Personal dosimeter for neutron and photon mixed radiation detection and detection method

Abstract

The invention relates to a personal dosimeter for neutron and photon mixed radiation detection and a detection method, the dosimeter comprises a base, an upper cover and three groups of detection assemblies, and the base is provided with three grooves which are distributed in parallel and respectively accommodate a first detection assembly, a second detection assembly and a third detection assembly. The first detection assembly and the second detection assembly each comprise a neutron absorber with grooves formed in the upper portion and the lower portion and a 6Li thermoluminescence detector pair and a 7Li thermoluminescence detector pair, and separated measurement of the dose of thermal neutrons, medium-high energy neutrons and gamma rays is achieved. The third detection assembly comprises a neutron moderator and a solid track detector. Separation measurement of neutrons with different energies is realized through track density difference. The upper cover is provided with a measuring window, and tissue equivalent materials with different mass thicknesses are adopted and correspond to the dose depth of Hp (10) and the dose depth of Hp (3). Through combination of multi-channel reading and correction coefficient calculation, dose evaluation of gamma rays, thermal neutrons and fast neutrons on different tissue depths can be realized. The device is compact in structure, and is suitable for personal dose monitoring in a complex radiation environment.

Description

Technical Field

[0001] This invention relates to the field of ionizing radiation dosimetry and protection technology, and in particular to a personal dosimeter and detection method for detecting mixed neutron and photon radiation. Background Technology

[0002] When ionizing radiation enters the human body, it reacts with cells and tissues through physical, chemical, and biological processes, potentially triggering both stochastic and deterministic tissue effects. In severe cases, it can lead to genetic damage or impaired tissue function. Therefore, to ensure the legitimacy of radiation practices, the optimization of protection, and the reasonable control of dose limits, accurate and traceable monitoring and assessment of individual radiation doses are essential.

[0003] In applications such as nuclear fuel cycle, industrial radiation source management, medical radiotherapy, and basic radiation research, neutrons are often a significant radiation source, and may even constitute a major source of radiation dose. Compared to photons or electrons, the radiation weighting factor (W) of neutrons is higher. R The concentrations are typically higher, ranging from 2 to over 20, reflecting a significantly stronger degree of harm to biological tissues. However, because neutrons are electrically neutral particles, their interactions with matter are more complex, with a lower reaction cross-section. Furthermore, they are easily affected by photon deposition dose interference in mixed radiation fields (n, γ), leading to technical challenges in neutron dose monitoring, such as inaccurate and undetectable measurements.

[0004] According to relevant international protection requirements and my country's standard guidelines in GBZ 128-2019 Occupational External Exposure Personal Monitoring Specification, in neutron-photon mixed radiation fields, it is necessary to use a discriminative personal dosimeter that can measure neutron dose and photon dose separately, and assess the personal dose equivalent of neutron and photon separately to obtain a total dose evaluation value.

[0005] Currently, passive dosimeters are the most commonly used methods for neutron dose monitoring, including albedosimeters, solid track neutron dosimeters, and bubble dosimeters. Among these, albedometers are simple in structure and low in cost, but have poor energy response, are highly susceptible to gamma-ray interference, and their neutron response is often submerged in the photon response. Solid track dosimeters are suitable for medium- and high-energy neutron fields, but their thermal neutron response is low, and their readout process is complex. Bubble dosimeters have good energy response, but are sensitive to temperature changes, have poor vibration resistance, and their use is limited. Existing single-type neutron dosimeters generally suffer from narrow response ranges, large gamma-ray interference, and high measurement uncertainty, which can easily lead to misjudgment, underestimation, or overestimation of individual neutron dose.

[0006] In view of the above problems, this invention is proposed. Summary of the Invention

[0007] This invention discloses a personal dosimeter and detection method for detecting mixed neutron and photon radiation, aiming to solve the technical problems existing in the prior art.

[0008] To achieve the above objectives, according to one aspect of the present invention, a personal dosimeter for detecting mixed neutron and photon radiation is provided, comprising a dosimeter base, a first detection component, a second detection component, a third detection component, and a dosimeter cover. The dosimeter base and the dosimeter cover are fastened together to form a sealed cavity. The dosimeter base is provided with a first groove, a second groove, and a third groove that are arranged in parallel to each other, each accommodating three sets of detection components.

