CN108445529B - Active neutron personal dosimeter based on three-layer silicon detector and measuring method thereof - Google Patents
Active neutron personal dosimeter based on three-layer silicon detector and measuring method thereof Download PDFInfo
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Abstract
The invention relates to an active neutron personal dosimeter based on a three-layer silicon detector and a measuring method thereof, the personal dosimeter comprises a main detector, an outer layer transformant component and a nuclear electronics system, the main detector and the outer layer transformant component comprise three passivation injection plane silicon detectors, an Open detector, a Fast detector and an Albedo detector are arranged in sequence from top to bottom, and a polyethylene layer and a nuclear electronics system are arranged in front of the Open detector 6 A LiF coating, a polyethylene layer arranged in front of the Fast detector, a boron-containing polyethylene layer, a polyethylene layer and a boron-containing polyethylene layer arranged in front of the Albedo detector 6 A LiF coating; the nuclear electronics system provides high voltage to each detector and acquires detection signals for multi-channel analysis. The invention provides two measurement modes, wherein neutron fluence and individual dose equivalent information can be given in real time in a direct reading mode, and neutron field energy spectrum can be given more accurately in a spectrum resolving mode. The invention has wide energy measuring range and energy resolving power, and can be used in n-gamma mixed fields.
Description
Technical Field
The invention relates to a neutron personal dosimeter technology, in particular to an active neutron personal dosimeter based on a three-layer silicon detector and a measuring method thereof.
Background
The neutron personal dose meter that is currently more commonly used is a CR-39 solid plastic track detector. After the plastic track detector is irradiated by neutrons, C, H atoms, O atoms and the like in the plastic track detector collide with neutrons to generate backflushing nuclei, and when the energy of the backflushing nuclei is large enough, damage potential tracks can be generated in the material. Then, chemical etching or electrochemical etching is used to etch easily observable tracks in the detector material and observe under a microscope. The solid track personal dosage meter has the advantages of small volume, no need of electronic instrument during measurement, etc. But its shortcomings are also quite evident: 1. the lower measurement limit is high, about 100keV, due to the basic principle of such a detector to measure the recoil kernel, and the nature of the material itself, so if such a personal dosimeter is used to measure a broad or energy spectrum softer neutron field, the dose of that neutron field will be underestimated; 2. the personal dosimeter can only measure the number of neutrons, but cannot learn the energy of neutrons, and if the difference between the use place and the energy spectrum of a neutron source used for calibration is large, the reading is inevitably higher or lower; 3. the passive neutron personal dosimeter of the personal dose meter needs to be worn by staff for a period of time and then reads, and the dose cannot be displayed in real time, so that if the dose limit value is exceeded, the passive neutron personal dosimeter cannot be used for alarming, and can only be used for conventional monitoring and dose evaluation under a non-accident state; 4. when the plastic track detector is read, a microscope and other instruments are needed to be used, and the number of tracks in the detector is counted, so that the time and the labor are consumed.
References to CR-39 neutron personal dosimeters are as follows:
cao Lei, deng Jun, zhang Guiying, et al, CR39 neutron personal dosimeter performance experimental studies, radiation protection [ J ],2012,32 (3): 103-107.
Feng Yushui, li Jun, lin Zhikai, et al, CR-39 plastic recoil track personal neutron dosimeter nuclear technology [ J ],1988,11 (9): 44-46.
Since the 60 s of the 20 th century, semiconductor detectors have begun to be used more widely for neutron detection. The seed personal dose timer is manufactured by using a semiconductor detector, a neutron converter film or coating is usually added outside the detector, secondary charged particles are generated by nuclear reaction of neutrons and certain specific nuclides, and then the relevant information of the neutrons is obtained. These nuclear reactions are mainly 6 Li(n,α)T， 10 B(n,α) 7 Li，He(n,p)T， 155,157 Gd(n,γ) 156,158 Gd, and the like. A more typical semiconductor neutron personal dosimeter is DOS-2002 manufactured by German PTB. The personal dosimeter is mainly based on a single-layer silicon detector, and polyethylene are arranged on the outer layer of the detector 6 LiF, as a fast neutron and thermal neutron transformant, by recording neutron counts to infer personal dose equivalents. DOS-2002 has some neutron and gamma-ray resolving power. The disadvantages of DOS-2002 are mainly that: neutron energy cannot be resolved. The nature of such personal dosimeters is to record the number of thermal and fast neutrons, while their personal dose equivalent response varies greatly with energy, the minimum and maximum values may differ by several orders of magnitude. Thus, it is clear that recording only the number does not give a more accurate personal dose equivalent.
