Method and apparatus for monitoring neutron ambient dose equivalent in a wide energy region

Abstract

This invention relates to the field of neutron dose monitoring, specifically to a method and apparatus for monitoring the dose equivalent around neutrons over a wide energy range. The apparatus includes an outer moderator, a thermal neutron absorbing layer, a high-energy neutron compensation layer, and an inner moderator. The high-energy neutron compensation layer is composed of a heavy metal material capable of undergoing a significant neutron multiplication reaction with ultra-high-energy neutrons. The thermal neutron detector is a radiation-resistant solid-state semiconductor detector, comprising a silicon carbide semiconductor substrate and an enrichment layer closely adhering to its surface. 6 The device is equipped with a lithium fluoride thermal neutron converter; an initial thermal neutron detector group is deployed within the device to collect thermal neutron fluence information under different slowing degrees; based on the initial thermal neutron detector group, position optimization and weight coefficient output are performed, and multiple groups of thermal neutron detectors in the first depth order are selected to make the total dose response flat in a wide energy range for dose equivalent monitoring.

Description

Technical Field

[0001] This invention relates to the field of thermal neutron energy region monitoring technology, and in particular to a method and device for monitoring the dose equivalent around neutrons in a wide energy region. Background Technology

[0002] Neutron ambient dose equivalent is a core physical quantity for assessing the dose received by personnel in mixed radiation fields. Traditional moderator dosimeters achieve measurement through a moderator and a thermal neutron detector, and their design quality depends on how well their dose response function conforms to the standard neutron fluence-dose equivalent conversion factor curve recommended by the International Commission on Radiation Protection (ICRP). However, current technology faces insurmountable physical bottlenecks when dealing with ultra-wide energy spectra up to the GeV level.

[0003] First, GeV-class neutrons possess extremely high penetrability and are difficult to slow down effectively within moderated volumes of conventional sizes, resulting in very low detection efficiency. Second, the interaction of high-energy neutrons with matter produces complex secondary particle showers, causing significant interference to traditional detector signals.

[0004] Currently, some advanced extended-range neutron dosimeters, while attempting to improve high-energy response by adding heavy metal layers (such as tungsten), typically have energy limits limited to a few hundred MeV, and their response functions deviate sharply from the standard curve at higher energies. The reason for this is that these designs have not fundamentally solved the problem of effectively picking up weak signals in the GeV energy range and achieving a systematic balance between response across the entire energy range.

[0005] Furthermore, existing multi-detector schemes rely heavily on empirical layout and lack a scientific method to systematically and globally optimize the number of detectors, their spatial positions, and signal weights. This results in a long-standing contradiction between hardware resource redundancy and poor response flatness, failing to meet the growing urgent demand from cutting-edge fields such as high-energy physics experiments for compact, efficient, and accurate ultra-wide energy range neutron monitoring equipment.

[0006] To address this, a method and device for monitoring the dose equivalent around neutrons in a wide energy range are proposed. Summary of the Invention

[0007] To address the aforementioned technical problems, this invention provides a method and apparatus for monitoring the dose equivalent around neutrons in a wide energy range.

[0008] The method for monitoring neutron ambient dose equivalent in a wide energy range, which outputs neutron ambient dose equivalent through a combination of multiple detectors, includes:

[0009] S10: Calculate the flux response values ​​of the initial thermal neutron detector array within the moderated body and the flux response function formed by the incident neutron energy, where the flux response function of the i-th detector is R. i(E), the average fluence response function of the i-th group of detectors in the spherical structure is expressed as follows:

[0010] ;

[0011] in, Let represent the flux response function of the j-th detector in the i-th group of detectors;

[0012] S20: Calculate the relative dose response function formed by comparing the fluence response function of the initial thermal neutron detector group with the ICRP74 neutron fluence-ambient dose equivalent conversion function. The expression for the relative dose response function of the i-th (group) detector of the dose equivalent monitoring device is as follows:

[0013] ;

[0014] in, Represents the relative dose response function. This represents the ICRP74 neutron dose-surround dose equivalent conversion function;

[0015] S30: Select an optimized detector combination from the initial thermal neutron detector group of the dose equivalent monitoring device, and multiply the relative dose response function matrix D of the detector combination with the corresponding linear combination coefficient vector w to obtain the total relative dose response function R(E). The formula for calculating the linear combination coefficient vector w is as follows:

[0016] ;

[0017] in, D yes k OK l The coefficient matrix of the column, l Each column in the column represents the flux response function of the selected detector (group of detectors). k Each value represents the flux response function within a specified energy range. k One response value; w It is a length of l The column vectors represent l The weighting coefficients corresponding to each response function; D ideal It is a length of k The column vector, with a value of 1, represents the flat relative dose response within the specified energy range;

[0018] S40: The neutron ambient dose equivalent is obtained by combining and outputting the counts of each detector in the detector combination selected from the dose equivalent monitoring device according to the solved linear combination coefficients. The expression for the neutron ambient dose equivalent output by the multi-detector combination is as follows:

[0019] ;

[0020] in, w i For the first i Weighting coefficients for individual (group) detectors; R i For the first i The average flux response of each (group of) detectors; C i For the first i The (average) count of each (group of) detectors; This represents the neutron fluence per unit energy.

[0021] The optimized thermal neutron detector array, obtained from the initial array, involves the following steps:

[0022] S11: Traverse all combinations of 2 to n detectors from the initial thermal neutron detector group of the dose equivalent monitoring device, and calculate the total relative dose response function for each detector combination;

[0023] S12: Calculate the relative standard deviation of the total relative dose response function for all detector combinations. Take the combination with the smallest relative standard deviation of the total relative dose response function among all combinations of n detectors (groups) as the optimal combination of n detectors (groups). Obtain the curve of the relative standard deviation of the optimal combination as the number of detectors increases.

