A parametric real-time measurement method for on-site neutron spectrum and dose

Abstract

The application provides a parameterized real-time measurement method for field neutron energy spectrum and dose, which comprises the following steps: using a three-section function model to represent the field neutron energy spectrum; constructing simulated field neutron energy spectrums of multiple moderation levels, and using a detector to measure and calculate a thermal-fast ratio; simulating and calculating the simulated field neutron energy spectrum of a test point by using a Monte Carlo model to obtain model parameters; establishing a corresponding relationship between the model parameters and the thermal-fast ratio; measuring the fast neutron pulse count rate and the thermal neutron pulse count rate by using the detector in the field, determining the value of the model parameters by looking up a table or interpolation, and the like according to the corresponding relationship between the model parameters and the thermal-fast ratio; calculating the field neutron energy spectrum according to the model parameters, and calculating the field neutron dose by using the field neutron energy spectrum and a neutron fluence-dose conversion coefficient.

Description

TECHNICAL FIELD

[0001] The present application belongs to the technical field of nuclear radiation monitoring, and particularly relates to a parameterized real-time measurement method for on-site neutron energy spectrum and dose. BACKGROUND

[0002] Affected by nuclear fission chain reaction, a reactor will generate a large amount of neutrons and gamma rays. Measuring the neutron and gamma ray radiation level of the environment around the reactor is of great significance for monitoring the operation state of the reactor. The measurement of the radiation level around the reactor includes fixed monitoring and patrol monitoring in terms of mode, and dose measurement and energy spectrum measurement in terms of content. The fixed monitoring has lower requirements for the volume and weight of the instrument, and the patrol monitoring requires the instrument to be portable. The dose is a core parameter for evaluating the radiation level, and the energy spectrum can be used to identify the types of radionuclides and analyze the cause and scale of nuclear leakage.

[0003] Neutrons and gamma rays are both indirect ionizing radiation, and the mechanisms and types of interaction with matter are different, and the commonly used sensitive materials of the detectors are also different. Therefore, different instruments are usually used to measure neutrons and gamma rays respectively. At present, there are various types of gamma ray measuring instruments, and the technology is relatively mature, such as high-pressure ionization chambers for fixed monitoring, portable gamma radiation instruments and portable gamma spectrometers for patrol monitoring, etc. However, neutron measuring instruments are relatively scarce, mainly including remmeters for measuring the neutron ambient dose equivalent, time-of-flight spectrometers, organic scintillators and He proportional counters, etc. The neutron energy spectrum measuring instruments are affected by factors such as method principle, energy spectrum inversion and detection efficiency, and are not suitable for on-site neutron energy spectrum measurement. In addition, due to the presence of polyethylene moderators, the volume and weight of the remmeter are much larger than those of the portable gamma radiation instrument, which does not affect the fixed monitoring, but it is difficult to use portably. Therefore, the miniaturization and integration of on-site neutron energy spectrum and dose measurement have not been solved. 3 He proportional counter, etc. are affected by factors such as method principle, energy spectrum inversion and detection efficiency, and are not suitable for on-site neutron energy spectrum measurement. In addition, due to the presence of polyethylene moderators, the volume and weight of the remmeter are much larger than those of the portable gamma radiation instrument, which does not affect the fixed monitoring, but it is difficult to use portably. Therefore, the miniaturization and integration of on-site neutron energy spectrum and dose measurement have not been solved. SUMMARY

[0004] (I) Technical problem to be solved

[0005] The present application provides a parameterized real-time measurement method for on-site neutron energy spectrum and dose to solve the technical problem of realizing portable integrated real-time measurement of on-site neutron energy spectrum and dose.

