy peak area of ​​characteristic gamma rays in the gamma spectrum of the nuclear material retention. Characteristic gamma rays refer to gamma rays of specific energies emitted by nuclides in the nuclear material retention, including... 235 The characteristic 186keV gamma rays emitted by U, 239 Pu emits characteristic gamma rays of 129 keV, 375 keV, or 413 keV, etc.

[0065] Step S4: Subtract the full-energy peak area of ​​the characteristic γ-rays of the environmental background to obtain the net count rate of characteristic γ-rays in the nuclear material retention amount;

[0066] Step S5: Using the first radioactive source model, detector model, and detection object model, determine the intrinsic detection efficiency of the detector (that is, determine the intrinsic detection efficiency of the characteristic γ-ray count of the nuclear material retention amount). Under the basic conditions such as the source-detection distance between the detector and the first radioactive source, establish the intrinsic detection efficiency function of the characteristic γ-ray, that is, establish the functional relationship between the source-detection distance between the detector and the first radioactive source and the energy of the characteristic γ-ray.

[0067] Step S6: Using the first radioactive source model, detector model, and detection object model, perform geometric detection efficiency correction, determine the geometric detection efficiency of the source and detector (the source and detector refer to the first radioactive source and detector) at their respective spatial positions (here, geometry refers to the different spatial positions of the detection object and the detector), and establish a geometric detection efficiency function related to the geometric structure of the characteristic gamma rays and the nuclear material retention amount, that is, establish the correspondence between the geometric structure of the nuclear material retention amount and the energy of the characteristic gamma rays.

[0068] Step S7: Based on the physical parameters of the object being detected, introduce attenuation correction parameters for the characteristic gamma rays of nuclear material retention; the physical parameters of the object being detected include material, shape, and size;

[0069] Step S8: Correct the net count rate for attenuation absorption efficiency, further correct it using intrinsic detection efficiency and geometric detection efficiency, and calculate the amount of nuclear material retained.

[0070] In step S1, a mathematical model is established using a Monte Carlo simulation tool. In this application, the object of detection is an actual pipe structure. The physical structure and material of the object of detection are modeled using the Monte Carlo simulation tool, as well as a detector model and a first radiation source model are established, so that the corresponding detection efficiency can be simulated using the Monte Carlo simulation tool.

[0071] In step S5, through several experiments, the distance and angle between the detector model, the first radioactive source model, and the detection object model are adjusted in each experiment. Based on the experimental results, the functional relationship between the source-detection distance of the detector model and the first radioactive source model is obtained, and then the functional relationship between the actual source-detection distance of the detector and the first radioactive source and the energy of the characteristic γ-ray is obtained, and formula (3) is derived:

[0072]

[0073] In the formula,

[0074] ε intrinsic The intrinsic detection efficiency is ε1 in formula (2);

[0075] E is the energy of the characteristic γ-ray, expressed in keV;

[0076] x is the source-detection distance of the first radioactive source, in cm;

[0077] a i These are the mathematical fitting parameters;

[0078] In practical applications, if the source-detection distance of the first radioactive source is much greater than its geometric dimensions, the detection efficiency can be obtained directly using this function.

[0079] (The purpose of the previous experiments was to fine-tune the virtual model. Because the detector model was modeled according to the design dimensions of the various objects inside the actual detector, there would be some error compared to the actual detector. To reduce this error, experiments were conducted, and the detector model was fine-tuned based on the experimental data. The goal was to obtain a simulation model that more closely matches the actual detector; in subsequent steps, the detector model will remain unchanged.)

[0080] In step S6, based on the detector model and the first radioactive source model after the intrinsic detection efficiency correction in step S5, formula (4) is derived according to the correspondence between the geometric structure of the nuclear material retention and the energy of the characteristic γ-rays:

[0081]

[0082] In the formula,

[0083] ε geometryThe geometric detection efficiency is the ε1·ε2 part in formula (2);

[0084] n is the quantity obtained by discretizing the geometric distribution of nuclear material retention;

[0085] x represents the source-detection distance of the first radioactive source, in cm;

[0086] b i These are the mathematical fitting coefficients.