[0024] S13: Determine the minimum number of detectors required to approach the minimum relative standard deviation and the optimal combination method from the curve of the relative standard deviation of the optimal combination method as the number of detectors increases, as the final optimized thermal neutron detector combination.

[0025] The optimization results of multi-detector combination are used to guide the structural optimization of the device, including:

[0026] S21: The relative standard deviation of the optimized combination of multiple detectors and the total relative dose response function of the initial structure of the device is calculated;

[0027] S22: Based on the energy range of the flatness difference of the total relative dose response function of the multi-detector in the initial structure optimization, optimize the structural parameters for the corresponding energy range response compensation, and calculate the multi-detector combination optimization result under the optimized structural parameters;

[0028] S23: Evaluate the optimization results of the multi-detector combination of the optimized structure, and further optimize the structural parameters until the device structure with the flattest dose response is obtained.

[0029] A wide-energy-range neutron ambient dose equivalent monitoring device is used to implement the aforementioned wide-energy-range neutron ambient dose equivalent monitoring method. The dose equivalent monitoring device includes an outer moderator, a thermal neutron absorbing layer, a high-energy neutron compensation layer, and an inner moderator.

[0030] The high-energy neutron compensation layer is composed of a heavy metal material that can undergo a significant neutron multiplication reaction with high-energy neutrons;

[0031] The dose equivalent monitoring device contains an initial thermal neutron detector group deployed in the moderator in a first depth order for collecting thermal neutron fluence information under different levels of moderation; multiple groups of thermal neutron detectors optimized from the initial thermal neutron detector group are used for neutron ambient dose equivalent monitoring;

[0032] Thermal neutron detectors are radiation-resistant solid-state semiconductor detectors, comprising a silicon carbide semiconductor substrate and an enrichment layer attached to its surface. 6 Lithium fluoride thermal neutron converter.

[0033] The high-energy neutron compensation layer is made of lead and tungsten, and the thermal neutron absorption layer is made of boron carbide.

[0034] The dose equivalent monitoring device comprises a multi-layered concentric spherical structure and a cylindrical structure.

[0035] The initial thermal neutron detector group is configured as follows: when the dose equivalent monitoring device is a multi-layer concentric spherical structure, a set of thermal neutron detectors is arranged 1 cm away from the front surface in the radial direction of the outer layer moderator, in front, behind, left, right, top and bottom, and one detector is arranged in the center of the inner layer moderator and multiple sets of thermal neutron detectors are arranged alternately with equal radial depth intervals of 5 mm.

[0036] When the dose equivalent monitoring device is a cylindrical structure, multiple thermal neutron detectors are arranged alternately at equal intervals of 5 mm along the axial depth in the inner and outer layers of the moderator.

[0037] The embodiments of the present invention have the following technical effects:

[0038] 1. Structural design of multi-detector neutron ambient dose equivalent meter: The combination of thermal neutron absorption layer and high-energy neutron compensation layer divides a whole cylindrical moderator into two parts, forming a low-energy neutron dose response region (outer moderator) and a high-energy neutron dose response region (inner moderator), respectively. The dose response functions of the thermal neutron detectors in the two regions are similar in shape to the low-energy and high-energy parts of the ICRP74 dose conversion function, respectively.

[0039] 2. Method for outputting neutron ambient dose equivalent using multiple detector combinations: Arrange a sufficient number of thermal neutron detectors in the moderator to fully acquire detailed response functions at different depths of the moderator. Iterate through the combination methods under different numbers of response functions, solve for the total dose response function under the optimal combination method (the combination with the smallest relative standard deviation of the total relative dose response function) for different numbers of detectors, and determine the optimal number of detectors and the corresponding combination method based on the relationship between the relative standard deviation of the total relative dose response function and the number of detectors, and combine them to output the ideal total dose response function. Attached Figure Description

[0040] To more clearly illustrate the specific embodiments of the present invention or the technical solutions in the prior art, the drawings used in the description of the specific embodiments or the prior art will be briefly introduced below. Obviously, the drawings described below are some embodiments of the present invention. For those skilled in the art, other drawings can be obtained from these drawings without creative effort.

[0041] Figure 1 This is a schematic diagram of the multi-layer concentric spherical structure provided in an embodiment of the present invention;

[0042] Figure 2 This is a schematic flowchart of a wide-energy-range neutron dose equivalent monitoring method provided in an embodiment of the present invention;

[0043] Figure 3 This is a schematic diagram of the combination matrix when selecting 3 groups of thermal neutron detectors according to an embodiment of the present invention;

[0044] Figure 4 This is a schematic diagram of the total dose response function obtained by weighting according to the optimal detector combination provided in an embodiment of the present invention;

[0045] Figure 5 This is a graph of the total relative dose response function obtained by weighting according to the optimal detector combination, provided in an embodiment of the present invention.

[0046] Figure 6 This is a schematic diagram of the cylindrical structure provided in an embodiment of the present invention. Detailed Implementation

[0047] To make the objectives, technical solutions, and advantages of this invention clearer, the technical solutions of this invention will be clearly and completely described below. Obviously, the described embodiments are only a part of the embodiments of this invention, and not all of them. Based on the embodiments of this invention, all other embodiments obtained by those skilled in the art without creative effort are part of this invention.