[0009] The first detection component and the second detection component are arranged adjacent to each other on one side. The first detection component includes a first neutron absorber enclosed by a four-sided frame. The center of the first neutron absorber is a rectangular platform with grooves on its upper and lower sides for encapsulating the upper and lower distributed second thermoluminescent detectors and the first thermoluminescent detector. The first thermoluminescent detector is in contact with the first groove. The second detection component includes a second neutron absorber enclosed by a four-sided frame. The center of the second neutron absorber is also a rectangular platform with grooves on its upper and lower sides for encapsulating the upper and lower distributed fourth thermoluminescent detectors and the third thermoluminescent detector. The third thermoluminescent detector is in contact with the second groove. All four thermoluminescent detectors are... 6 Li and 7 The paired design of Li detector elements enables dose decoupling and quantitative differentiation of neutrons and photons;

[0010] The third detection component is disposed adjacent to the second detection component on the other side, and includes a solid track detector and a neutron moderator. The neutron moderator is disposed above the solid track detector and covers part of its area.

[0011] The dosimeter cover is equipped with a measurement window. The dosimeter cover and the measurement window are made of tissue equivalent materials of different mass and thickness, respectively, corresponding to different standard dose depths.

[0012] As a preferred technical solution, the dosimeter cover and the measuring window are respectively made of materials with a thickness of 1000 mg / cm. 2 With 300mg / cm 2 Tissue equivalent materials are used for simultaneous dose measurement of deep subcutaneous tissues and the lens of the eye.

[0013] As a preferred technical solution 6 Li and 7 Li detection elements are all cylindrical, cuboid, or disc-shaped, and are arranged in close contact or adjacent to each other.

[0014] As a preferred technical solution, the neutron moderator is a hydrogen-rich polymer.

[0015] As a preferred technical solution, the neutron moderator is made of one of high-density polyethylene (HDPE), polyoxymethylene, paraffin or hydrogenated silicone rubber.

[0016] As a preferred technical solution, both the first neutron absorber and the second neutron absorber are made of单质, compounds and mixtures of neutron absorption materials such as cadmium and boron.

[0017] According to another aspect of the present invention, there is also provided a method for detecting mixed neutron and photon radiation based on the above dosimeter, including:

[0018] Record the readings of the solid track detector, establish a response model by comparing the track densities of the two regions on the solid track detector covered and not covered by the neutron moderator, and fit to obtain the neutron response correction coefficient S1 for different energies;

[0019] Record the readings of the second thermoluminescent detector and the first thermoluminescent detector, and fit according to the difference to obtain the correction coefficient S2;

[0020] Record the readings of the fourth thermoluminescent detector and the third thermoluminescent detector, and fit according to the difference to obtain the correction coefficient S3;

[0021] Based on the above readings and correction coefficients, calculate the personal dose equivalent Hp(10) and Hp(3).

[0022] As a preferred technical solution, when calculating the personal dose equivalent Hp(10):

[0023] Based on the readings of the solid track detector, the first thermoluminescent detector and the second thermoluminescent detector, perform multi-channel fitting by combining the correction coefficients S1, S2 and S3 to obtain the deep tissue dose Hp(10)_light caused by gamma rays and the deep tissue dose Hp(10)_neutron caused by neutrons, and calculate Hp(10) according to the following formula,

[0024] Hp(10) = Hp(10)_light + Hp(10)_neutron.

[0025] As a preferred technical solution, when calculating the personal dose equivalent Hp(3): ​​​​​​​The technical solution adopted in this invention can achieve at least one of the following beneficial effects:

[0029] 1. This invention employs paired configurations 6 Li and 7 The Li thermoluminescent detector element, combined with a neutron absorber with a double-groove structure, is packaged and arranged to fully utilize the advantages of this technology. 6 Li's high sensitivity to thermal neutrons and 7 Li's selective response to gamma rays effectively distinguishes between thermoluminescent signals caused by thermal neutrons and gamma rays, enabling dose decoupling and quantitative differentiation between the two in a mixed radiation field, and significantly improving the accuracy and reliability of dose measurement in a combined field.