References to semiconductor neutron personal dosimeters are as follows:
Luszik-Bhadra.Electronic personal doesmeters:The solution to problems of individual monitoring in mixed neutron fields？.Radiation Protection Dosimetry[J],2004,110(1-4):747-752.
M.Luszuk-Bhadra,W.Wendt and M.Weierganz.The Electronic Neutron/Photon Doesmeter PTB DOC-2002.Radiation Protection Dosimetry[J],2004,110(1-4):291-295.
disclosure of Invention
The invention aims to overcome the defects of the prior art, and provides an active neutron personal dosimeter which has wide energy measuring range and energy resolution capability, can be used in an n-gamma mixed field and provides two measuring modes of the neutron personal dosimeter.
The technical scheme of the invention is as follows: an active neutron personal dosimeter based on three layers of silicon detectors comprises a main detector, an outer layer transformant component and a nuclear electronics system, wherein the main detector and the outer layer transformant component comprise three passivation injection plane silicon detectors, an Open detector, a Fast detector and an Albedo detector are sequentially arranged from top to bottom, and a polyethylene layer and a nuclear electronics layer are arranged in front of the Open detector 6 A LiF coating, a polyethylene layer arranged in front of the Fast detector, a boron-containing polyethylene layer, a polyethylene layer and a boron-containing polyethylene layer arranged in front of the Albedo detector 6 A LiF coating; the nuclear electronics system provides high voltage to each detector and acquires detection signals for multi-channel analysis.
Further, the active neutron personal dosimeter based on the three-layer silicon detector comprises a preamplifier, a main amplifier, a low-voltage power supply, a high-voltage power supply and a multi-channel analyzer, wherein detection signals of the three passivated injection plane silicon detectors are sent into the multi-channel analyzer through the preamplifier and the main amplifier.
Further, the active neutron personal dosimeter based on the three-layer silicon detector is described above, wherein the three passivation injection plane silicon detectors are encapsulated in an aluminum housing.
The measuring method of the active neutron personal dosimeter adopts a direct reading mode, and the method firstly divides neutron energy into three parts: the method comprises the steps of obtaining the fluence responses of each detector of a personal dosimeter under different energies through a single-energy neutron fluence scale experiment in a 20 meV-1 keV slow neutron energy region, a 1 keV-1 meV medium energy neutron energy region and a 1 meV-20 meV fast neutron energy region, calculating the average fluence responses of different energy regions, and obtaining the personal dose equivalent response by using fluence-personal dose equivalent conversion coefficients; the count of the Fast detector is divided by the personal dose equivalent of the Fast neutron contribution, the difference between the count of the Open detector and the count of the Fast detector is divided by the personal dose equivalent of the Open detector of the low energy neutron contribution, the difference between the count of the Albedo detector and the count of the Open detector is divided by the medium energy neutron response, the personal dose equivalent of the medium energy neutron contribution, and the sum of the three is the total neutron personal dose equivalent.
In another measurement method of the active neutron personal dosimeter, a spectrum decomposition mode is adopted, and the method utilizes the relationship between the energy and angle of recoil protons and neutron energy as follows:
E p ＝E n cos 2 φ
wherein E is p For recoil proton energy, E n For the energy of an incident neutron, phi is the recoil angle,
deconvolving the pulse amplitude spectrum measured by the Fast detector into neutron energy spectrum.
The beneficial effects of the invention are as follows: 1) The neutron energy measuring range of the personal dosimeter of the invention covers the common neutron field energy range of radiation protection, and the heat energy is 20MeV; 2) The invention has certain energy resolving power and can respectively give low-energy neutrons, medium-energy neutrons and fast neutronsPersonal dose equivalent of son; 3) The invention has lower detection lower limit of 9.86×10 -1 Mu Sv; 4) The invention has the gamma-ray screening capability and can be used in an n-gamma mixed field; 5) The invention provides two measurement modes, wherein neutron fluence and individual dose equivalent information can be given in real time in a direct reading mode, and neutron field energy spectrum can be given more accurately in a spectrum resolving mode.