[0048] Example 1:

[0049] This invention proposes a wide-energy-range neutron ambient dose equivalent monitoring device for implementing the aforementioned wide-energy-range neutron ambient dose equivalent monitoring method. The dose equivalent monitoring includes an outer moderator, a thermal neutron absorbing layer, a high-energy neutron compensation layer, and an inner moderator.

[0050] The high-energy neutron compensation layer is composed of a heavy metal material that can undergo a significant neutron multiplication reaction with high-energy neutrons;

[0051] The dose equivalent monitoring device contains an initial thermal neutron detector group deployed in the moderator in a first depth order for collecting thermal neutron fluence information under different levels of moderation; multiple groups of thermal neutron detectors optimized from the initial thermal neutron detector group are used for neutron ambient dose equivalent monitoring;

[0052] Thermal neutron detectors are radiation-resistant solid-state semiconductor detectors, comprising a silicon carbide semiconductor substrate and an enrichment layer attached to its surface. 6 Lithium fluoride thermal neutron converter.

[0053] The high-energy neutron compensation layer is made of lead, and the thermal neutron absorption layer is made of boron carbide. The initial thermal neutron detector group includes: a group of detectors arranged radially upwards at a distance of 1 cm from the front surface of the outer moderator (front, back, left, right, top, and bottom); and 8 groups of thermal neutron detectors arranged radially at equal intervals of 5 mm, with one detector arranged at the center of the inner moderator, for a total of 10 groups and 55 detectors.

[0054] The dose equivalent monitoring device comprises a multi-layered concentric spherical structure and a cylindrical structure; wherein, the structure of the multi-layered concentric spherical structure is as follows: Figure 1 As shown, the structure is spherical. 1 is the neutron moderator, used to reduce the energy of incident neutrons, forming thermal neutrons. 1.1 is the outer moderator, corresponding to the low-energy neutron dose response region. 1.2 is the inner moderator, corresponding to the fast neutron dose response region. 4 is the thermal neutron detector. One detector is located at the center of the inner moderator 1.2. The other detectors are arranged in a first-depth order with equal spacing in six radial directions (front, back, left, right, top, and bottom) according to their depth from the center of the moderator 1. The six detectors at the same depth are grouped together, and the average response value of each group is used as the thermal neutron response value at the corresponding depth of the moderator, thus improving the angular response consistency of the device. 2 is the thermal neutron absorption layer, used to absorb thermal neutrons formed from incident low-energy neutrons, reducing the response of the thermal neutron detectors in the inner moderator to incident low-energy neutrons. 3 is a high-energy neutron compensation layer, which utilizes the (n,xn) multiplication reaction between the metal material and fast neutrons to improve the response of the thermal neutron detector in the inner moderator to incident fast neutrons.

[0055] The neutron source used in the calculation was a circular monoenergetic neutron source, the size of which was the same as the front surface of the moderator, such as... Figure 1 The apparatus shown has a neutron source incident from the left side onto the front surface of the moderator. The incident neutron energy ranges from 0.01 eV to 1 GeV. 111 neutron energy points are taken at equal intervals in logarithmic coordinates. A thermal neutron detector records the response values ​​caused by the neutrons at each energy point incident at different depths in the moderator.

[0056] This invention, by incorporating a high-energy neutron compensation layer, utilizes its significant neutron multiplication reaction with ultra-high-energy neutrons to amplify the originally weak and difficult-to-detect signal of a single incident high-energy neutron into multiple secondary, lower-energy neutrons in a cascading manner. This fundamentally improves the detection efficiency and response sensitivity of the device in the ultra-high-energy region, which is key to extending the energy upper limit to 1 GeV. Secondly, the introduction of a thermal neutron absorption layer constructs a physical "low-pass filter," effectively shielding the internal high-energy signal region from interference from external low-energy neutrons and secondary thermal neutrons, achieving physical decoupling of the internal and external response regions, and ensuring the purity of high-energy event measurements. Furthermore, the use of a silicon carbide-based radiation-resistant detector ensures long-term stable operation of the device against the strong ionizing radiation background generated by high-energy particle showers, guaranteeing measurement reliability. Finally, the proposed "selection from the initial set" concept indicates the device's high optimization potential, aiming to achieve optimal performance with the most economical hardware configuration, offering greater design flexibility and efficiency compared to traditional fixed-layout instruments.

[0057] The procedure for monitoring the dose equivalent around neutrons in a wide energy range is as follows: Figure 2 As shown, it includes:

[0058] S10: Calculate the flux response values ​​of the initial thermal neutron detector array within the moderated body and the flux response function formed by the incident neutron energy, where the flux response function of the i-th detector is R. i (E), the average fluence response function of the i-th group of detectors in the spherical structure is expressed as follows:

[0059] ;

[0060] in, Let represent the flux response function of the j-th detector in the i-th group of detectors;

[0061] S20: Calculate the relative dose response function formed by comparing the fluence response function of the initial thermal neutron detector group with the ICRP74 neutron fluence-ambient dose equivalent conversion function. The expression for the relative dose response function of the i-th (group) detector of the dose equivalent monitoring device is as follows:

[0062] ;

[0063] in, Represents the relative dose response function. This represents the ICRP74 neutron dose-surround dose equivalent conversion function;

[0064] S30: Select an optimized detector combination from the initial thermal neutron detector group of the dose equivalent monitoring device, and multiply the relative dose response function matrix D of the detector combination with the corresponding linear combination coefficient vector w to obtain the total relative dose response function R(E). The formula for calculating the linear combination coefficient vector w is as follows:

[0065] ;

[0066] in, D yes k OK l The coefficient matrix of the column, l Each column in the column represents the flux response function of the selected detector (group of detectors). k Each value represents the flux response function within a specified energy range. k One response value; w It is a length of l The column vectors represent l The weighting coefficients corresponding to each response function; D ideal It is a length of k The column vector, with a value of 1, represents the flat relative dose response within the specified energy range;

[0067] S40: The neutron ambient dose equivalent is obtained by combining and outputting the counts of each detector in the detector combination selected from the dose equivalent monitoring device according to the solved linear combination coefficients. The expression for the neutron ambient dose equivalent output by the multi-detector combination is as follows:

[0068] ;

[0069] in, w i For the first i Weighting coefficients for individual (group) detectors; R i For the first i The average flux response of each (group of) detectors; C i For the first i The (average) count of each (group of) detectors; This represents the neutron fluence per unit energy.