[0030] 2. This invention, by setting up a third detection component including a neutron moderator and a solid track detector, utilizes the moderator's effect on fast neutrons and the track density difference between the moderated and unmoderated regions to derive the response of neutrons of different energies and establish a correction model. Simultaneously, by combining the thermoluminescence component and comparing the three sets of readings, multi-channel fitting is used to obtain the dose components of γ, thermal neutrons, and fast neutrons, achieving independent and correctable responses of the three types of radiation particles during dose measurement, thus meeting the needs of refined protection assessment in complex radiation fields.

[0031] 3. This invention collects readings from multiple thermoluminescent detectors and solid track detectors to establish response relationships related to thermal neutrons, fast neutrons, and gamma rays, respectively. Multi-channel fitting is performed using three correction coefficients, S1, S2, and S3, to calculate the dose contributions of photons and neutrons in Hp(10) and Hp(3), respectively, thereby improving the accuracy, stability, and algorithm controllability of dose measurement in complex neutron / photon combined radiation fields.

[0032] 4. The dosimeter cover and measuring window of the present invention are respectively made of materials with a thickness of 1000 mg / cm. 2 and 300mg / cm 2 The tissue equivalent materials correspond to the Hp(10) and Hp(3) dose depth standards defined by ICRP, respectively, which can simultaneously assess the radiation dose of deep organs and superficial tissues (such as the lens of the eye), effectively covering the relevant tissue exposure situations that may be received during actual wear, and enhancing the authority and medical comparability of the test results. Attached Figure Description

[0033] To more clearly illustrate the technical solutions of the embodiments of the present invention, the accompanying drawings used in the description of the embodiments will be briefly introduced below, forming part of the present invention. The illustrative embodiments of the present invention and their descriptions explain the present invention and do not constitute an improper limitation of the present invention. In the accompanying drawings:

[0034] Figure 1This is a schematic diagram of the structure of a personal dosimeter for detecting mixed neutron and photon radiation according to the present invention;

[0035] Figure 2 This is a schematic flowchart of the neutron and photon mixed radiation detection method of the present invention.

[0036] Explanation of reference numerals in the attached figures:

[0037] 1. Dosimeter base; 11. First groove; 12. Second groove; 13. Third groove; 2. First detection component; 21. First thermoluminescent detector; 22. First neutron absorber; 23. Second thermoluminescent detector; 3. Second detection component; 31. Third thermoluminescent detector; 32. Second neutron absorber; 33. Fourth thermoluminescent detector; 4. Third detection component; 41. Solid track detector; 42. Neutron moderator; 5. Dosimeter top cover; 6. Measurement window. Detailed Implementation

[0038] 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 in conjunction with specific embodiments and corresponding drawings. In the description of this invention, it should be noted that the term "or" is generally used to include the meaning of "and / or," unless otherwise expressly indicated.

[0039] In the description of this invention, it should be noted that, unless otherwise explicitly specified and limited, the terms "installation," "connection," and "linking" should be interpreted broadly. For example, they can refer to a fixed connection, a detachable connection, or an integral connection; they can refer to a mechanical connection or a magnetic connection; they can refer to a direct connection or an indirect connection through an intermediate medium; and they can refer to the internal communication between two components. Those skilled in the art can understand the specific meaning of the above terms in this invention based on the specific circumstances. Furthermore, in the description of this application, the terms "first," "second," etc., are used only for distinguishing descriptions and should not be construed as indicating or implying relative importance.

[0040] Obviously, the described embodiments are only some, not all, of the embodiments of the present invention. All other embodiments obtained by those skilled in the art based on the embodiments of the present invention without inventive effort are within the scope of protection of the present invention.

[0041] First, to facilitate understanding of the embodiments of the present invention, some terms or nouns involved in the present invention will be explained below:

[0042] Hp(10) is an important dose measure in radiation dosimetry, used to assess the radiation protection effect of external exposure on personnel. It refers to the personal dose equivalent at a depth of 10 mm from the skin surface in the equivalent soft tissue of the human body.

[0043] Hp(3) refers to the individual dose equivalent at a depth of 3 mm from the skin surface in the equivalent soft tissue of the human body.

[0044] Solid track detectors are polymer detectors that can generate permanent microscopic damage tracks through high-energy particle impacts. When these detectors are subjected to secondary charged particles (such as protons and alpha particles) produced by the reaction of fast neutrons and a moderation medium, they will form hidden tracks in their structure. These tracks can be amplified by chemical corrosion and observed and counted under a microscope, thereby deducing the fast neutron flux and dose.