[0087] In practical work, the discretization method for the geometric distribution of nuclear material retention can vary depending on the actual physical model of the probe object. For example, the geometric features of the probe object (length, volume, surface area, surface area-to-volume ratio, number, etc.) can be meshed to perform geometric efficiency fitting.

[0088] The geometric detection efficiency correction is obtained through a virtual model. A small number of experiments are also conducted to verify the correctness of the virtual model and the accuracy of the simulation results. During the experimental verification, a second radioactive source, which is different from the first radioactive source, was used. Previously, some standard point sources were used to ensure the accuracy of the established detector model. The second radioactive source is a slightly larger cylindrical uranium source that is geometrically related to the retention distribution.

[0089] In step S7, the attenuation correction parameter is obtained by actual measurement of the probe object or by simulation measurement using a simulation model of the probe object (if actual measurement is used, a second radiation source is used; the positional relationship between the detector, radiation source, and probe object is shown in [reference needed]). Figure 3 Different substances have different attenuation and absorption capabilities for characteristic gamma rays of different energies. Utilizing the physical mechanism of attenuation absorption, the attenuation correction parameter under a certain source distance condition is denoted as K, as shown in Formula 5:

[0090]

[0091] In the formula,

[0092] ε total The detection efficiency of the characteristic γ-ray after attenuation and absorption is represented by ε1·ε2·ε3 in formula (2);

[0093] ε geometry This represents the geometric detection efficiency obtained in step S6.

[0094] In step S8, the detection efficiency of nuclear material retention under the actual detection range of the radioactive source is obtained by multiplying the attenuation correction parameters by Equation 4 (e.g., 235 The detection efficiency of the characteristic gamma rays emitted by U with an energy of 185 keV after passing through a certain absorbing material is shown in Formula 6:

[0095]

[0096] Formula 6 is the final correction result for the amount of nuclear material retained in the probe (pipeline).

[0097] Both experimental and simulation methods can be performed in this step. Experimental methods require the preparation of corresponding source samples, while simulation methods require understanding the specific parameters of the target object, such as geometric dimensions, design specifications, and materials.

[0098] The detection efficiency correction method provided by this invention requires: 1) ensuring the stable operation of the detector; 2) controlling the background radioactivity intensity; 3) accurately establishing the geometric model; and 4) accurately selecting the attenuation absorption material.

[0099] This invention also discloses a detector for a detection efficiency correction method used in the nuclear material retention analysis described above (see [link to detector]). Figure 1 The detector includes a detector crystal 4 disposed inside a lead collimator 2. The detector crystal 4 is surrounded by a shell 5. A reflective layer 6 is disposed between the shell 5 and the detector crystal 4. The top of the detector crystal 4 is close to the opening of the lead collimator 2, and the bottom of the detector crystal 4 is provided with optical glass 7.

[0100] It also includes a stainless steel outer shell 1 surrounding the lead collimator 2, and a polytetrafluoroethylene material 3 between the lead collimator 2 and the outer shell 5.

[0101] The detector crystal 4 is a lanthanum bromide crystal (LaBr3 detector, 2 inches in diameter); the casing 5 is made of aluminum, and the reflective layer 6 is made of MgO.

[0102] The detector also includes an electron spectrometer (Inspector2000) and energy spectrum acquisition software (Genie2000). The area of ​​the characteristic γ-ray peak is obtained by manually truncating and subtracting the background count. This detector is used for the entire measurement work after step S5, and the detector model is also consistent with this detector.

[0103] Example

[0104] In step S5, during the fitting of the detector's intrinsic detection efficiency...

[0105] The first radioactive source used in the experiment included 133 Ba、 137 Cs、 152 Eu、 60Co sources with activities of 12394 Bq, 7995 Bq, 791880 Bq, and 5411 Bq (calibrated on November 29, 2021) were used. The corresponding efficiencies were measured at different source distances (15 cm, 20 cm, 25 cm, and 30 cm). After curve fitting, the relationship between efficiency and distance and energy was obtained, as shown in formula (7). Parameters are shown in Table 1, and the correlation coefficient R0 is given. 2 It is 0.9904.