[0070] The relative dose response functions of multiple sets of thermal neutron detectors arranged at different depths were calculated using the Monte Carlo method.

[0071] The optimized thermal neutron detector array, obtained from the initial array, involves the following steps:

[0072] S11: Traverse all combinations of 2 to n detectors from the initial thermal neutron detector group of the dose equivalent monitoring device, and calculate the total relative dose response function for each detector combination;

[0073] S12: Calculate the relative standard deviation of the total relative dose response function for all detector combinations. Take the combination with the smallest relative standard deviation of the total relative dose response function among all combinations of n detectors (groups) as the optimal combination of n detectors (groups). Obtain the curve of the relative standard deviation of the optimal combination as the number of detectors increases.

[0074] S13: Determine the minimum number of detectors required to approach the minimum relative standard deviation and the optimal combination method from the curve of the relative standard deviation of the optimal combination method as the number of detectors increases, as the final optimized thermal neutron detector combination.

[0075] The optimization results of multi-detector combination are used to guide the structural optimization of the device, including:

[0076] S21: The relative standard deviation of the optimized combination of multiple detectors and the total relative dose response function of the initial structure of the device is calculated;

[0077] S22: Based on the energy range of the flatness difference of the total relative dose response function of the multi-detector in the initial structure optimization, optimize the structural parameters for the corresponding energy range response compensation, and calculate the multi-detector combination optimization result under the optimized structural parameters;

[0078] S23: Evaluate the optimization results of the multi-detector combination of the optimized structure, and further optimize the structural parameters until the device structure with the flattest dose response is obtained.

[0079] In this embodiment, the outer moderator 1.1 has an inner diameter of 10.7 cm and an outer diameter of 12.7 cm, and the inner moderator 1.2 has a diameter of 5.7 cm. The thermal neutron detector 4 is a 4H-SiC detector containing a thermal neutron converter, with dimensions of 1 cm * 1 cm. 6 LiF, arranged close to the detector surface. 6 The LiF converter has a thickness of 5 μm and an enrichment of 90% for 6Li atoms. The thermal neutron absorber layer 2 is made of boron carbide, with an inner diameter of 6.2 cm and an outer diameter of 10.7 cm. The high-energy neutron compensation layer 3 is made of lead, with an inner diameter of 5.7 cm and an outer diameter of 6.2 cm.

[0080] The relative dose response values ​​detected by 10 sets of thermal neutron detectors at 111 energy nodes are multiplied by the linear combination coefficients, and then summed to obtain a constant value of 1 representing the flat relative dose response function. For example, when n is 3, ... Figure 3 As shown, selecting 3 out of 10 thermal neutron detectors results in 120 possible scenarios. The linear combination coefficients for each scenario are then calculated. In summary, the linear combination coefficients are calculated by taking the 111 response values ​​of each response function in the selected relative dose response function combination between 0.01 eV and 1 GeV as columns of the coefficient matrix, setting the constant term to 1 to represent an ideal flat relative dose response function, and solving the least-squares solution of this overdetermined system of equations as the linear combination coefficients for each detector.

[0081] According to an embodiment of the present invention, the method for selecting the optimal combination of relative dose response functions is as follows: 2 to 5 groups are selected from 10 groups of relative dose response functions of thermal neutron detectors, and the linear combination coefficients of all possible response function combinations are calculated. Then, these coefficients are linearly combined to obtain a flat relative dose response function, and the relative standard deviation of its relative dose response value in the energy range of 0.01 eV–1 GeV is calculated. Among combinations with the same number of relative dose response functions, the response function combination with the smallest relative standard deviation is selected, and the curve of the smallest relative standard deviation versus the number of relative dose response functions of the thermal neutron detector is plotted.

[0082] The curve showing the minimum relative standard deviation as a function of the number of response functions no longer decreases significantly after the number of response functions exceeds two. Therefore, the appropriate number of detectors is two sets, with the distances between the front surface of the detectors and the front surface of the device being 1 cm and 12.7 cm, respectively, one set at the outer moderator and one at the center of the sphere, for a total of seven detectors. The corresponding weight values ​​for the detectors are 2652.7 and 45025.7, respectively. The total relative dose response function obtained by summing the relative dose response functions of each detector according to the optimized detector combination and weighted calculation is as follows: Figure 3 As shown, the relative standard deviation is 18.6%, and the relative dose response range is 0.54–1.49. The total dose response function obtained by summing the fluence response functions of each detector according to the optimized detector combination and weighted average is shown below. Figure 4 As shown, the total relative dose response function is as follows: Figure 5 As shown, the average deviation from the ICRP74 conversion function is 13.1%.

[0083] This invention creatively transforms the detector configuration problem into a solvable mathematical optimization problem, namely, solving a least-squares solution to an overdetermined system of equations aimed at achieving an ideal flat response. This method completely eliminates the reliance on engineer experience in traditional design, replacing it with rigorous and repeatable mathematical algorithms, thus achieving scientific and automated design.