[0045] To address the problems existing in the prior art, embodiments of the present invention provide a personal dosimeter for detecting mixed neutron and photon radiation, such as... Figure 1 As shown, the device includes a dosimeter base 1, a first detection component 2, a second detection component 3, a third detection component 4, and a dosimeter cover 5. The dosimeter base 1 and the dosimeter cover 5 are interlocked to form a sealed cavity for accommodating the detection components. In this embodiment, the interlocking method can employ a snap-fit ​​structure, threaded connection, screw fixing, or magnetic combination structure to ensure good mechanical stability and sealing performance of the overall structure under prolonged wear or vibration impact conditions. Preferably, the dosimeter base 1 has an annular flange or boss structure around its periphery, which cooperates with the corresponding groove or snap-fit ​​structure on the inner side of the dosimeter cover 5 to achieve quick and reliable snap-fit ​​fixing. To further enhance the sealing effect, rubber rings or silicone sealing gaskets can be embedded in the contact areas to effectively prevent external impurities such as dust and moisture from entering the internal detection cavity, ensuring long-term stable operation of the device. Both the dosimeter base 1 and the dosimeter cover 5 are made of tissue-equivalent materials or polymer materials to ensure good tissue equivalence of the dosimeter during actual wear and use. The aforementioned tissue equivalence refers to the fact that the interaction characteristics (such as absorption cross section, scattering cross section, and energy response) of the selected material with neutrons and photons when exposed to radiation are similar to those of human tissue, thereby ensuring good consistency between the dose received by the detector and the actual biological dose received by the wearer. Preferably, the tissue equivalent material can be polymethyl methacrylate (PMMA), polyethylene (PE), polycarbonate (PC), epoxy resin, or other radiation-responsive calibrated polymer materials. These materials have similar properties to human soft tissue in terms of mass density, hydrogen content, and atomic number composition, making them particularly suitable for dose response simulation under gamma-ray and slow neutron conditions.

[0046] The dosimeter base 1 includes a first groove 11, a second groove 12, and a third groove 13, which are used to accommodate the first detection component 2, the second detection component 3, and the third detection component 4, respectively, to achieve a separate detection arrangement for gamma rays and neutrons. The three grooves are parallel and evenly distributed on the dosimeter base 1. By setting up a multi-groove structure, the various detection units do not interfere with each other in terms of physical structure, while also having good spatial isolation and electromagnetic compatibility, thus improving the overall detection sensitivity and stability of the device.

[0047] The first detection assembly 2 includes a first thermoluminescent detector 21, a first neutron absorber 22, and a second thermoluminescent detector 23, used to achieve differentiated cumulative dose detection of gamma rays and neutrons. The first thermoluminescent detector 21 includes a pair of thermoluminescent elements arranged in parallel, namely... 6 Li detector elements and 7 The Li detector element consists of two thermoluminescent elements, both cylindrical, cuboid, or disc-shaped, to achieve dose decoupling and quantitative differentiation of neutrons and photons. These two thermoluminescent elements are closely fitted or arranged adjacently and encapsulated within a first neutron absorber 22, serving as a single unit to respond to gamma rays. 7 The Li detector plate does not react with neutrons at all, but only responds to gamma rays; 6 The Li detector sheet can respond to both gamma rays and thermal neutrons via the 6Li(n,α)T reaction. Therefore, the contribution of neutrons to the thermoluminescence signal can be quantitatively distinguished by the difference between the two readings.

[0048] The second thermoluminescent detector 23 also consists of a pair of thermoluminescent elements, with a structure identical to that of the first thermoluminescent detector 21, i.e., it includes one... 6 Li detector element and a 7 Li detection element. A first thermoluminescent detector 21 is positioned below the second thermoluminescent detector 23 and is in direct contact with the first groove 11. The second thermoluminescent detector 23 is also encapsulated within a first neutron absorber 22, used for direct response to the radiation source term, thereby obtaining a thermoluminescent signal primarily originating from the interaction of neutrons in the radiation source term. By combining the readings with those of the first thermoluminescent detector 21, the separability and response accuracy of the dose component under the γ / neutron recombination field can be enhanced.