[0006] (II) Technical scheme

[0007] In order to solve the above technical problem, the present application provides a parameterized real-time measurement method for on-site neutron energy spectrum and dose, which comprises the following steps:

[0008] S1. Using a three-section function model to represent the on-site neutron energy spectrum

[0009] The on-site neutron energy spectrum is divided into a thermal energy region, a medium energy region and a fast energy region, and the on-site neutron energy spectrum is represented by a three-section function model. is described by formula (1):

[0010] (1)

[0011] wherein, , and are the neutron energy spectrum of the thermal energy region, the intermediate energy region and the fast energy region, respectively, is the demarcation energy point of the thermal energy region and the intermediate energy region, is the demarcation energy point of the intermediate energy region and the fast energy region; , and are respectively as formula (2)-(6):

[0012] (2)

[0013] (3)

[0014] (4)

[0015] wherein, A and B are the fluences of and respectively, representing the amplitude of the function; a and b are the energy corresponding to the temperature respectively, representing the spectral shape of the function; and are the intercept and slope of the logarithmic linear function respectively; and are calculated by and at and respectively, as (5) and formula (6):

[0016] (5)

[0017] (6)

[0018] S2. Construct simulated field neutron energy spectrum of multiple moderation levels using different thicknesses of shielding materials, and use a detector to measure to obtain fast neutron pulse count rate and thermal neutron pulse count rate , and calculate the thermal fast ratio r using formula (7):

[0019] (7)

[0020] S3. Simulate and calculate the simulated field neutron energy spectrum of the test point by the Monte Carlo model , and fit using formula (1) to obtain the model parameters A, a, B and b in ;

[0021] S4. Establish the corresponding relationship between model parameters A, a, B and b and r

[0022] S5. Measuring the fast neutron pulse count rate using a detector on site and thermal neutron pulse count rate , according to the corresponding relationship between model parameters A, a, B and b and r, determine the values ​​of A, a, B and b by looking up the table or interpolation;

[0023] S6. Calculate the on-site neutron spectrum according to the model parameters A, a, B and b according to formula (1) , and then using on-site neutron spectroscopy and neutron fluence-dose conversion factor Calculate the on-site neutron dose H according to formula (8):

[0024] ⑻

[0025] Where c is the calibration factor used to correct Different conversion factors are used to calculate different neutron doses. ; The unit is fluence rate or fluence. When fluence rate is used, it is calculated in real time based on the fast neutron pulse count rate and the thermal neutron pulse count rate. Formula (8) calculates the on-site neutron dose rate; when fluence is used, it is calculated based on the total fast neutron pulse count and the total thermal neutron pulse count. At this time, Formula (8) calculates the on-site neutron cumulative dose.

[0026] Further, in step S2, the simulated on-site neutron spectrum is obtained by 252 A Cf isotope neutron source or an accelerator monoenergetic neutron source is constructed in conjunction with a moderator material, and the energy spectrum distribution is obtained through experimental measurement or Monte Carlo simulation calculation.

[0027] Furthermore, in step S2, the fast neutron pulse count rate and the thermal neutron pulse count rate are measured by a fast neutron detector and a thermal neutron detector respectively, or by using a detector containing 10 B. 6 Li detector or composite detector is measured together.

[0028] Furthermore, in step S4, the fluence parameter A and the thermal neutron pulse count rate are first calculated. Ratio , and the fluence parameter B and the fast neutron pulse count rate Ratio , then create and The corresponding relationship with r.

[0029] Further, in step S5, first use and Calculate r according to formula (I), and then determine the values of a and b according to the corresponding relationship between a, b and r; then determine the values of a and b according to the corresponding relationship between a, b and r; then determine the values of a and b according to the corresponding relationship between a, b and r; then A= And And And The value of a and b is determined according to the corresponding relationship between a, b and r; then A= × , B= × .

[0030] (Three) beneficial effects

[0031] The present application provides a kind of parameterization real-time measurement method of field neutron spectrum and dose, which comprises using three-stage function model to represent field neutron spectrum;A plurality of simulated field neutron spectrum of moderation level is constructed, and is measured using detector, and the calculation of thermal fast ratio;The simulated field neutron spectrum of test point is calculated by Monte Carlo model simulation, and the model parameters are obtained;Establish the corresponding relationship between model parameters and thermal fast ratio;Field using detector measures fast neutron pulse count rate and thermal neutron pulse count rate, according to the corresponding relationship between model parameters and thermal fast ratio, the value of the model parameter is determined by table lookup or interpolation etc.;According to model parameters, the field neutron spectrum is calculated, and then the field neutron dose is calculated using field neutron spectrum and neutron fluence-dose conversion coefficient.