[0084] Secondly, by traversing all possible combinations of different numbers of detectors and using the relative standard deviation (RSD) of the total relative dose response function as the sole evaluation metric, this method can perform global optimization in a vast solution space, ensuring that it finds not only a usable solution but also the theoretically optimal one. Finally, by selecting the combination corresponding to the "minimum relative standard deviation," this method establishes a quantitative decision criterion, systematically balancing the performance and cost of the device, and achieving optimal response flatness with the minimum number of detectors.

[0085] Ultimately, this method enables the device to achieve an excellent RSD of 18.6% using only 2 sets (7 detectors), fully demonstrating its significant technical advantages in improving performance, reducing costs, and increasing design efficiency.

[0086] Example 2:

[0087] This invention proposes a wide-energy-range neutron ambient dose equivalent monitoring device. The device is a multi-layered concentric spherical structure, comprising, radially from the outside to the inside: an outer moderator, a thermal neutron absorbing layer, a high-energy neutron compensation layer, and an inner moderator. The high-energy neutron compensation layer is composed of a heavy metal material capable of undergoing a significant neutron multiplication reaction with ultra-high-energy neutrons. The thermal neutron detector is a radiation-resistant solid-state semiconductor detector, comprising a silicon carbide semiconductor substrate and an enrichment layer closely attached to its surface. 6 Lithium fluoride thermal neutron converter;

[0088] This embodiment focuses on the analysis within the range of 0.01 eV to 20 MeV, specifically including:

[0089] The outer moderator has a body diameter of 10.7 cm and an outer diameter of 12.7 cm, while the inner moderator has a diameter of 4.7 cm. The thermal neutron detector is a 4H-SiC detector containing a thermal neutron converter, measuring 1 cm x 1 cm. 6 LiF, arranged close to the detector surface. 6 The LiF conversion body has a thickness of 5 μm. 6 The enrichment of Li atoms is 90%. A set of detectors is arranged radially upwards at a distance of 1 cm from the front surface of the outer moderator (front, back, left, right, top, and bottom). A single detector is placed at the center of the inner moderator, and seven sets of thermal neutron detectors are arranged alternately at 5 mm intervals along the radial depth, for a total of 9 sets and 49 detectors. The thermal neutron absorption layer 2 is made of boron carbide, with an inner diameter of 5.7 cm and an outer diameter of 10.7 cm. The high-energy neutron compensation layer 3 is made of tungsten, with an inner diameter of 4.7 cm and an outer diameter of 5.7 cm.

[0090] The neutron source used in the calculation was a circular monoenergetic neutron source, the size of which was the same as the front surface of the moderator, such as... Figure 1The apparatus shown has a neutron source incident from the left side onto the front surface of the moderator. The incident neutron energy ranges from 0.01 eV to 20 MeV. 94 neutron energy points are taken at equal intervals in logarithmic coordinates. The response value caused by the neutrons at each energy point being incident at different depths of the moderator is recorded by a thermal neutron detector.

[0091] The relative dose response values ​​at 94 energy nodes detected by 9 sets of thermal neutron detectors are multiplied by the linear combination coefficients, and then summed to obtain a constant value of 1 representing the flat relative dose response function. For example, when n is 3, if 3 sets of thermal neutron detectors are selected, there are 84 possible cases. The linear combination coefficients for each case are then solved.

[0092] In summary, the method for calculating the linear combination coefficients is to take the 94 response values ​​of each response function in the selected combination of relative dose response functions between 0.01eV and 20MeV as columns of the coefficient matrix, set the constant term to 1 to represent the ideal flat relative dose response function, and solve the least squares solution of the overdetermined equation system as the linear combination coefficients of each detector.

[0093] According to an embodiment of the present invention, the method for selecting the optimal combination of relative dose response functions is to select 2 to 5 groups from 9 groups of relative dose response functions of thermal neutron detectors and calculate the linear combination coefficients of all possible combinations of response functions.

[0094] Then, a flat relative dose response function is obtained by linearly combining these coefficients, and the relative standard deviation of its relative dose response value in the energy range of 0.01 eV-20 MeV is calculated. Among the combinations with the same number of relative dose response functions, the combination with the smallest relative standard deviation is selected, and the curve of the smallest relative standard deviation as a function of the number of relative dose response functions of the thermal neutron detector is plotted.

[0095] The curve showing the minimum relative standard deviation as a function of the number of response functions no longer decreases significantly after the number of response functions exceeds two. Therefore, the appropriate number of detectors is two groups, with the distances between the front surface of the detectors and the front surface of the device being 1 cm and 12.7 cm, respectively, one group at the outer moderator and one at the center of the sphere, for a total of seven detectors. The corresponding weight values ​​for the detectors are 2685.6 and 65564.7, respectively. The total relative dose response function, obtained by summing the relative dose response functions of each detector using the optimized detector combination and weighted calculation, has a relative standard deviation of 12.2% and a relative dose response range of 0.69–1.24. The average deviation of the total dose response function obtained by summing the fluence response functions of each detector using the optimized detector combination and weighted calculation from the ICRP74 conversion function is 9.4%.

[0096] This invention arranges a sufficient number of thermal neutron detectors in a moderator to fully acquire the detailed response functions at different depths of the moderator. It iterates through the combination methods under different numbers of response functions, solves the total dose response function under the optimal combination of different numbers of detectors (the combination with the smallest relative standard deviation of the total relative dose response function), and determines the optimal number of detectors and the corresponding combination method based on the relationship between the relative standard deviation of the total relative dose response function and the number of detectors. The combined output obtains the ideal total dose response function.