[0049] like Figure 1As shown, the first neutron absorber 22 has a flat, box-like structure with a closed frame on all four sides. The central area is a rectangular platform, with a fourth groove and a fifth groove (not shown) on its upper and lower surfaces, respectively, for embedding the second thermoluminescent detector 23 and the first thermoluminescent detector 21. Preferably, the first neutron absorber 22 is made of cadmium, boron, or other elements, compounds, or mixtures thereof with good neutron absorption properties, to effectively shield against non-target radiation. The cross-sectional shapes of the fourth and fifth grooves match the corresponding cross-sections of the second and first thermoluminescent detectors 23 and 21, thus preventing detector damage from shaking. Furthermore, the overall dimensions of the first neutron absorber 22 match the first groove 11, allowing for a tight fit during assembly and ensuring component stability and structural consistency. The first groove 11 is a dedicated mounting slot for fixing the first detection component 2, featuring a stepped positioning platform, side wall slots, or bottom limiting posts, matching the outer frame structure of the first neutron absorber 22 for fixing the first detection component 2. To prevent the thermoluminescent detector from detaching during use, a flexible pad, micro-spring, or adhesive layer can be provided inside the first groove 11 to improve its impact resistance.

[0050] Similar in structure to the first detection component 2, the second detection component 3 includes a third thermoluminescent detector 31, a second neutron absorber 32, and a fourth thermoluminescent detector 33. It is primarily used for ray and energy discrimination, dose compensation, redundancy verification, or directional response extension in a mixed neutron and gamma-ray radiation field. Both the third thermoluminescent detector 31 and the fourth thermoluminescent detector 33 consist of a pair of complementary thermoluminescent elements, namely, elements rich in... 6 Li thermoluminescent element and a sheet rich in ... 7 Li thermoluminescent element. This structural design allows the detector to respond to gamma rays while simultaneously comparing... 6 Li and 7 The difference in readings on the Li sheet accurately reflects the dose component of thermal neutrons. The third thermoluminescent detector 31 is positioned below the fourth thermoluminescent detector 33 and is in direct contact with the second groove 12. The second neutron absorber 32 is also a flat, box-shaped structure with a closed perimeter and a rectangular platform in the center. Its upper and lower surfaces are respectively provided with a sixth and a seventh groove (not shown in the figure) for embedding the fourth thermoluminescent detector 33 and the third thermoluminescent detector 31. Preferably, the second neutron absorber 32 is made of cadmium, boron, or other elements, compounds, or mixtures thereof with good neutron absorption properties to achieve effective shielding against non-target radiation. The cross-sectional shapes of the sixth and seventh grooves match the corresponding cross-sections of the fourth and third thermoluminescent detectors 33 and 31, improving impact resistance. Furthermore, the overall dimensions of the second neutron absorber 32 match the second groove 12, allowing for a tight fit during assembly and ensuring component stability and structural consistency.

[0051] The second detection component 3 is arranged parallel to the first detection component 2. Their structures are coordinated in terms of dimensions, thickness, and installation direction, forming a symmetrical arrangement within the dosimeter base 1. This facilitates the overall structural balance and dose response uniformity design. This parallel layout not only improves the spatial consistency of the detection data but also provides technical support for subsequent data analysis, including dual-channel comparison, outlier detection, and directional dose correction.

[0052] The second detection component 3 is installed in the pre-set second groove 12 in the dosimeter base 1. The groove structure is precisely designed according to the shape of the component, and has a stepped positioning platform, side wall slots or bottom limiting posts, which match the outer frame structure of the second neutron absorber 32 to fix the second detection component 3. To prevent the structure from detaching during use, a flexible pad, micro springs or adhesive layer can be set inside the second groove 12, which also improves the impact resistance.

[0053] The third detection component 4 is parallel and uniformly distributed with the first detection component 2 and the second detection component 3. It includes a solid-state track detector 41 and a neutron moderator 42, primarily used for indirect measurement and quantitative recording of fast neutron dose. It is suitable for personal dose assessment in high-energy neutron fields or environments dominated by fast neutrons. The solid-state track detector 41 is a polymer detector that can generate permanent microscopic damage tracks through high-energy particle impacts, preferably made of polycarbonate (such as Makrofol), CR-39 (polyalkylene dicarboxylate), or CAP (Cellulose Acetate Polymer). The neutron moderator 42 is used to slow down incident high-energy fast neutrons to the thermal neutron or slow neutron energy range, thereby increasing the possibility of indirect interaction with the solid-state track detector. Preferred materials are hydrogen-rich polymers, such as high-density polyethylene (HDPE), polyoxymethylene, paraffin wax, or hydrogenated silicone rubber, which possess good neutron moderation capabilities, mechanical strength, and dimensional stability.

