A method for measuring the monoenergetic neutron fluence rate above 1MeV

Abstract

The embodiment of the present application discloses a method for measuring the monoenergetic neutron fluence rate above 1MeV, which relates to the field of neutron measurement, makes the measurement result more accurate and reduces the time for measuring the neutron fluence rate. The method for measuring the monoenergetic neutron fluence rate above 1MeV includes the following steps: presetting the distance between the neutron source and the fission ionization chamber, calculating the response parameters of the fission ionization chamber to the monoenergetic neutron source; experimentally obtaining the fission fragment pulse amplitude spectrum, and obtaining the monoenergetic neutron source strength according to the above-threshold fission fragment count rate, the below-threshold fission fragment count correction factor, the U target self-absorption correction factor, and the fission ionization chamber's response parameters to the monoenergetic neutron source; obtaining the agreed true value of the neutron fluence rate at the pre-placement position of the neutron measurement device to be calibrated according to the obtained monoenergetic neutron source strength and the corresponding neutron angular distribution. The method for measuring the monoenergetic neutron fluence rate above 1MeV of the present application is used to measure the neutron fluence rate.

Description

Technical Field

[0001] The present application relates to but is not limited to the field of neutron measurement, and in particular to a method for measuring the monoenergetic neutron fluence rate above 1 MeV. Background Art

[0002] At present, neutron detectors are widely used in security inspection and aerospace fields. The method of detecting radiation in the universe is to use space radiation neutron detectors to detect neutron rays in the universe. In order to ensure that the detection accuracy of the neutron detector can meet the detection requirements, it is necessary to use a neutron reference radiation field to calibrate the neutron detector.

[0003] In order to enable the monoenergetic neutron reference radiation field to be used to determine the fluence response of the neutron measuring device, the neutron fluence rate at the calibration position of the neutron measuring device must be accurately measured first. The traditional measurement method is to determine the position of the fluence response of the neutron measuring device in the direction of the neutron beam or the charged particle beam and use the corresponding neutron fluence rate standard device to measure the neutron fluence rate at the corresponding position. For the case of using a charged particle beam to produce monoenergetic neutrons, the law that the neutron fluence rate is inversely proportional to the square of the distance from the nuclear reaction target can also be used to determine the neutron fluence rate at other positions in the direction of the charged particle beam, and use the appropriate neutron fluence rate to calibrate the corresponding neutron measuring device. For a neutron fluence standard device of a certain size, only by placing the standard device at a distance away from the nuclear reaction target, Figure 1 For the neutron fluence marking equipment fluence response model, for the neutron fluence standard equipment 01, the neutron field 02 at the position where the neutron fluence rate is to be measured can be regarded as a parallel aligning field. Figure 1 The calculation model of the laboratory's existing U-235 fission chamber calculated the fluence response results at 1.2MeV, 2.5MeV, 5MeV, 14.8MeV, and 19MeV as shown in Table 1:

[0004] Table 1

[0005] Neutron energy / MeV <![CDATA[注量响应 / cm 2 > 1.2 2.57620E-06 2.5 2.67276E-06 5 2.29504E-06 14.8 4.15980E-06 19 4.15915E-06

[0006] Typical values ​​of neutron flux rates above 1 MeV at 1 m from the target produced by the 5SDH-2 tandem accelerator of the China Institute of Atomic Energy are shown in Table 2:

[0007] Table 2

[0008] Neutron energy / MeV <![CDATA[Typical value of neutron fluence rate / (cm -2 s -1 )]]> 1.2 5.60000E+02 2.5 8.00000E+02 5 2.40000E+03 14.8 5.60000E+02 19 1.60000E+02

[0009] From Tables 1 and 2, it can be estimated that the time required for the U-235 fission ionization chamber to obtain 10,000 fission fragment counts at 1 m from the target is shown in Table 3:

[0010] Table 3

[0011] Neutron energy / MeV <![CDATA[Fission fragment count rate / s -1 > Time for 10,000 fission fragment counts / h 1.2 1.44267E-03 1.92544E+03 2.5 2.13821E-03 1.29911E+03 5 5.50810E-03 5.04308E+02 14.8 2.32949E-03 1.19244E+03 19 6.65464E-04 4.17420E+03

[0012] It can be seen from Table 3 that it takes at least 500 hours to obtain 10,000 fission fragment counts, and the experimental time cost is too expensive.

[0013] In order to make the measurement result more accurate and reduce the time of measuring the neutron fluence rate, the present application proposes a method for measuring the monoenergetic neutron fluence rate above 1 MeV. Summary of the invention

[0014] The embodiment of the present application provides a method for measuring the monoenergetic neutron fluence rate above 1 MeV, which makes the measurement result more accurate and reduces the time for measuring the neutron fluence rate.