[0032] Compared with the prior art, the beneficial effects of the present application are as follows:

[0033] 1. The present application can realize the measurement of field neutron spectrum only by using pulse count rate, without spectrum calculation, with small calculation amount and low difficulty of use, which is significantly superior to organic scintillator and 3He proportional counter tube, etc. neutron spectrometer, and can realize online real-time measurement through single-chip microcomputer or FPGA.

[0034] 2. The method proposed in the present application can realize integrated real-time measurement of field neutron spectrum and dose, without polyethylene moderator, which is significantly superior to remmeter in volume, weight and measurement parameter type, and is more suitable for use in the field around the reactor.

[0035] 3. The method proposed in the present application uses spectrum to calculate field neutron dose, without involving energy response problem, can integrate multiple conversion coefficients to calculate multiple doses at the same time, and only needs to replace neutron-dose conversion coefficient when replacing neutron dose utility, without changing the instrument structure. BRIEF DESCRIPTION OF DRAWINGS

[0036] Figure 1 Comparison of neutron simulation spectrum and fitting spectrum under different thickness of polyethylene;

[0037] Figure 2 Variation of model parameters with thermal fast ratio;

[0038] Figure 3 The neutron ambient dose equivalent rate measurement value and the theoretical value and the relative deviation change with the thickness of polyethylene. DETAILED DESCRIPTION

[0039] In order to make the purpose, content and advantages of the present application clearer, the specific embodiments of the present application are described in further detail below in combination with the drawings and examples.

[0040] The embodiment proposes a parameterized real-time measurement method for field neutron spectrum and dose, which specifically includes the following steps:

[0041] S1. Using a three-section function model to represent the field neutron spectrum

[0042] The reactor nuclear fission mainly produces fast neutrons, which are consistent with the fission spectrum distribution. When the fast neutrons transport in the shielding material, a part of them is scattered and moderated. With the increase of the thickness of the shielding material, the fraction of the moderated neutrons increases, and the fraction of the fast neutrons decreases. Part of the moderated neutrons will be captured by the surrounding matter, resulting in a decrease in the total neutron flux. After multiple scattering, the neutrons and the surrounding matter reach thermal equilibrium, and their energy basically no longer changes, thereby forming a thermal neutron peak. The scattered neutrons that do not reach thermal equilibrium are between fast neutrons and thermal neutrons.

[0043] According to ISO12789:1-2008, the field neutron spectrum can be divided into a thermal energy region, an intermediate energy region and a fast energy region. The field neutron spectrum can be described by formula (1):

[0044] (1)

[0045] In the formula, , and are the neutron spectra of the thermal energy region, the intermediate energy region and the fast energy region respectively, is the demarcation energy point of the thermal energy region and the intermediate energy region, is the demarcation energy point of the intermediate energy region and the fast energy region. Formula (2), formula (3) and formula (4) are respectively , and a form of expression.

[0046] (2)

[0047] (3)

[0048] (4)

[0049] In the formula, A and B are respectively and The amount of flux characterizes the amplitude of the function; a and b are the energies corresponding to the temperature, characterizing the spectral shape of the function; and are the intercept and slope of the log-linear function, respectively; and Depend on and respectively and The function value at is calculated as (5) and Equation 6) to reduce the number of parameters:

[0050] ⑸

[0051] ⑹

[0052] S2. Use shielding materials of different thicknesses to construct simulated on-site neutron energy spectra at various moderation levels, and use detectors to measure and obtain the fast neutron pulse count rate. and thermal neutron pulse count rate And use formula ⑺ to calculate the thermal ratio r:

[0053] ⑺

[0054] The moderation level of the shielding material affects and The thermal neutron pulse counts are primarily generated by neutrons in the thermal energy region, while the fast neutron pulse counts are generated only by neutrons in the fast energy region. Therefore, r can represent the moderation level of the shielding material.

[0055] Simulating the on-site neutron spectrum can be achieved by 252 Neutron sources with isotopes such as Cf or accelerator monoenergetic neutron sources are constructed in combination with moderator materials, and their energy spectrum distribution can be calculated through Monte Carlo simulation.