[0054] The solid track detector 41 is divided into two functional regions, or is composed of two independent solid track detection elements spliced ​​together, referred to as the first track region and the second track region, respectively. The materials, sizes, thicknesses, and detection sensitivities of the two regions are basically the same to ensure comparability in fast neutron response capabilities and mutual reference of results.

[0055] The neutron moderator 42 is positioned above the solid-state track detector 41 and covers one of its regions, such as... Figure 1 As shown, the first track region is preferably covered. The shape and size of the neutron moderator 42 are designed to be basically matched with the covered region, ensuring that the moderated neutron beam can fully act on the track detector below the region, thereby forming a recordable track through the indirect reaction between the moderated neutrons and the detection material (such as secondary protons generated by elastic scattering).

[0056] The other uncovered region (i.e., the second track region) is directly exposed to the neutron field in the environment and responds only to unmoderated high-energy fast neutrons. By comparing and analyzing the track densities recorded in the two regions, the distribution information of fast neutrons at different energy spectra can be derived, or it can be used to eliminate the offset of the overall response caused by the moderator material, thereby achieving selective identification of fast neutron spectral bands and energy response correction.

[0057] To achieve the measurement and differentiation of dose equivalents at different depths, in this embodiment, a measurement window 6 is provided on the dosimeter cover 5 to provide a controlled incident path for a specific detector, simulating the radiation dose response characteristics at different depths of human tissue. The measurement window 6 is an opening structure with a defined area, and its material is the same as or similar to that of the cover body. Preferably, a tissue-equivalent polymer material is used, such as polycarbonate (PC), polymethyl methacrylate (PMMA), or other calibrated and certified equivalent materials, to maintain the consistency of particle energy deposition characteristics.

[0058] The dosimeter cover 5 has an overall thickness set to 1000 mg / cm. 2 The tissue equivalent material, i.e., the mass thickness is equivalent to 1.0 g / cm. 2 The thickness is designed according to the definition of personal dose equivalent Hp(10) by relevant international standards organizations, representing a dose measurement reference point of 10 mm tissue depth. It is suitable for assessing the dose of deep organs (such as bone marrow, internal organs, etc.) to γ, n rays or high-energy particle radiation, and is used for simultaneous measurement of subcutaneous deep tissue and eye lens dose.

[0059] Measuring window 6 has a thickness set to 300 mg / cm. 2 The tissue equivalent material, i.e., a mass thickness of 0.3 g / cm. 2 It is equivalent to the human tissue dose depth at 3 mm below the skin, which conforms to the international standard of personal dose equivalent Hp(3). It is used to assess the radiation dose received by superficial tissues (such as the lens of the eye), and is especially suitable for dose monitoring of sensitive areas of the lens of the eye. It is used for simultaneous measurement of dose in deep subcutaneous tissues and the lens of the eye.

[0060] Through the above structural design, and the parallel distribution of the three types of detection components, independent responses and dose assessments for gamma rays, thermal neutrons, and fast neutrons are achieved, respectively. In the first detection component 2 and the second detection component 3, neutron absorbers (first neutron absorber 22 and second neutron absorber 32) are respectively embedded in the upper and lower double-groove neutron absorbers. 6 Li and 7The Li thermoluminescent detector utilizes the difference in response to thermal neutrons and gamma rays to distinguish between neutron and photon doses. The third detection component 4 employs a combination of a neutron moderator 42 and a solid-state track detector 41 to indirectly measure fast neutron doses, expanding the adaptability of traditional dosimeters to high-energy neutron fields. The overall structure of the dosimeter is constructed from tissue-equivalent materials, including the dosimeter cover 5, which has a measurement window 6. The mass and thickness of the dosimeter cover 5 and the measurement window 6 are both 1000 mg / cm². 2 With 300mg / cm 2 These correspond to the dose measurement depths of Hp(10) and Hp(3), respectively, ensuring consistency between the detection response and the actual biological dose in the human body. The components are designed with multi-level sealing, limiting, and coordination to enhance structural stability and environmental adaptability, making them suitable for accurate and reliable dose monitoring in complex radiation environments.