[0106] After establishing the working model of the detector, it was verified under the same conditions as the experiment, and the parameter correlation coefficient R was used. 2 It is 0.9981.

[0107]

[0108] In the formula ε intrinsic Let E be the detection efficiency, E be the energy (keV), and x be the distance (cm) from the point source to the front edge of the detector. In the formula, a... i The parameters are shown in Table 1 below.

[0109] Table 1. Parameters for determining the intrinsic detection efficiency of the detector

[0110]

[0111] In step S6, during the geometric detection efficiency fitting...

[0112] The target object was divided into grids, and detection was performed at distances of 15cm, 20cm, and 30cm from the detector. The target object model used in the simulation was modified to obtain detection efficiency under more conditions. The detection efficiency was fitted, resulting in Equation 8, with a correlation coefficient of 0.9985. Where ε... geometry For detection efficiency, n is the number of grids where radiation locations exist. In this experiment, after discretizing the number of grids, m represents the number of volume sources used, and x is the distance (cm) from the source to the front edge of the detector. In the formula, b... i The parameters are shown in Table 3 below. Partial data obtained from fitting calculations and actual measurements are shown in Table 4 below.

[0113]

[0114] Table 3 Parameters of Geometric Detection Efficiency Function

[0115]

[0116] Table 4 shows the simulation and experimental results and relative errors of the volume source detection efficiency used for correction simulation.

[0117]

[0118] Step S7, attenuation correction parameter fitting

[0119] Different materials have different mass absorption factors for rays of different energies. For gamma rays of a specific energy, the higher the density of the absorbing material, the more severe the absorption. For a certain material, the higher the energy of the gamma rays, the stronger the penetrating power. Using experiments or simulations, the detection efficiency at a certain distance is obtained without a conduit (the conduit being the object of detection) and with a conduit inserted. The ratio is denoted as K.

[0120]

[0121] In the formula, ε total ε represents the detection efficiency after being placed in the pipe. geometry This represents the detection efficiency without the addition of pipe material.

[0122] The detection efficiency of a cylindrical source (i.e., the second radioactive source) inside a pipe at 185 keV is:

[0123]

[0124] Experimental verification

[0125] Using other numbers of cylindrical uranium sources, detection was performed at different distances from the detector to obtain experimental data, which was then compared with simulated data to verify the correctness of the geometric detection efficiency fitting. The results are shown in Table 5.

[0126] Table 5 Simulation and experimental results of detection efficiency and relative error of U-sources with different numbers of cylinders.

[0127]

[0128] Different numbers of cylindrical uranium sources were placed inside a metal pipe container (the object of detection), and detection was conducted at different distances from the detector to obtain gamma-ray energy spectra. After calibration, the average activity of a single cylindrical uranium source was found to be 7.74 × 10⁻⁶. 4 Bq, and its calibration value of 8.55 × 10 4 Compared to Bq, the deviation is 9.45%.

[0129] Table 6 Simulation and experimental results and relative errors of detection efficiency when different numbers of cylindrical U-sources are placed inside the tube.

[0130]

[0131] The device described in this invention is not limited to the embodiments described in the specific implementation. Other implementation methods derived by those skilled in the art based on the technical solution of this invention also fall within the scope of technical innovation of this invention.