Drawings
FIG. 1 is a schematic diagram of the structure of a main detector and its outer transformant components according to the present invention;
FIG. 2 is a schematic diagram of the structure of a personal dosimeter core electronics system;
FIG. 3 is a graphical representation of individual dose equivalent response results;
FIG. 4 is a graph showing the results of a personal dosimeter 7MeV photon response calculation;
FIG. 5 is a schematic representation of neutron spectrum results in the spectral resolution mode.
Detailed Description
The present invention will be described in detail with reference to the accompanying drawings and examples.
The invention provides a neutron personal dosimeter based on a three-layer silicon detector, which mainly comprises the following parts: a main detector, an outer transformant component thereof, and a nuclear electronics system. The invention also provides two measurement modes of the neutron personal dosimeter, including a direct reading method and spectrum resolution software.
The neutron personal dosimeter detector body is 20mm in diameter and 20mm in height, and is enclosed in a 20mm by 40mm by 60mm aluminum housing. The assembly mainly comprises three passivation injection plane silicon detectors, polyethylene, boron-containing polyethylene, and, 6 LiF and other materials are used as a moderator, an absorber and a transformant to form a sandwich structure. The silicon detector is named an Open detector, a Fast detector and an Albedo detector from top to bottom in sequence.
As shown in FIG. 1, only a polyethylene layer 4 is arranged in front of the Fast detector 2, and the detector can only record recoil protons generated by the elastic scattering of Fast neutrons and hydrogen nuclei. The Fast detector 2 responds only to neutrons with energies greater than 1 MeV.
Open probes 1 and AThe lbedo detector 3 is provided with not only a polyethylene layer 4 but also in front of it 6 LiF coatings 5, so that they are able to record low-energy neutrons and 6 alpha particles generated by Li reaction and tritium; and simultaneously, recoil protons generated by fast neutrons can be recorded. In addition to the surrounding components, the Albedo detector 3 is also provided with a boron-containing polyethylene layer 6 in front of and closer to the human body or body membrane during operation, so that the low-energy neutron response is slightly lower than that of the Open detector, and the medium-energy neutron response is slightly higher than that of the Open detector.
As a specific example, the detector is a PIPS detector of the MSD011 type manufactured by MICRON Inc. of England, whose sensitive region has a diameter of 10mm and a maximum depletion layer thickness of 300. Mu.m. In practice, the detector high voltage is set to 3.5V, at which time the sensitive area thickness is about 100 μm. At this sensitive thickness, the silicon detector is able to fully deposit neutrons and 6 the Li reacts to produce alpha particles and tritium particles, for which the detector is a fully deposited detector. But the range of recoil protons in silicon is long, so that only recoil protons with an energy of 3.2MeV can be completely deposited. When the energy of the recoil protons is higher, the detector is a penetration type detector, and the particle energy can be judged only by the energy deposition spectrum shape difference generated by the recoil protons with different energies. The Open detector responds most to thermal neutrons, the Albedo detector responds most to energetic neutrons, and the Fast detector responds only to Fast neutrons.
In the outer transformant assembly, the polyethylene and the boron-containing polyethylene were discs of thickness 2mm and diameter 20 mm. The boron-containing polyethylene is B 4 C。 6 The LiF film plating adopts ion sputtering film plating technology, and the technology utilizes positive ions generated by gas discharge to bombard a target material at high speed under the action of an electric field so that atoms in the target material escape and are deposited on the surface of polyethylene to form 6 LiF film plating. 6 LiF film thickness 4 μm, mass thickness 1mg/cm 2 . Because the scattering cross section of aluminum to neutrons is very small, and the strength and the cost are considered, the shell is made of aluminum alloy, and the thickness is 1mm.
As shown in fig. 2, the nuclear power system includes a preamplifier, a main amplifier, a low voltage power supply, a high voltage power supply, and a multi-channel analyzer and software thereof. The system mainly provides high voltage for the detector, the output voltage of the high-voltage power supply is 3.5V, and the system is kept to work normally. The pulse signal can be amplified, the signal to noise ratio can be improved, and analog-to-digital conversion, multichannel analysis and the like can be performed.