[0061] In some preferred embodiments, the dosimeter cover 5 is affixed with a label to identify unique information associated with the dosimeter, such as a serial number, letters, numbers, or other identifiable identifiers. This label enables individualized management, tracking, and data binding of the dosimeter, and is particularly suitable for use scenarios involving multiple users, periodic replacement, or batch management. More preferably, the label can be located on the outer surface or inner edge of the dosimeter cover 5, without obstructing the radiation incident path and without affecting measurement accuracy.

[0062] In some preferred embodiments, the opening area of ​​the measuring window 6 is circular or rectangular, and the area can be optimized according to the size of the detection element (e.g., diameter 6–10 mm).

[0063] In some preferred embodiments, the personal dosimeter of this embodiment is placed on the chest surface, either on the left or right chest. Preferably, it can be placed in the chest pocket of work clothes or held in place by a chest strap.

[0064] When the personal dosimeter of this embodiment is exposed to a mixed photon-neutron radiation field, its overall system works collaboratively through three types of detection components to achieve the distinguishing detection of gamma-ray and neutron radiation and accurate assessment of personal dose. This embodiment also provides a method for detecting mixed neutron and photon radiation based on the above-described dosimeter, including:

[0065] Step 1: A solid-state track detector 41 is used to record track events caused by fast neutrons, and the readings of the solid-state track detector are recorded. Its surface is divided into two regions, one part of which is covered by a neutron moderator 42, and the other part is uncovered. Because the neutron moderator 42 has a slowing effect on fast neutrons, the neutron energy spectrum received in the covered region changes, resulting in a difference in the number of tracks recorded in this region and the uncovered region. By comparing the track densities of the two regions, a response model is established, and the neutron response correction coefficient S1 for different energies is obtained through fitting.

[0066] Step 2: In the first detection component 2, the second thermoluminescent detector 23 is directly exposed to the initial radiation field, recording the superposition response of γ and thermal neutrons. The first thermoluminescent detector 21, due to the presence of the first neutron absorber 22, has its neutron response partially shielded in the initial radiation field, primarily recording the responses of γ and neutrons scattered by the human body. The correction coefficient S2 is obtained by fitting the reading difference between the first thermoluminescent detector 21 and the second thermoluminescent detector 23.

[0067] Step 3: Similar to the structure and function of the second detection component, the different readings of the fourth thermoluminescent detector 33 and the third thermoluminescent detector 31 are fitted to obtain the correction coefficient S3;

[0068] Step 4: Calculate the individual dose equivalents Hp(10) and Hp(3):

[0069] Based on the readings of the solid track detector 41, the first thermoluminescent detector 21, and the second thermoluminescent detector 23, multi-channel fitting was performed using correction coefficients S1, S2, and S3 to obtain the following results:

[0070] The deep tissue dose Hp(10) caused by gamma rays and the deep tissue dose Hp(10) caused by neutrons (fast neutrons + thermal neutrons);

[0071] Thus, the total deep personal dose equivalent is obtained: Hp(10) = Hp(10)light + Hp(10)middle;

[0072] Based on the readings from the solid-state track detector 41, the third thermoluminescent detector 31, and the fourth thermoluminescent detector 33, multi-channel fitting was performed using correction coefficients S1, S2, and S3 to obtain the following results:

[0073] The superficial tissue dose (e.g., the lens of the eye) caused by gamma rays Hp(3) and the superficial dose Hp(3) caused by light and neutrons;

[0074] Thus, the total superficial personal dose equivalent is obtained: Hp(3) = Hp(3)light + Hp(3)middle;

[0075] The above results enable the separate measurement of gamma rays and neutron radiation in mixed radiation fields, simultaneous assessment of deep and shallow tissue doses, and correction of fast / thermal neutron spectrum response, significantly improving the accuracy and reliability of the measurement results. It is particularly suitable for personal radiation protection monitoring in various scenarios such as nuclear medicine, nuclear industry, and irradiation operations.

[0076] The embodiments of the present invention have been described above with reference to the accompanying drawings. However, the present invention is not limited to the specific embodiments described above. The specific embodiments described above are merely illustrative and not restrictive. Those skilled in the art can make many other forms under the guidance of the present invention without departing from the spirit and scope of the claims, and all of these forms are within the protection scope of the present invention.