[0015] In order to achieve the above-mentioned purpose, the technical solution of the embodiment of the present application is implemented as follows:

[0016] The present application provides a method for measuring a monoenergetic neutron fluence rate of more than 1 MeV, comprising the following steps: presetting a distance between a neutron source and a fission ionization chamber, and calculating a response parameter of the fission ionization chamber to the monoenergetic neutron source; experimentally obtaining a fission fragment pulse amplitude spectrum, and obtaining a monoenergetic neutron source intensity according to an above-threshold fission fragment count rate, a below-threshold fission fragment count correction factor, a U target self-absorption correction factor, and a response parameter of the fission ionization chamber to the monoenergetic neutron source; and obtaining a conventional true value of a neutron fluence rate at a pre-placed position of a neutron measurement device to be calibrated according to the obtained monoenergetic neutron source intensity and a corresponding neutron angular distribution.

[0017] The present application provides a method for measuring a monoenergetic neutron fluence rate above 1 MeV, comprising the following steps: firstly, presetting the distance between a neutron source and a fission ionization chamber, and calculating the response parameters of the fission ionization chamber to the monoenergetic neutron source; then, obtaining a fission fragment pulse amplitude spectrum through an experiment, and obtaining a monoenergetic neutron source intensity according to an above-threshold fission fragment count rate, a below-threshold fission fragment count correction factor, a U target self-absorption correction factor, and the response parameters of the fission ionization chamber to the monoenergetic neutron source; and finally, obtaining a conventional true value of a neutron fluence rate at a pre-placed position of a neutron measurement device to be calibrated according to the obtained monoenergetic neutron source intensity and the corresponding neutron angular distribution. The present application uses the monoenergetic neutron source strength and the corresponding neutron angular distribution to obtain the agreed true value of the neutron fluence rate at the pre-placed position of the neutron measurement device to be calibrated. Compared with the related art in which the standard device is placed at a distance away from the nuclear reaction target, the neutrons at the position where the neutron fluence rate to be measured are regarded as a parallel and aligned field, so that it tends to the ideal state of a parallel and aligned field. There is a deviation between the measurement result and the actual neutron fluence rate. The present application reduces the distance between the neutron fluence rate measurement standard device and the nuclear reaction target, obtains the source strength of the corresponding monoenergetic neutron source through measurement, and then selects a sampling area of ​​the neutron source according to the angular distribution of the monoenergetic neutrons generated by the nuclear reaction to infer the neutron fluence rate used to determine the fluence or energy response position of the neutron measurement device. The measurement result is more accurate and the time for measuring the neutron fluence rate is reduced.

[0018] In a possible implementation of the present application, the distance between the neutron source and the fission ionization chamber is preset, and the step of calculating the response parameters of the fission ionization chamber to the monoenergetic neutron source includes: according to the preset distance between the neutron source and the fission ionization chamber and the angular distribution of the neutron source, in order to reduce the time used for calculation, the sampling area is determined, and the response parameters of the fission ionization chamber to the neutrons in the sampling area are calculated; according to the angular distribution of neutrons in the neutron source, the proportion of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle is determined, and the response parameters of the fission ionization chamber to the monoenergetic neutron source are determined in combination with the response parameters of the fission ionization chamber to the neutrons in the sampling area. Due to the low measurement efficiency of the fission ionization chamber, the technical solution provided by the present application only needs to calculate the response parameters of the fission ionization chamber to the neutrons in the sampling area, and then determine the proportion of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle in combination with the angular distribution of neutrons in the neutron source to determine the response parameters of the fission ionization chamber to the monoenergetic neutron source, which saves time compared with directly calculating the response parameters of the fission ionization chamber to the monoenergetic neutron source.

[0019] In a possible implementation of the present application, an agreed true value of the neutron fluence rate at a pre-placed position of a neutron measurement device to be calibrated is obtained based on the obtained monoenergetic neutron source strength and the corresponding neutron angular distribution, including: determining the ratio of the number of neutrons in a sampling area to the number of neutrons in a 4π solid angle of the neutron source based on the angular distribution of the neutron source; calculating the ratio of neutrons in the sampling area reaching the front surface of the neutron measurement device to be calibrated; obtaining the agreed true value of the neutron fluence rate at a pre-placed position of the neutron measurement device to be calibrated based on the monoenergetic neutron source strength, the ratio of the number of neutrons in the sampling area to the number of neutrons in a 4π solid angle, the ratio of neutrons in the sampling area that can cover the front surface of the neutron measurement device to be calibrated, and the area that can cover the front surface. The technical solution provided in the present application determines the ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle through the neutron angular distribution of the neutron source, the ratio of the neutrons in the sampling area that can cover the front surface of the neutron measurement device to be calibrated, the area that can cover the front surface, and the strength of the monoenergetic neutron source, and determines four parameter values ​​required for the calculation of the neutron fluence rate. The agreed true value of the neutron fluence rate at the pre-placement position of the neutron measurement device to be calibrated can be obtained through the four parameter values.