The personal dosimeter of the invention has two modes in total: a direct reading mode and a spectrum resolving mode. The direct reading mode can give neutron fluence or personal dose equivalent of three energy areas in real time, and is a relatively rough but rapid method; the spectrum deconvolution mode deconvolutes the pulse amplitude spectrum measured by the Fast detector into a neutron energy spectrum by using spectrum deconvolution software, and the mode result is more accurate but cannot be calculated in real time.
1) Direct reading mode
The basis of the direct reading mode is the difference in the responses of the three detectors to neutrons of different energies.
The neutron personal dosimeter divides the neutron energy into three parts: a slow neutron energy region (thermal) of 20 meV-1 keV; an energy neutron energy region (interval) between 1keV and 1 MeV; 1 MeV-20 MeV fast neutron energy region (fast). Therefore, the neutron fluence phi in the total energy area is the sum of the fluence of the three:
φ＝φ thermal +φ interval +φ fast (1)
for each detector, the total count M is the sum of the neutron counts in three energy regions:
M＝M thermal +M interval +M fast (2)
in addition, the detector count is again equal to the neutron fluence phi in a unit energy interval, as defined by fluence response E (E) Integration within the total region of the product of the fluence response R (E) with the corresponding energy interval:
M＝∫φ E (E)R(E)dE (3)
discretizing equation (3), the detector count can be approximated as the sum of the count of the three energy regions multiplied by the corresponding average fluence response:
M＝φ thermal R thermal +φ interval R interval +φ fast R fast (4)
since the transformants in front of the three detectors are different, their responses are slightly different for neutrons of different energies. Applying equation (4) to each detector can result in:
M O ＝φ thermal R O,thermal +φ interval R O,interval +φ fast R O,fast
M F ＝φ fast R F,fast
M A ＝φ thermal R A,thermal +φ interval R A,interval +φ fast R A,fast (5)
wherein R is thermal 、R interval And R is fast The average fluence responses of the detector to low energy neutrons, medium energy neutrons and fast neutrons are respectively represented and can be obtained through Monte Carlo calculation or single energy fluence response scale experiments. Considering the characteristic that the active personal dosimeter needs to display parameters such as personal dose equivalent in real time, the calculation of equation (5) is too complex, difficult to miniaturize, and not suitable for the personal dosimeter. Thus, equation (5) can be reduced to:
M O ＝R O,thermal ×φ thermal ×C 1
M F ＝R F,fast ×φ fast ×C 2
M A ＝R A,interval ×φ interval ×C 3 (6)
in equation (6), C represents the calibration coefficient for each detector. The fluence response can be converted to a personal dose equivalent response using the neutron fluence-personal dose equivalent conversion coefficient given in ICRP74 report.
And calculating the average personal dose equivalent response of each detector to neutrons in different energy regions according to the design and experimental results. When in measurement, the neutron counts of the three detectors are divided by the corresponding individual dose equivalent responses and their calibration factors to obtain the neutron fluence or individual dose equivalent of the three energy regions.
As a specific embodiment, firstly, the fluence responses of the detectors of the personal dosimeter under different energies are obtained through a monoenergetic neutron fluence scale experiment. The experimental energy points included 144keV, 250keV, 565keV, 1.2MeV, 2.5MeV, 5MeV, and 14.8MeV. And selecting 1MeV as a gamma ray threshold according to the Monte Carlo calculation result. After subtraction of the corresponding counts, background contributions of the chamber scatter were subtracted using either cone of view or extrapolation. Thus, the average fluence responses of the different energy areas can be calculated, and then the fluence-personal dose equivalent conversion coefficient is used for obtaining the personal dose equivalent response. According to the experimental result, the count of the Fast detector can be considered to be divided by the response thereof, namely, the personal dose equivalent which is the Fast neutron contribution; the difference in the counts of the Open detector and the Fast detector divided by the personal dose equivalent for which the Open detector response to low energy neutrons is a low energy neutron contribution; and the difference between the counts of the Albedo detector and the Open detector divided by the neutron response is the personal dose equivalent of the neutron contribution. The sum of the three is the total neutron personal dose equivalent. The neutron personal dose equivalent can be obtained by using the formula (7) and the counts of the three detectors. Wherein H is p (10) Representing individual dose equivalents in each energy region; m is M F Representing Fast detector counts; m is M A Representing Albedo detector counts; m is M O Representing an Open detector count;representing the average personal dose equivalent response of the fast neutron energy region; />Represents the average personal dose equivalent response of the medium energy neutron energy region; />Indicating the average personal dose equivalent response in low energy neutrons.