[0056] The fast neutron pulse count rate and thermal neutron pulse count rate can be measured by fast neutron detectors (such as organic scintillators, hydrogen-filled proportional counters or diamond detectors) and thermal neutron detectors (such as 3 He proportional counter tube, 10 BF3 proportional counter tube or 6 LiI scintillator, etc.) can be measured separately, or the 10 B (or 6 Li, etc.) or composite detectors (such as composite scintillators) are measured together.

[0057] In this embodiment, use 252 Cf source, lead bricks and polyethylene plates were used to construct a simulated on-site neutron spectrum. 252Cf source releases neutrons by spontaneous fission, which has a similar neutron spectrum to that of a nuclear reactor. To simulate the moderating effect of shielding materials and the slowing down of polyethylene plates, a lead brick was placed behind the Cf source, and polyethylene plates of different thicknesses were used to build up a variety of moderated neutron spectra. A composite scintillator was placed in the radiation field to measure the fast neutron pulse rate 252 and the thermal neutron pulse rate for each thickness of polyethylene plate, and r was calculated.

[0058] S3. Simulate the test point's simulated field neutron spectrum by Monte Carlo model simulation and fit it with equation (1) to obtain the model parameters A, a, B and b in .

[0059] The moderation level is the main factor affecting the field neutron spectrum. When the thickness of the shielding material increases, the moderation level of the shielding material for neutrons increases, the total neutron fluence of the field decreases, the probability of the source term neutrons not interacting decreases, the share of the fast energy region decreases, and the share of the thermal energy region increases. Therefore, different moderation levels correspond to different model parameters.

[0060] In this embodiment, a Monte Carlo model of the entire laboratory is established, the test point's simulated field neutron spectrum is simulated by Monte Carlo simulation, and equation (1) is fitted to obtain the model parameters A, a, B and b for each thickness of polyethylene plate.

[0061] S4. Establish the correspondence between the model parameters A, a, B and b and r

[0062] The spectrum parameters a and b are determined by the moderation level, so the correspondence between a and b and r can be established directly. The fluence parameters A and B are affected by the moderation level and also related to the source term intensity. Therefore, first, the ratio of the fluence parameter A to the thermal neutron pulse rate , and the ratio of the fluence parameter B to the fast neutron pulse rate are calculated, and then the correspondence between and and r is established.

[0063] In this embodiment, the correspondence between the model parameters A, a, B and b and r is established, and the results are shown in Figure 2 . As can be seen from Figure 2 , a, b, and have a clear correspondence with r.

[0064] S5. Measure the fast neutron pulse rate and the thermal neutron pulse rate​​ According to the corresponding relationship between the model parameters A, a, B and b and r, the values of A, a, B and b are determined by table lookup or interpolation.

[0065] Firstly, the model parameters A, a, B and b are determined according to the corresponding relationship between the model parameters A, a, B and b and r. and r is calculated according to formula (2), and then the values of a and b are determined according to the corresponding relationship between a and b and r; then the values of A and B are determined according to the corresponding relationship between A and B and r, and the values of a and b are determined according to the corresponding relationship between a and b and r, then A= and and and . , B= .

[0066] In this embodiment, the simulated in-situ neutron energy spectrum constructed in step S2 is taken as the in-situ neutron energy spectrum, and the values of A, a, B and b are determined by cubic spline interpolation according to the corresponding relationship between the model parameters and r, according to the fast neutron pulse count rate and the thermal neutron pulse count rate .

[0067] S6. According to the model parameters A, a, B and b, the in-situ neutron energy spectrum is calculated according to formula (1), and then the in-situ neutron dose H is calculated according to formula (3) by using the in-situ neutron energy spectrum and the neutron fluence-dose conversion coefficient .

[0068] (3)

[0069] In the formula, c is a calibration factor for correcting the overall deviation of . Different conversion coefficients can be selected to calculate different neutron doses, such as peripheral dose equivalent, effective dose and peripheral dose. The unit of can be fluence rate or fluence. The former is calculated in real time according to the fast neutron pulse count rate and the thermal neutron pulse count rate, and at this time formula (3) calculates the in-situ neutron dose rate; the latter is calculated according to the total count of fast neutron pulses and thermal neutron pulses, and at this time formula (3) calculates the in-situ neutron cumulative dose.