[0020] In a possible implementation of the present application, the agreed true value of the neutron fluence rate at the pre-placement position of the neutron measurement device to be calibrated is obtained according to the obtained monoenergetic neutron source strength and the corresponding neutron angular distribution. The calculation formula of the neutron fluence rate is as follows:

[0021]

[0022] in, represents the neutron injection rate; R s Represents the ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle; R a represents the proportion of the surface that can be covered by the neutrons in the sampling area that reach the neutron measurement device to be calibrated; s represents the area of ​​the surface that can be covered by the neutrons in the sampling area that reach the neutron measurement device to be calibrated.

[0023] In a possible implementation of the present application, the calculation formula for the monoenergetic neutron source strength is as follows:

[0024]

[0025] Among them, Q E Represents the neutron source intensity; n E represents the count rate of the above-threshold effect plus background fission fragments measured by the fission ionization chamber; c a represents the U target self-absorption correction factor; c t represents the subthreshold fission fragment correction factor; n Eb Represents the background fission fragment count rate measured by the fission ionization chamber; R E Represents the response parameters of the fission ionization chamber.

[0026] In a possible implementation of the present application, the target self-absorption correction factor C a The calculation formula is as follows:

[0027]

[0028] Among them, C a represents the target self-absorption correction factor; t u represents the mass thickness of the target material; σ represents the standard deviation of the mass thickness of the target material; R represents the average range of the fission fragments in the target material.

[0029] In a possible implementation of the present application, the subthreshold fission fragment correction factor c t The calculation formula is as follows:

[0030]

[0031] Among them, c t represents the subthreshold fission fragment correction factor; N 阈上 represents the above-threshold fission fragment count; N 阈下 Stands for subthreshold fission fragment count.

[0032] In a possible implementation of the present application, the energy of the monoenergetic neutron source is 1.2 MeV.

[0033] In a possible implementation of the present application, the fission chamber is a U-235 fission chamber, and a certain thickness of gadolinium and boron-containing polyethylene are attached to the outside of the U-235 fission chamber to prevent low-energy neutrons scattered by the shielding chamber from entering the U-235 fission chamber.

[0034] In a possible implementation of the present application, the distance between the monoenergetic neutron source and the front surface of the neutron measurement device to be calibrated is 150 cm, and the radius of the circle that can cover the front surface of the neutron measurement device is 21 cm. BRIEF DESCRIPTION OF THE DRAWINGS

[0035] Figure 1 A calculation model for the fluence response of a neutron fluence standard device of the related technology;

[0036] Figure 2 This is one of the flow charts of the method for measuring the monoenergetic neutron fluence rate above 1 MeV in the embodiment of the present application;

[0037] Figure 3 This is the second flow chart of the method for measuring the monoenergetic neutron fluence rate above 1 MeV in the embodiment of the present application;

[0038] Figure 4 This is the third flow chart of the method for measuring the monoenergetic neutron injection rate above 1MeV in the embodiment of the present application;

[0039] Figure 5 It is a schematic diagram of the structure of a conventional back-to-back U-235 fission ionization chamber;

[0040] Figure 6 It is the neutron response calculation model of the U-235 fission ionization chamber in the embodiment of the present application;

[0041] Figure 7 The neutron angular distribution of the 1.2 MeV monoenergetic neutron field in the embodiment of the present application;

[0042] Figure 8 This is a schematic diagram of the proton beam and neutron change process in the embodiment of the present application;

[0043] Fig. 9 This is the 1.2MeV monoenergetic neutron spectrum in the embodiment of this application;

[0044] Fig.10 This is the pulse amplitude spectrum obtained by measuring the 1.2 MeV monoenergetic neutron source intensity in the examples of this application.

[0045] Reference numerals

[0046] 01- Neutron injection standard equipment; 02- Neutron beam; 1- Fission ionization chamber; 101- Ionization chamber shell; 102- Inflatable tube; 103- Top cover; 104- Copper column; 105- Copper cap; 106- Insulating column; 107- Protective ring; 108- High voltage pole and collector; 109- Insulating cap; 110- Remo female head base; 111- Base and low voltage pole; 2- Neutron beam; 3- Neutron source; 4- Vacuum target tube; 41- Target material; 42- Base; 5- Proton beam. DETAILED DESCRIPTION

[0047] It should be noted that, in the absence of conflict, the embodiments and technical features in the embodiments of the present application can be combined with each other, and the detailed description in the specific implementation method should be understood as an explanation of the purpose of the present application and should not be regarded as an improper limitation on the present application.