2) De-spectral mode
Because the detector is provided with a layer of polyethylene sheet before Fast, when neutron energy is higher than 1MeV, the neutron energy can elastically scatter with hydrogen nuclei in the detector to eject recoil protons. The energy and angle of recoil protons are related to neutron energy as follows:
E p ＝E n cos 2 φ (8)
wherein E is p For recoil proton energy, E n For the energy of an incident neutron, φ is the recoil angle. According to the relation, the energy information of the incident neutrons can be reversely deduced through spectrum resolution software by using a deconvolution method.
As a specific example, the response functions of three silicon detectors are first determined by comparing experimental measurements with theoretical calculations. Using the GEANT4 Monte Carlo program, simulating and calculating the response functions of 39 energy points in the energy range of 20 meV-15 meV of the detector Open and the detector Albedo, and calculating 30 response functions in the energy range of 1 meV-15 meV of the detector Fast. And comparing the theoretical calculation result with the response result of the monoenergetic neutron response scale experiment, and correcting the theoretical calculation result. And carrying out multi-channel spectrum decomposition calculation by adopting a genetic algorithm, and carrying out processes of determining a solution space, determining an adaptability function, determining a genetic operator and the like to deconvolve the measured pulse amplitude spectrum into a neutron energy spectrum.
The design-related parameters of the spectral analysis calculation are shown in table 1.
Table 1 relevant parameters involved in the calculation of the solution spectrum
The neutron personal dosimeter provided by the invention adopts the passivation injection plane silicon detector, and the silicon detector has the characteristics of small volume, light weight, high sensitivity and the like. The volume of one probe and its external transformant was only 20mm by 5mm. The overall external dimension of the personal dosimeter is 20mm multiplied by 40mm multiplied by 60mm, the mass is only 55g, and the personal dosimeter is suitable for daily wearing of individuals.
A monoenergetic neutron fluence scale experiment is carried out by utilizing a quasi monoenergetic neutron reference radiation field established on a 5SDH-2 serial accelerator of the metering and testing part of the national institute of atomic energy science. Experimental energy points included 144keV, 250keV, 565keV, 1.2MeV, 2.5MeV, 5MeV, and 14.8MeV, each of which was tested at two different distances to subtract the chamber scatter background. The results of the individual dose equivalent response experiments are shown in figure 3. The individual dose equivalent response in fig. 3 varies with energy, covering the thermal energy to 20MeV, for radiation protection in most locations.
The average individual dose equivalent response of the individual dosimeter to neutrons of different energies can also be obtained by single energy neutron response scale experiments and Monte Carlo simulations, as shown in Table 2.
TABLE 2 average personal dose equivalent response (s/. Mu.Sv) for different energy regions of a personal dosimeter
According to the principle formulas in the table 2 and the technical proposal, the detection lower limit of the personal dosimeter is 9.86 multiplied by 10 -1 Mu Sv is far below the existing passive personal dosimeter detection lower limit.
The personal dosimeter deducts the influence caused by gamma rays through a card domain mode, and can more accurately measure the personal dose equivalent caused by the neutron.
Fig. 4 is a 7MeV photon energy deposition spectrum calculated using the GEANT4 monte carlo program. It can be seen from the figure that photons with an energy of 7MeV deposit an energy of at most 1MeV in a silicon detector with a sensitive area thickness of 100 μm. Thus, the present neutron personal dosimeter is able to subtract up to 7MeV photons, as long as the photon threshold is set to 1 MeV.
Placing the personal dosimeter in 241 The test experiments were performed with an Am-Be neutron source. The results of the direct reading mode are shown in table 3, and the results of the spectrum resolution mode are shown in fig. 5.
Table 3. 241 Am-Be neutron source personal dose equivalent experimental results (mu Sv/h)
241 The true value of the personal dose equivalent convention at 101cm of Am-Be neutron source is (3.24E+02) mu Sv/h, the experimental relative error is 52.6%, the true value of the convention at 67cm is (3.53E+02) mu Sv/h, and the experimental relative error is 52.1%.