Claims

1. A method for detecting efficiency correction of nuclear material retention analysis, comprising the following steps: Step S1, establishing a mathematical model suitable for a detection object, the mathematical model comprising a virtual first radioactive source model for a first radioactive source, a virtual detector model for a detector, and a virtual detection object model for the detection object; Step S2, acquiring an environmental background gamma spectrum by using the detector, and acquiring a nuclear material retention gamma spectrum in the detection object by using the detector; Step S3, analyzing a full-energy peak area of a characteristic gamma ray in the environmental background gamma spectrum, the characteristic gamma ray being a gamma ray of a specific energy emitted by a nuclide in the nuclear material retention in the detection object; Step S4, after deducting the environmental background, analyzing a full-energy peak area of a characteristic gamma ray in the nuclear material retention gamma spectrum, to obtain a net count rate of the characteristic gamma ray of the nuclear material retention; Step S5, determining an intrinsic detection efficiency of the detector by using the first radioactive source model, the detector model, and the detection object model, to establish an intrinsic detection efficiency function of the characteristic gamma ray; Step S6, determining a geometric detection efficiency of a source-detector spatial position by using the first radioactive source model, the detector model, and the detection object model, to establish a geometric detection efficiency function of the characteristic gamma ray related to a geometric structure of the nuclear material retention; Step S7, introducing an attenuation correction parameter of the characteristic gamma ray of the nuclear material retention according to physical parameters of the detection object, the physical parameters of the detection object including material, shape, and size; Step S8, performing attenuation and absorption efficiency correction on the net count rate, and further correcting by using the intrinsic detection efficiency and the geometric detection efficiency, to calculate the nuclear material retention.

2. A method of correcting the detection efficiency of a nuclear material inventory analysis according to claim 1, characterized in that the step of In the step S1, the mathematical model is established by using a Monte Carlo simulation tool, and the detection object is an actually existing pipeline structure.

3. The method for detecting efficiency correction of nuclear material retention analysis according to claim 1, characterized in that: in the step S5, the detector model is obtained according to experimental results by adjusting a distance and an angle between the detector model, the first radioactive source model, and the detection object model in each test, and a formula (3) is derived: wherein, ε intrinsic is the intrinsic detection efficiency; E is an energy of the characteristic gamma ray, in keV; x is a source-detector distance of the first radioactive source, in cm. a i is a mathematical fitting parameter.

4. The method for detecting efficiency correction of nuclear material retention analysis according to claim 1, characterized in that: in the step S6, the geometric detection efficiency function of the characteristic gamma ray is established according to a geometric structure of the nuclear material retention by using the detector model and the first radioactive source model, to obtain a formula (4): wherein, ε geometry Geometric detection efficiency; n is a number obtained by discretizing the geometric distribution of the nuclear material retention; x is a source-detector distance of the first radioactive source, in cm. b i are the mathematical fitting coefficients.

5. The method of claim 1, wherein the step S7 is characterized in that: the decay correction parameter is obtained by actual measurement of the detection object or simulation measurement of a simulation model of the detection object; the decay correction parameter is denoted as K, and the formula (5) is: K = 1 / (1 - e -λt) (5) wherein, λ is a decay constant of the detection object, and t is a time interval of the detection object.

6. The method of claim 4, wherein the step S8 is characterized in that: the detection efficiency of the nuclear material inventory under the detection distance condition of the radioactive source is obtained by multiplying the decay correction parameter by the formula (4), and the formula (6) is: E = K * E0 (6) wherein, E0 is the detection efficiency of the nuclear material inventory under the detection distance condition of the radioactive source. The detector comprises a detector crystal (4) arranged in a lead collimator (2), an outer periphery of the detector crystal (4) is provided with a cladding (5), a reflection layer (6) is further arranged between the cladding (5) and the detector crystal (4), a top end of the detector crystal (4) is close to an opening of the lead collimator (2), and an optical glass (7) is arranged at a bottom end of the detector crystal (4). ε total The detection efficiency of the characteristic γ-ray after the attenuation absorption ε geometry represent the geometrical detection efficiency obtained in said step S6. Further comprising a stainless steel shell (1) arranged at an outer periphery of the lead collimator (2), and a polytetrafluoroethylene material (3) is further arranged between the lead collimator (2) and the cladding (5). In The detector crystal (4) is a lanthanum bromide crystal, the material of the cladding (5) is aluminum, and the material of the reflection layer (6) is MgO.

7. A probe for use in a method as claimed in claim 1, characterised in that: ​ 8. The probe of claim 7, wherein: ​ 9. The probe of claim 8, wherein: ​

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