[0097] Example 3:

[0098] This invention proposes a wide-energy-range neutron ambient dose equivalent monitoring device, the device comprising an outer moderator, a thermal neutron absorbing layer, a high-energy neutron compensation layer, and an inner moderator;

[0099] The high-energy neutron compensation layer is composed of a heavy metal material capable of undergoing a significant neutron multiplication reaction with ultra-high-energy neutrons; the thermal neutron detector is a radiation-resistant solid-state semiconductor detector, comprising a silicon carbide semiconductor substrate and an enrichment layer closely attached to its surface. 6 Lithium fluoride thermal neutron converter;

[0100] The device is equipped with an initial thermal neutron detector group for collecting thermal neutron flux information under different slowing degrees; multiple groups of thermal neutron detectors of the first depth order are selected from the initial thermal neutron detector group for equivalent monitoring.

[0101] In this embodiment, the device has a cylindrical structure, specifically as follows: Figure 6 As shown.

[0102] Figure 6 This is a cross-sectional view of the multi-detector neutron ambient dose equivalent meter structure, which is cylindrical. 1 is the neutron moderator, used to reduce the energy of incident neutrons, forming thermal neutrons. 1.1 is the outer moderator, corresponding to the low-energy neutron dose response region, and 1.2 is the inner moderator, corresponding to the high-energy neutron dose response region. 4 is the thermal neutron detector, arranged in first depth order and at equal intervals according to the neutron incident depth in moderator 1, to obtain the response values ​​of thermal neutrons at different depths within the moderator. 2 is the thermal neutron absorption layer, used to absorb thermal neutrons formed from incident low-energy neutrons, reducing the response of the thermal neutron detector in the inner moderator to incident low-energy neutrons. 3 is the high-energy neutron compensation layer, which utilizes the (n,xn) multiplication reaction between the metallic material and high-energy neutrons to improve the response of the thermal neutron detector in the inner moderator to incident high-energy neutrons. 5 is a neutron-scattering shielding layer, used to absorb scattered neutrons incident from the side and rear, preventing scattered neutrons from entering the thermal neutron detector and generating a count, thus improving the counting accuracy of the thermal neutron detector for neutrons incident orthogonally on the front surface of the moderator.

[0103] Neutron moderator 1 is made of polyethylene, with the outer moderator 1.1 having a diameter of 10 cm and a height of 4.5 cm, and the inner moderator 1.2 having a diameter of 10 cm and a height of 7.2 cm. Thermal neutron detector 4 is a 4H-SiC detector containing a thermal neutron converter, measuring 1 cm x 1 cm. 6 LiF, arranged close to the detector surface. 6 The LiF conversion body has a thickness of 5 μm. 6 The enrichment of Li atoms is 90%. Eighteen thermal neutron detectors are staggered along the axis at 5mm depths in both the inner and outer moderator layers. Thermal neutron absorption layer 2 is made of boron carbide, with a diameter of 10cm and a height of 2.5cm. High-energy neutron compensation layer 3 is made of lead, with a diameter of 10cm and a height of 0.8cm. The scattering neutron shielding layer 5 is made of boron-containing polyethylene with a boron-10 content of 40%. The shielding layer is annular and coaxial with the cylindrical moderator, with an outer diameter of 16cm, an inner diameter of 10cm, and a height of 18cm. The thickness of the bottom and walls of the barrel is 3cm.

[0104] The neutron source used in the calculation was a circular monoenergetic neutron source, the size of which was the same as the front surface of the moderator, such as... Figure 6 The apparatus shown has a neutron source incident from the left side onto the front surface of the moderator. The incident neutron energy ranges from 0.01 eV to 1 GeV. 111 neutron energy points are taken at equal intervals in logarithmic coordinates. A thermal neutron detector records the response values ​​caused by the neutrons at each energy point incident at different depths in the moderator.

[0105] The relative dose response values ​​at 111 energy nodes detected by 18 thermal neutron detectors are multiplied by the linear combination coefficients, and then summed to obtain a constant value of 1 representing the flat relative dose response function. For example, when n is 3, selecting 3 of the 18 thermal neutron detectors results in 816 possible cases. The linear combination coefficients for each case are then calculated. In summary, the linear combination coefficients are calculated by taking the 111 response values ​​of each response function in the selected combination of relative dose response functions between 0.01 eV and 1 GeV as columns of the coefficient matrix, setting the constant term to 1 to represent the ideal flat relative dose response function, and solving the least-squares solution of this overdetermined system of equations as the linear combination coefficients for each detector.

[0106] According to an embodiment of the present invention, the method for selecting the optimal combination of relative dose response functions is as follows: 2 to 5 relative dose response functions are selected from 18 thermal neutron detectors, and the linear combination coefficients of all possible response function combinations are calculated. These coefficients are then linearly combined to obtain a flat relative dose response function, and the relative standard deviation of its relative dose response value in the energy range of 0.01 eV–1 GeV is calculated. Among combinations with the same number of relative dose response functions, the response function combination with the smallest relative standard deviation is selected, and the curve of the smallest relative standard deviation versus the number of relative dose response functions of the thermal neutron detector is plotted.