In FIG. 5, the two lines are each given in ISO8529 241 Am-Be neutron source standard spectrum and spectrum resolution results. As can be seen from the figure, the spectrum resolution result is better matched with the standard spectrum. The personal dose equivalent obtained by the resolution spectrum at 101cm is (1.90E+02) mu Sv/h, the agreed true value is (3.53E+02) mu Sv/h, and the experimental relative error is 41.5%; the individual dose equivalent obtained by the resolution at 67cm is (5.51E+02) mu Sv/h, the agreed true value is (7.36E+02) mu Sv/h, and the experimental relative error is 25.1%.
It will be apparent to those skilled in the art that various modifications and variations can be made to the present invention without departing from the spirit or scope of the invention. Thus, it is intended that the present invention also include such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
Claims (5)
1. An active neutron personal dosimeter based on a three-layer silicon detector, which is characterized in that: the nuclear power system comprises a main detector, an outer layer transformant component and a nuclear electronics system, wherein the main detector and the outer layer transformant component comprise three passivation injection plane silicon detectors, an Open detector, a Fast detector and an Albedo detector are sequentially arranged from top to bottom, and a polyethylene layer are arranged in front of the Open detector 6 A LiF coating, a polyethylene layer arranged in front of the Fast detector, a boron-containing polyethylene layer, a polyethylene layer and a boron-containing polyethylene layer arranged in front of the Albedo detector 6 A LiF coating; the Fast detector only responds to neutrons with energy greater than 1MeV, and can only record recoil protons generated by elastic scattering of Fast neutrons and hydrogen nuclei; the Open detector and the Albedo detector can record low-energy neutrons and low-energy neutrons 6 Alpha particles generated by Li reaction are identical to tritiumRecoil protons generated by fast neutrons can be recorded; the low-energy neutron response of the Albedo detector is lower than that of the Open detector, and the medium-energy neutron response is higher than that of the Open detector; the nuclear electronics system provides high voltage to each detector and acquires detection signals for multi-channel analysis.
2. The active neutron personal dosimeter based on a three-layer silicon detector of claim 1, wherein: the nuclear optics system comprises a pre-amplifier, a main amplifier, a low-voltage power supply, a high-voltage power supply and a multi-channel analyzer, wherein detection signals of three passivation injection plane silicon detectors are sent into the multi-channel analyzer through the pre-amplifier and the main amplifier.
3. The active neutron personal dosimeter based on a three-layer silicon detector of claim 1, wherein: the three passivation injection plane silicon detectors are packaged in an aluminum shell.
4. A method of measuring an active neutron personal dosimeter based on a three layer silicon detector according to any of claims 1 to 3, wherein: the method adopts a direct reading mode, and firstly, neutron energy is divided into three parts: the method comprises the steps of obtaining the fluence responses of each detector of a personal dosimeter under different energies through a single-energy neutron fluence scale experiment in a 20 meV-1 keV slow neutron energy region, a 1 keV-1 meV medium energy neutron energy region and a 1 meV-20 meV fast neutron energy region, calculating the average fluence responses of different energy regions, and obtaining the personal dose equivalent response by using fluence-personal dose equivalent conversion coefficients; the count of the Fast detector is divided by the personal dose equivalent of the Fast neutron contribution, the difference between the count of the Open detector and the count of the Fast detector is divided by the personal dose equivalent of the Open detector of the low energy neutron contribution, the difference between the count of the Albedo detector and the count of the Open detector is divided by the medium energy neutron response, the personal dose equivalent of the medium energy neutron contribution, and the sum of the three is the total neutron personal dose equivalent.
5. A method of measuring an active neutron personal dosimeter based on a three layer silicon detector according to any of claims 1 to 3, wherein: the spectrum decomposition mode is adopted, and the method utilizes the relation between the energy and angle of recoil protons and neutron energy as follows:
E p ＝E n cos 2 φ
wherein E is p For recoil proton energy, E n For the energy of an incident neutron, phi is the recoil angle,
deconvolving the pulse amplitude spectrum measured by the Fast detector into neutron energy spectrum.
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