Claims

1. A personal dosimeter for neutron and photon mixed radiation detection, characterized in that, The dose meter comprises a dose meter base, a first detection assembly, a second detection assembly, a third detection assembly and a dose meter cover, the dose meter base and the dose meter cover are coupled to form a closed cavity, the dose meter base is provided with first, second and third grooves arranged in parallel to accommodate the three detection assemblies respectively; The first detection component is arranged adjacent to one side of the second detection component, the first detection component comprises a first neutron absorber enclosed by four peripheral frames, the center of the first neutron absorber is a rectangular platform, and grooves are arranged on the upper and lower parts of the platform respectively for encapsulating a second thermoluminescence detector and a first thermoluminescence detector distributed on the upper and lower parts, and the first thermoluminescence detector is in contact with the first groove; the second detection component comprises a second neutron absorber enclosed by four peripheral frames, the center of the second neutron absorber is also a rectangular platform, and grooves are arranged on the upper and lower parts of the platform respectively for encapsulating a fourth thermoluminescence detector and a third thermoluminescence detector distributed on the upper and lower parts, and the third thermoluminescence detector is in contact with the second groove; the four thermoluminescence detectors all adopt 6 The pair design of Li and 7 Li detection elements realizes dose decoupling and quantitative differentiation of neutrons and photons. The third detection assembly is arranged adjacent to the other side of the second detection assembly and comprises a solid track detector and a neutron moderator, the neutron moderator is arranged above the solid track detector and covers part of the area of the solid track detector; The dose meter cover is provided with a measurement window, the dose meter cover and the measurement window are made of tissue-equivalent materials with different mass thicknesses respectively to correspond to different standard dose depths.

2. The personal dosimeter of claim 1, wherein, The dosimeter upper cover and the measurement window have a mass thickness of 1000 mg / cm 2 and 300 mg / cm 2 of tissue equivalent material for subcutaneous deep tissue and eye lens dose simultaneous measurement.

3. The personal dosimeter of claim 1, wherein, The 6 Li and 7 The Li detection elements are cylindrical or cuboid or round, closely fitted or arranged adjacent.

4. The personal dosimeter of claim 1, wherein, The neutron moderator is made of a hydrogen-rich polymer.

5. The personal dosimeter of claim 4, wherein, The neutron moderator is made of one of high-density polyethylene (HDPE), polyformaldehyde, paraffin wax or hydrogenated silicone rubber.

6. The personal dosimeter of claim 1, wherein, The dose meter cover is attached with a label, which is arranged on the outer surface or the inner edge of the dose meter cover.

7. The personal dosimeter of claim 1, wherein, The first and second neutron absorbers are made of single elements, compounds and mixtures of neutron absorbing materials such as cadmium and boron.

8. A method for detecting mixed neutron and photon radiation based on the personal dosimeter according to any one of claims 1 to 7, characterized in that, The method comprises: Recording the readings of the solid track detector, establishing a response model by comparing the track densities of the two areas covered and not covered by the neutron moderator on the solid track detector, and fitting to obtain a correction coefficient S1 of the response of neutrons with different energies; Recording the readings of the second and first thermoluminescent detectors and fitting according to the difference to obtain a correction coefficient S2; Recording the readings of the fourth and third thermoluminescent detectors and fitting according to the difference to obtain a correction coefficient S3; Based on the readings and correction coefficients, calculating the personal dose equivalent Hp(10) and Hp(3).

9. The detection method according to claim 8, characterized in that, When calculating the personal dose equivalent Hp(10): Based on the readings of the solid track detector, the first and second thermoluminescent detectors, combining the correction coefficients S1, S2 and S3 for multi-channel fitting, obtaining the deep tissue dose Hp(10) caused by gamma rays and the deep tissue dose Hp(10) caused by neutrons, and calculating Hp(10) according to the following formula, Hp(10) = Hp(10) light + Hp(10) medium.

10. The detection method of claim 8, wherein, When calculating the personal dose equivalent Hp(3): Based on the readings of the solid track detector, the first and second detection assemblies, combining the correction coefficients S1, S2 and S3 for multi-channel fitting, obtaining the shallow tissue dose Hp(3) caused by gamma rays and the shallow dose Hp(3) caused by neutrons, and calculating Hp(3) according to the following formula, Hp(3) = Hp(3) light + Hp(3) medium.

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