[0107] The curve showing the minimum relative standard deviation as a function of the number of response functions no longer decreases significantly after the number of response functions exceeds four. Therefore, the appropriate number of detectors is four, with distances between the front surface of the detector and the front surface of the device being 1 cm, 8.3 cm, 9.7 cm, and 11.3 cm, respectively. The corresponding weight values ​​for the detectors are 356.6, -2734335.2, 2812225.0, and -1073624.7, respectively. The total relative dose response function, obtained by summing the relative dose response functions of each detector using the optimized detector combination and weighted calculation, has a relative standard deviation of 15.8% and a relative dose response range of 0.62–1.39. The average deviation of the total dose response function obtained by summing the fluence response functions of each detector using the optimized detector combination and weighted calculation from the ICRP74 conversion function is 11.9%.

[0108] Example 4:

[0109] This invention proposes a wide-energy-range neutron ambient dose equivalent monitoring device, the device comprising an outer moderator, a thermal neutron absorbing layer, a high-energy neutron compensation layer, and an inner moderator;

[0110] The high-energy neutron compensation layer is composed of a heavy metal material capable of undergoing a significant neutron multiplication reaction with ultra-high-energy neutrons; the thermal neutron detector is a radiation-resistant solid-state semiconductor detector, comprising a silicon carbide semiconductor substrate and an enrichment layer closely attached to its surface. 6 Lithium fluoride thermal neutron converter;

[0111] The device is equipped with an initial thermal neutron detector group for collecting thermal neutron flux information under different slowing degrees; multiple groups of thermal neutron detectors of the first depth order are selected from the initial thermal neutron detector group for equivalent monitoring.

[0112] In this embodiment, the device has a cylindrical structure;

[0113] The outer moderator 1.1 has a diameter of 10 cm and a height of 4 cm, while the inner moderator 1.2 has a diameter of 10 cm and a height of 6 cm. The thermal neutron detector 4 is a 4H-SiC detector containing a thermal neutron converter, with dimensions of 1 cm x 1 cm. 6 LiF, arranged close to the detector surface. 6The LiF converter is 5 μm thick, with a 6Li atom enrichment of 90%. Fifteen thermal neutron detectors are staggered along the axis at 5 mm depths in both the inner and outer moderator layers. Thermal neutron absorption layer 2 is made of boron carbide, with a diameter of 10 cm and a height of 4 cm. High-energy neutron compensation layer 3 is made of tungsten, with a diameter of 10 cm and a height of 1 cm. Scattering neutron shielding layer 5 is made of boron-containing polyethylene with a boron-10 content of 20%. The shielding layer is annular and coaxial with the cylindrical moderator, with an outer diameter of 16 cm, an inner diameter of 10 cm, and a height of 18 cm. The thickness of the bottom and walls of the barrel is 3 cm.

[0114] The neutron source used in the calculation was a circular monoenergetic neutron source, the size of which was the same as the front surface of the moderator, such as... Figure 6 The apparatus shown has a neutron source incident from the left side onto the front surface of the moderator. The incident neutron energy ranges from 0.01 eV to 20 MeV. 94 neutron energy points are taken at equal intervals in logarithmic coordinates. The response value caused by the neutrons at each energy point being incident at different depths of the moderator is recorded by a thermal neutron detector.

[0115] The product of the relative dose response values ​​at 94 energy nodes detected by 15 thermal neutron detectors and the linear combination coefficients is set, and then summed to obtain a constant value of 1 representing the flat relative dose response function. For example, when n is 3, there are 455 possible cases when 3 of the 15 thermal neutron detectors are selected. The linear combination coefficients for each case are solved.

[0116] In summary, the method for calculating the linear combination coefficients is to take the 94 response values ​​of each response function in the selected combination of relative dose response functions between 0.01eV and 20MeV as columns of the coefficient matrix, set the constant term to 1 to represent the ideal flat relative dose response function, and solve the least squares solution of the overdetermined equation system as the linear combination coefficients of each detector.

[0117] According to an embodiment of the present invention, the method for selecting the optimal combination of relative dose response functions is as follows: 2 to 5 relative dose response functions are selected from 15 thermal neutron detectors, and the linear combination coefficients of all possible response function combinations are calculated. These coefficients are then linearly combined to obtain a flat relative dose response function, and the relative standard deviation of its relative dose response value in the energy range of 0.01 eV–20 MeV is calculated. Among combinations with the same number of relative dose response functions, the response function combination with the smallest relative standard deviation is selected, and the curve of the smallest relative standard deviation versus the number of relative dose response functions of the thermal neutron detector is plotted.

[0118] The curve showing the minimum relative standard deviation as a function of the number of response functions no longer decreases significantly after the number of response functions exceeds three. Therefore, the appropriate number of detectors is three, with distances between the front surface of the detector and the front surface of the device of 1 cm, 11.5 cm, and 13.5 cm, respectively. The corresponding weight values ​​for the detectors are 360.0, 1528227.8, and -1689899.4, respectively. The total relative dose response function, obtained by summing the relative dose response functions of each detector using the optimized detector combination and weighted calculation, has a relative standard deviation of 9.2% and a relative dose response range of 0.79–1.28. The average deviation of the total dose response function obtained by summing the fluence response functions of each detector using the optimized detector combination and weighted calculation from the ICRP74 conversion function is 6.6%.

[0119] Finally, it should be noted that the above embodiments are only used to illustrate the technical solutions of the present invention, and not to limit them; although the present invention has been described in detail with reference to the foregoing embodiments, those skilled in the art should understand that modifications can still be made to the technical solutions described in the foregoing embodiments, or equivalent substitutions can be made to some or all of the technical features; and these modifications or substitutions do not cause the essence of the corresponding technical solutions to deviate from the technical solutions of the embodiments of the present invention.