[0048] In order to make the purpose, technical solution and advantages of the embodiments of the present application clearer, the specific technical solution of the present application will be further described in detail below in conjunction with the drawings in the embodiments of the present application. The following embodiments are used to illustrate the present application, but are not used to limit the scope of the present application.

[0049] In the embodiments of the present application, the terms "first" and "second" are used for descriptive purposes only and are not to be understood as indicating or implying relative importance or implicitly indicating the number of the indicated technical features. Thus, a feature defined as "first" or "second" may explicitly or implicitly include one or more of the features. In the description of the embodiments of the present application, unless otherwise specified, "multiple" means two or more.

[0050] In addition, in the embodiments of the present application, directional terms such as "up", "down", "left" and "right" are defined relative to the orientation of the components in the drawings. It should be understood that these directional terms are relative concepts. They are used for relative description and clarification, and they may change accordingly according to the changes in the orientation of the components in the drawings.

[0051] In the embodiments of the present application, unless otherwise clearly specified and limited, the term "connection" should be understood in a broad sense. For example, "connection" can be a fixed connection, a detachable connection, or an integral connection; it can be a direct connection or an indirect connection through an intermediate medium.

[0052] In the embodiments of the present application, the terms "include", "comprises" or any other variants thereof are intended to cover non-exclusive inclusion, so that a process, method, article or device including a series of elements includes not only those elements, but also includes other elements not explicitly listed, or also includes elements inherent to such process, method, article or device. In the absence of further restrictions, an element defined by the sentence "comprises a ..." does not exclude the presence of other identical elements in the process, method, article or device including the element.

[0053] In the embodiments of the present application, words such as "exemplary" or "for example" are used to indicate examples, illustrations or descriptions. Any embodiment or design described as "exemplary" or "for example" in the embodiments of the present application should not be interpreted as being more preferred or more advantageous than other embodiments or designs. Specifically, the use of words such as "exemplary" or "for example" is intended to present related concepts in a specific way.

[0054] At present, neutron detectors are widely used in security inspection and aerospace fields. The method of detecting radiation in the universe is to use space radiation neutron detectors to detect neutron rays in the universe. In order to ensure that the detection accuracy of the neutron detector can meet the detection requirements, it is necessary to use a neutron reference radiation field to calibrate the neutron detector.

[0055] The standard equipment used for measuring neutron fluence rate above 1MeV includes scintillator telescope, semiconductor telescope, U-235 fission ionization chamber, and U-238 fission ionization chamber. The fission ionization chamber uses neutrons to undergo fission reaction with U isotopes in the U target in the fission ionization chamber to produce fission fragments. The fission fragments are ionized in the working gas in the fission ionization chamber to produce electrical signals.

[0056] Reference Figure 2 The present application provides a method for measuring the monoenergetic neutron fluence rate above 1 MeV, comprising the following steps:

[0057] Step S1: presetting the distance between the neutron source and the fission ionization chamber, and calculating the response parameters of the fission ionization chamber to the monoenergetic neutron source;

[0058] Step S2: experimentally obtaining a fission fragment pulse amplitude spectrum, and obtaining a monoenergetic neutron source intensity according to an above-threshold fission fragment count rate, a below-threshold fission fragment count correction factor, a U target self-absorption correction factor, and a response parameter of a fission ionization chamber to a monoenergetic neutron source;

[0059] Step S3: obtaining the agreed true value of the neutron flux rate at the pre-placed position of the neutron measurement device to be calibrated according to the obtained monoenergetic neutron source strength and the corresponding neutron angular distribution.

[0060] The present application provides a method for measuring the monoenergetic neutron fluence rate above 1MeV, comprising the following steps: firstly, the distance between the neutron source and the fission ionization chamber is preset, and the response parameters of the fission ionization chamber to the monoenergetic neutron source are calculated; then, the fission fragment pulse amplitude spectrum is obtained experimentally, and the monoenergetic neutron source intensity is obtained according to the above-threshold fission fragment count rate, the below-threshold fission fragment count correction factor, the U target self-absorption correction factor, and the response parameters of the fission ionization chamber to the monoenergetic neutron source; finally, the agreed true value of the neutron fluence rate at the pre-placement position of the neutron measurement device to be calibrated is obtained according to the obtained monoenergetic neutron source intensity and the corresponding neutron angular distribution. The present application adopts the monoenergetic neutron source intensity and the corresponding neutron angular distribution to obtain the agreed true value of the neutron fluence rate at the pre-placement position of the neutron measurement device to be calibrated. Compared with the related art in which the standard equipment is placed at a distance from the nuclear reaction target, the neutrons at the position where the neutron fluence rate to be measured are regarded as a parallel aligning field, so that it tends to the ideal state of a parallel aligning field, and there is a deviation between the measurement result and the actual neutron fluence rate, the present application reduces the distance between the neutron fluence rate measurement standard equipment and the nuclear reaction target, obtains the source strength of the corresponding monoenergetic neutron source through measurement, and then selects a sampling area of ​​the neutron source according to the angular distribution of the monoenergetic neutrons generated by the nuclear reaction to infer the neutron fluence rate for determining the fluence or energy response position of the neutron measurement measurement equipment. The measurement result is more accurate and the time for measuring the neutron fluence rate is reduced.