Claims

1. A method for monitoring dose equivalent around neutrons in a wide energy range, characterized in that, The dose equivalent around neutrons is output through a combination of multiple detectors, including: S10: Calculate the flux response values ​​of the initial thermal neutron detector array within the moderated body and the flux response function formed by the incident neutron energy, where the flux response function of the i-th detector is R. i (E), the average fluence response function of the i-th group of detectors in the spherical structure is expressed as follows: ； in, Let represent the flux response function of the j-th detector in the i-th group of detectors; S20: Calculate the relative dose response function formed by comparing the fluence response function of the initial thermal neutron detector group with the ICRP74 neutron fluence-ambient dose equivalent conversion function. The expression for the relative dose response function of the i-th (group) detector of the dose equivalent monitoring device is as follows: ； in, Represents the relative dose response function. This represents the ICRP74 neutron dose-surround dose equivalent conversion function; S30: Select an optimized detector combination from the initial thermal neutron detector group of the dose equivalent monitoring device, and multiply the relative dose response function matrix D of the detector combination with the corresponding linear combination coefficient vector w to obtain the total relative dose response function R(E). The formula for calculating the linear combination coefficient vector w is as follows: ； in, D yes k OK l The coefficient matrix of the column, l Each column in the column represents the flux response function of the selected detector (group of detectors). k Each value represents the flux response function within a specified energy range. k One response value; w It is a length of l The column vectors represent l The weighting coefficients corresponding to each response function; D ideal It is a length of k The column vector, with a value of 1, represents the flat relative dose response within the specified energy range; S40: The neutron ambient dose equivalent is obtained by combining and outputting the counts of each detector in the detector combination selected from the dose equivalent monitoring device according to the solved linear combination coefficients. The expression for the neutron ambient dose equivalent output by the multi-detector combination is as follows: ； in, w i For the first i Weighting coefficients for individual (group) detectors; R i For the first i The average flux response of each (group of) detectors; C i For the first i The (average) count of each (group of) detectors; This represents the neutron fluence per unit energy.

2. The method for monitoring the dose equivalent around neutrons in a wide energy range according to claim 1, characterized in that, The optimized thermal neutron detector array, obtained from the initial array, involves the following steps: S11: Traverse all combinations of 2 to n detectors from the initial thermal neutron detector group of the dose equivalent monitoring device, and calculate the total relative dose response function for each detector combination; S12: Calculate the relative standard deviation of the total relative dose response function for all detector combinations. Take the combination with the smallest relative standard deviation of the total relative dose response function among all combinations of n detectors (groups) as the optimal combination of n detectors (groups). Obtain the curve of the relative standard deviation of the optimal combination as the number of detectors increases. S13: Determine the minimum number of detectors required to approach the minimum relative standard deviation and the optimal combination method from the curve of the relative standard deviation of the optimal combination method as the number of detectors increases, as the final optimized thermal neutron detector combination.

3. The method for monitoring the dose equivalent around neutrons in a wide energy range according to claim 2, characterized in that, The optimization results of multi-detector combination are used to guide the structural optimization of the device, including: S21: The relative standard deviation of the optimized combination of multiple detectors and the total relative dose response function of the initial structure of the device is calculated; S22: Based on the energy range of the flatness difference of the total relative dose response function of the multi-detector in the initial structure optimization, optimize the structural parameters for the corresponding energy range response compensation, and calculate the multi-detector combination optimization result under the optimized structural parameters; S23: Evaluate the optimization results of the multi-detector combination of the optimized structure, and further optimize the structural parameters until the device structure with the flattest dose response is obtained.

4. A wide-energy-range neutron ambient dose equivalent monitoring device, characterized in that, For implementing the wide-energy-range neutron ambient dose equivalent monitoring method according to any one of claims 1-3, the dose equivalent monitoring device includes an outer moderator, a thermal neutron absorbing layer, a high-energy neutron compensation layer, and an inner moderator; The high-energy neutron compensation layer is composed of a heavy metal material that can undergo a significant neutron multiplication reaction with high-energy neutrons; The dose equivalent monitoring device contains an initial thermal neutron detector group deployed in the moderator in a first depth order for collecting thermal neutron fluence information under different levels of moderation; multiple groups of thermal neutron detectors optimized from the initial thermal neutron detector group are used for neutron ambient dose equivalent monitoring; Thermal neutron detectors are radiation-resistant solid-state semiconductor detectors, comprising a silicon carbide semiconductor substrate and an enrichment layer attached to its surface. 6 Lithium fluoride thermal neutron converter.

5. A wide-energy-range neutron ambient dose equivalent monitoring device according to claim 4, characterized in that: The high-energy neutron compensation layer is made of lead and tungsten, and the thermal neutron absorption layer is made of boron carbide.

6. A wide-energy-range neutron ambient dose equivalent monitoring device according to claim 5, characterized in that, The dose equivalent monitoring device comprises a multi-layered concentric spherical structure and a cylindrical structure.

7. A wide-energy-range neutron ambient dose equivalent monitoring device according to claim 6, characterized in that, The initial thermal neutron detector group is configured as follows: when the dose equivalent monitoring device is a multi-layer concentric spherical structure, a set of thermal neutron detectors is arranged 1 cm away from the front surface in the radial direction of the outer layer moderator, in front, behind, left, right, top and bottom, and one detector is arranged in the center of the inner layer moderator and multiple sets of thermal neutron detectors are arranged alternately with equal radial depth intervals of 5 mm. When the dose equivalent monitoring device is a cylindrical structure, multiple thermal neutron detectors are arranged alternately at equal intervals of 5 mm along the axial depth in the inner and outer layers of the moderator.

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