[0061] Reference Figure 2The method for measuring the monoenergetic neutron fluence rate above 1MeV provided in the present application comprises the following steps: step S1: presetting the distance between the neutron source and the fission ionization chamber, and calculating the response parameters of the fission ionization chamber to the monoenergetic neutron source; step S2: experimentally obtaining the fission fragment pulse amplitude spectrum, and obtaining the monoenergetic neutron source strength according to the above-threshold fission fragment count rate, the below-threshold fission fragment count correction factor, the U target self-absorption correction factor, and the response parameters of the fission ionization chamber to the monoenergetic neutron source; step S3: obtaining the agreed true value of the neutron fluence rate at the pre-placement position of the neutron measurement device to be calibrated according to the obtained monoenergetic neutron source strength and the corresponding neutron angular distribution. The present application adopts the monoenergetic neutron source strength and the corresponding neutron angular distribution to obtain the agreed true value of the neutron fluence rate at the pre-placement position of the neutron measurement device to be calibrated. Compared with the related art in which the standard equipment is placed at a distance from the nuclear reaction target, the neutrons at the position where the neutron fluence rate to be measured are regarded as a parallel aligning field, so that it tends to the ideal state of a parallel aligning field, and there is a deviation between the measurement result and the actual neutron fluence rate, the present application reduces the distance between the neutron fluence rate measurement standard equipment and the nuclear reaction target, obtains the source strength of the corresponding monoenergetic neutron source through measurement, and then selects a sampling area of ​​the neutron source according to the angular distribution of the monoenergetic neutrons generated by the nuclear reaction to infer the neutron fluence rate for determining the fluence or energy response position of the neutron measurement equipment. The measurement result is more accurate and the time for measuring the neutron fluence rate is reduced.

[0062] Furthermore, if Figure 3 As shown, step S1: presetting the distance between the neutron source and the fission ionization chamber, and calculating the response parameters of the fission ionization chamber to the monoenergetic neutron source, including:

[0063] Step S11: according to the preset distance between the neutron source and the fission ionization chamber and the angular distribution of the neutron source, in order to reduce the time used for calculation, the sampling area is determined, and the response parameters of the fission ionization chamber to the neutrons in the sampling area are calculated;

[0064] Step S12: determining the ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle according to the neutron angular distribution of the neutron source, and determining the response parameters of the fission ionization chamber to the monoenergetic neutron source in combination with the response parameters of the fission ionization chamber to the neutrons in the sampling area.

[0065] Since the measurement efficiency of the fission ionization chamber is low, the technical solution provided in the present application only needs to calculate the response parameters of the fission ionization chamber to the neutrons in the sampling area, and then determine the ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle in combination with the neutron angular distribution of the neutron source to determine the response parameters of the fission ionization chamber to the monoenergetic neutron source, which saves time compared to calculating the response parameters of the fission ionization chamber to the monoenergetic neutron source.

[0066] Furthermore, if Figure 4As shown, step S3: obtaining the agreed true value of the neutron fluence rate at the pre-placed position of the neutron measurement device to be calibrated according to the obtained monoenergetic neutron source strength and the corresponding neutron angular distribution, including:

[0067] Step S31: determining the ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle of the neutron source according to the angular distribution of the neutron source;

[0068] Step S32: calculating the proportion of neutrons in the sampling area reaching the front surface of the neutron measurement device to be calibrated;

[0069] Step S33: Obtain the agreed true value of the neutron injection rate at the pre-placement position of the neutron measurement device to be calibrated according to the monoenergetic neutron source strength, the ratio of the number of neutrons in the sampling area to the number of neutrons in a 4π solid angle, the ratio of the neutrons in the sampling area reaching the front surface of the neutron measurement device to be calibrated that can cover the front surface, and the area of ​​the surface that can cover the front surface.

[0070] The technical solution provided by the present application determines the proportion of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle through the neutron angular distribution of the neutron source, the proportion of the surface that can cover the front surface of the neutron measurement device to be calibrated when the neutrons in the sampling area reach the front surface, the area of ​​the surface that can cover the front surface, and the monoenergetic neutron source strength. The four parameter values ​​required for the calculation of the neutron fluence rate are determined, and the agreed true value of the neutron fluence rate at the pre-placement position of the neutron measurement device to be calibrated can be obtained through the four parameter values. The present application provides a specific embodiment to describe the measurement method of the monoenergetic neutron fluence rate above 1MeV, and the embodiment is a 1.2MeV monoenergetic neutron source strength and the average neutron fluence rate of a circular surface with a radius of 21cm at 1.5m away from the target on the proton beam line.

[0071] Neutron detectors need to be checked and calibrated in a neutron reference radiation field before use. The energy and generation method of the monoenergetic neutron reference radiation field recommended by the International Organization for Standardization for determining the fluence and energy response of neutron measurement equipment are shown in Table 4:

[0072] Table 4

[0073]

[0074] Reference Figure 5The conventional fission ionization chamber 1 includes eleven parts, including the outer shell 101 of the ionization chamber, the gas filling tube 102, the top cover 103, the copper column 104, the copper cap 105, the insulating column 106, the protective ring 107, the high voltage and collecting electrode 108, the insulating cap 109, the Leimo mother casting base 110 and the bottom lining and the low voltage electrode 111. The substrate 111 is platinum with a diameter of 36 mm and a thickness of 0.3 mm. The high-voltage and collecting electrodes 108 are oxygen-free copper with a diameter of 36 mm and a thickness of 0.2 mm, and the inter-electrode distance is 20 mm. There is a protective ring 107 in the middle of the two electrodes to ensure the uniformity of the central electric field and prevent leakage current. The protective ring 107 has an inner diameter of 29 mm, an outer diameter of 36 mm, and a thickness of 0.2 mm. The material is oxygen-free copper. The uranium target is plated on the substrate and the low-voltage electrode 111. The diameter of the plated uranium target is 25 mm. Table 5 shows the abundance of U isotopes in the target, and Table 6 shows the number of U isotope nuclei in the target material.

[0075] Table 5

[0076] <![CDATA[ 234 In(%)]]> Std Err <![CDATA[ 235 In(%)]]> Std Err <![CDATA[ 236 In(%)]]> Std Err <![CDATA[ 238 In(%)]]> Std Err 1.262 0.004 90.118 0.0036 0.2294 0.0014 8.390 0.0031

[0077] Table 6

[0078] U isotope name Number of nuclei U-234 2.545E+16 U-235 1.82378E+18 U-236 4.75075E+15 U-238 1.6967E+17

[0079] When using a U-235 fission ionization chamber to measure the monoenergetic neutron source intensity, the neutron source intensity can be expressed as formula 1:

[0080]

[0081] In the formula, Q E Represents the neutron source intensity; n E represents the count rate of the above-threshold effect plus background fission fragments measured by the fission ionization chamber; c a represents the U target self-absorption correction factor; c t represents the subthreshold fission fragment correction factor; n Eb Represents the background fission fragment count rate measured by the fission ionization chamber; R E Represents the response parameters of the fission ionization chamber.

[0082] Among them, the U target self-absorption correction factor c a Calculated by formula 2:

[0083]

[0084] In the formula, c a represents the target self-absorption correction factor; t u represents the target material mass thickness; σ represents the standard deviation of the target material mass thickness; R represents the average range of fission fragments in the target material

[0085] Since the U-235 fission ionization chamber is selected in this embodiment, the target material is a uranium target, and the mass thickness of the uranium target is 161.1 μg / cm 2 The standard deviation of mass thickness is 1.10 μg / cm 2 The average range of fission fragments in the target material is 7.5E+3μg / cm 2 , the U target self-absorption correction factor C is calculated by formula 2 a It is 1.010778.

[0086] Next, calculate the response parameter R of the fission chamber E The response calculation model of the fission ionization chamber is as follows: Figure 6 As shown, the neutron source 3 emits a neutron beam 2, and the neutrons 2 in the sampling area of ​​the neutron source are selected to enter the uranium target of the fission ionization chamber 1 and undergo fission, so that the fission ionization chamber 3 responds to the neutrons 2 in the sampling area.

[0087] The fission ionization chamber used in the embodiments of the present application is Figure 5 On the basis of the 0.2mm thick gadolinium and 3mm thick boron-containing polyethylene, the shielding room scatters low-energy neutrons into the fission ionization chamber. The gas tube and the Lemo interface are not described in the response calculation. In order to improve the calculation efficiency, the sampling angle range is 0.945519~1. In order to make the simulation more accurate, the neutron angular distribution is used for sampling. The angular distribution data of the 1.2MeV monoenergetic neutron field is shown in Figure 7 shown.

[0088] Since the monoenergetic neutrons to be produced have a certain yield, refer to Figure 8 The proton beam 5 bombards the target material 42 in the vacuum target tube 4 to generate monoenergetic neutrons of a certain energy. The fission ionization chamber 1 can respond to the monoenergetic neutrons. Since the size of the monoenergetic neutrons in the embodiment of the present application is 1.2 MeV, the target material 42 is set as a tritium target. In order to ensure a certain yield of the emitted monoenergetic neutrons, the thickness of the tritium target used is 710 μg / cm 2 In addition, in order to support such a thin tritium target, a 3 mm thick tantalum substrate 41 was used. Therefore, the emitted 1.2 MeV monoenergetic neutrons have a certain degree of broadening. The neutron energy spectrum of the front surface of the U-235 fission ionization chamber calculated by the Target software is as follows: Fig. 9 shown.

[0089] use Fig. 9 The neutron spectrum and Figure 7 The response of the U-235 fission chamber to the neutrons in the sampling area calculated by the neutron angular distribution is: 1.39981E-07, due to Figure 6 In the calculation model, the number of neutrons in the sampling area accounts for only a part of the number of neutrons in the 4π solid angle. Figure 7The ratio of the number of neutrons in the sampling area calculated from the neutron angular distribution to the number of neutrons in the 4π solid angle is: 0.0480223. Therefore, the response of the U-235 fission chamber to the 1.2MeV monoenergetic neutron source is: 6.72221E-09.

[0090] The measurement of 1.2MeV monoenergetic neutron source intensity was carried out in 5SDH-2 tandem accelerator. The detector was placed at 0° angle and the distance from the center of the target was 10cm. The pulse amplitude spectrum obtained by measuring the 1.2MeV monoenergetic neutron source intensity is as follows: Fig.10 As shown:

[0091] Subthreshold fission fragment correction factor c t It can be expressed as formula 3:

[0092]

[0093] In the formula, c t represents the subthreshold fission fragment correction factor; N 阈上 represents the above-threshold fission fragment count; N 阈下 represents the subthreshold fission fragment count. The measurement time is 208796s, the above-threshold fission fragment count is 10873, the signal-noise discrimination threshold is 25, the extrapolated subthreshold fission fragment count is 263, and the subthreshold fission fragment correction factor is 1.024188.

[0094] In addition, according to the target self-absorption correction factor of 1.010778, the total fission fragment count is 11256, and the fission fragment count rate is 5.39092E-02s -1 , the monitor count for the same time period is 22181919.

[0095] The background is measured using the shadow cone method. The live time of the background measurement is 14899s, the background fission fragment count is 10, the monitor count is 1514804, and the background fission fragment count rate normalized by the monitor count rate when the detector is 10cm away from the center of the target is 7.01327E-04s -1 , then the detector effect fission fragment counting rate is 5.32078E-02s -1 , then according to the response of the U-235 fission chamber to 1.2MeV monoenergetic neutrons, the 1.2MeV monoenergetic neutron source intensity is calculated to be 7.91522E+06s -1 .

[0096] Here, we take the working standard long counter of the calibration laboratory as an example to give the neutron flux rate at the typical placement position of the long counter at 150 cm from the target. The radius of the circular surface that can cover the front surface of the long counter is 21 cm. The neutron flux rate on the front surface of the long counter can be expressed as Formula 4:

[0097]

[0098] In the formula, represents the neutron injection rate; R s Represents the ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle; R a represents the proportion of neutrons in the sampling area reaching the front surface of the long counter; s represents the area of ​​the surface that can cover the front surface of the long counter.

[0099] The sampling area used to calculate the neutron flux rate on the front surface of the long counter is the same as that used to calculate the neutron source strength. The ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle is 0.0480223; the proportion of neutrons reaching the front surface of the long counter calculated by Monte Carlo is 3.65931E-01; the neutron flux rate at 150 cm from the target calculated by formula 4 is 1.00396E+02cm -2 s -1 .

[0100] From the above description, it can be seen that the time for obtaining the effect plus background fission fragment pulse amplitude spectrum of 1.2MeV monoenergetic neutrons by measuring the neutron source intensity is 57.99889h, while the traditional method of measuring neutron fluence rate uses 1.92544E+03h. The time used by the traditional method is 33 times that of the current method. It can be seen that using the current method to measure the neutron fluence rate greatly saves time cost.

[0101] The above description is only a preferred embodiment of the present invention and is not intended to limit the scope of protection of the present invention. Any modifications, equivalent substitutions and improvements made within the spirit and principles of the present invention should be included in the scope of protection of the present invention.

Claims

1. A method for measuring the monoenergetic neutron fluence rate above 1 MeV, characterized in that: The steps include: Preset the distance between the neutron source and the fission ionization chamber, and calculate the response parameters of the fission ionization chamber to the monoenergetic neutron source; The fission fragment pulse amplitude spectrum is obtained experimentally, and the monoenergetic neutron source intensity is obtained according to the above-threshold fission fragment count rate, below-threshold fission fragment count correction factor, U target self-absorption correction factor, and the response parameters of the fission ionization chamber to the monoenergetic neutron source. According to the obtained monoenergetic neutron source intensity and the corresponding neutron angular distribution, the agreed true value of the neutron injection rate at the pre-placed position of the neutron measurement device to be calibrated is obtained; The calculation formula of the monoenergetic neutron source strength is as follows: Among them, Q E represents the monoenergetic neutron source intensity; n E represents the count rate of the above-threshold effect plus background fission fragments measured by the fission ionization chamber; c a represents the U target self-absorption correction factor; c t represents the subthreshold fission fragment correction factor; n Eb Represents the background fission fragment count rate measured by the fission ionization chamber; R E Represents the response parameters of the fission ionization chamber.

2. The method for measuring the monoenergetic neutron fluence rate above 1 MeV according to claim 1, characterized in that: The preset distance between the neutron source and the fission ionization chamber and the calculation of the response parameters of the fission ionization chamber to the monoenergetic neutron source include: According to the preset distance between the neutron source and the fission ionization chamber and the angular distribution of the neutron source, a sampling area is determined to reduce the time used for calculation, and the response parameters of the fission ionization chamber to the neutrons in the sampling area are calculated; The ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle is determined according to the neutron angular distribution of the neutron source, and the response parameters of the fission ionization chamber to the monoenergetic neutron source are determined in combination with the response parameters of the fission ionization chamber to the neutrons in the sampling area.

3. The method for measuring the monoenergetic neutron fluence rate above 1 MeV according to claim 1, characterized in that: The method of obtaining the agreed true value of the neutron flux rate at the pre-placed position of the neutron measurement device to be calibrated according to the obtained monoenergetic neutron source strength and the corresponding neutron angular distribution includes: Determine the ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle of the neutron source according to the angular distribution of the neutron source; Calculating the proportion of neutrons in the sampling area reaching the front surface of the neutron measurement device to be calibrated; According to the monoenergetic neutron source strength, the ratio of the number of neutrons in the sampling area to the number of neutrons in a 4π solid angle, the ratio of the neutrons in the sampling area reaching the front surface of the neutron measurement device to be calibrated that can cover the front surface, and the area of ​​the surface that can cover the front surface, the agreed true value of the neutron injection rate at the pre-placement position of the neutron measurement device to be calibrated is obtained.

4. The method for measuring the monoenergetic neutron fluence rate above 1 MeV according to claim 3, characterized in that: The agreed true value of the neutron flux rate at the pre-placement position of the neutron measurement device to be calibrated is obtained according to the obtained monoenergetic neutron source strength and the corresponding neutron angular distribution. The calculation formula of the neutron flux rate is as follows: in, represents the neutron injection rate; R s Represents the ratio of the number of neutrons in the sampling area to the number of neutrons in the 4π solid angle; R a represents the proportion of neutrons in the sampling area that reach the surface that can cover the front surface of the neutron measurement device to be calibrated; s represents the area of ​​the surface that can cover the front surface of the neutron measurement device to be calibrated.

5. The method for measuring the monoenergetic neutron fluence rate above 1 MeV according to claim 1, characterized in that: The U target self-absorption correction factor c a The calculation formula is as follows: Among them, c a represents the target self-absorption correction factor; t u represents the mass thickness of the target material; σ represents the standard deviation of the mass thickness of the target material; R represents the average range of the fission fragments in the target material.

6. The method for measuring the monoenergetic neutron fluence rate above 1 MeV according to claim 1, characterized in that: The subthreshold fission fragment correction factor c t The calculation formula is as follows: Among them, c t represents the subthreshold fission fragment correction factor; N 阈上 represents the above-threshold fission fragment count; N 阈下 Stands for subthreshold fission fragment count.

7. The method for measuring the monoenergetic neutron fluence rate of 1 MeV or above according to any one of claims 1 to 6, wherein the fission ionization chamber is a U-235 fission ionization chamber, and a certain thickness of gadolinium and boron-containing polyethylene are attached to the outside of the U-235 fission ionization chamber to prevent low-energy neutrons scattered by the shielding chamber from entering the U-235 fission ionization chamber.

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