[0070] In this embodiment, the in-situ neutron energy spectrum is calculated according to formula (1) according to the model parameters A, a, B and b, and then the peripheral dose equivalent is calculated according to formula (3) by using the in-situ neutron energy spectrum and the neutron fluence-peripheral dose equivalent conversion coefficient Figure 3 ​​As shown, the measured values are in good agreement with the theoretical values, proving the feasibility of the method.

[0071] The above merely describes the preferred embodiments of the present application, and it should be noted that those skilled in the art can make several improvements and modifications without departing from the technical principles of the present application, and these improvements and modifications should also be considered as falling within the protection scope of the present application.

Claims

1. A parametric real-time measurement method for on-site neutron spectrum and dose, characterized in that: The parameterized real-time measurement method comprises the following steps: S1. Using a three-segment function model to represent the on-site neutron energy spectrum On-site neutron energy spectrum is divided into thermal energy zone, medium energy zone and fast energy zone. Use formula (1) to describe: Where, 、 and These are the neutron energy spectra in the thermal energy region, the medium energy region, and the fast energy region, It is the boundary energy point between the thermal energy zone and the medium energy zone. It is the dividing energy point between the medium energy zone and the fast energy zone; 、 and Respectively as formula (2) to (4): ⑵ ⑶ ⑷ Where, A and B They are and The fluence, characterizing the amplitude of the function; a and b are the energies corresponding to the temperatures, representing the spectrum of the function; and are the intercept and slope of the log-linear function, respectively; and Depend on and respectively and The function value at is calculated as follows: ⑸ ⑹ S2. Use shielding materials of different thicknesses to construct simulated on-site neutron energy spectra at various moderation levels, and use detectors to measure and obtain the fast neutron pulse count rate. and thermal neutron pulse count rate and use formula ⑺ to calculate the thermal fast ratio r : ⑺ S3. Calculate the simulated on-site neutron spectrum of the test point through Monte Carlo simulation , and fit it with formula (1) to obtain Model parameters in A 、 a 、 B and b ; S4. Establish model parameters A 、 a 、 B and b and r The corresponding relationship S5. Measuring the fast neutron pulse count rate using a detector on site and thermal neutron pulse count rate , according to the model parameters A 、 a 、 B and b and r The corresponding relationship is determined by looking up the table or interpolation A 、 a 、 B and b The value of S6. According to the model parameters A 、 a 、 B and b , calculate the on-site neutron spectrum according to the formula , and then using on-site neutron spectroscopy and neutron fluence-dose conversion factor Calculate the on-site neutron dose according to formula ⑻ H : ⑻ Where, c is the calibration factor used to correct Different conversion factors are used to calculate different neutron doses. ; The unit is fluence rate or fluence. When fluence rate is used, it is calculated in real time based on the fast neutron pulse count rate and the thermal neutron pulse count rate. Formula (8) calculates the on-site neutron dose rate; when fluence is used, it is calculated based on the total fast neutron pulse count and the total thermal neutron pulse count. At this time, Formula (8) calculates the on-site neutron cumulative dose.

2. The parametric real-time measurement method for on-site neutron spectrum and dose according to claim 1, characterized in that: In step S2, the simulated on-site neutron spectrum is obtained by 252 A Cf isotope neutron source or an accelerator monoenergetic neutron source is constructed in conjunction with a moderator material, and the energy spectrum distribution is obtained through experimental measurement or Monte Carlo simulation calculation.

3. The parametric real-time measurement method for on-site neutron spectrum and dose according to claim 1, characterized in that: In step S2, the fast neutron pulse count rate and the thermal neutron pulse count rate are measured by a fast neutron detector and a thermal neutron detector respectively, or by using a detector containing 10 B. 6 Li detector or composite detector is measured together.

4. The parametric real-time measurement method for on-site neutron spectrum and dose according to claim 1, characterized in that: In step S4, the injection parameters are first calculated A Thermal neutron pulse count rate Ratio , and the fluence parameters B Fast neutron pulse count rate Ratio , then build and and r The corresponding relationship.

5. The parametric real-time measurement method for on-site neutron spectrum and dose according to claim 4, characterized in that: In step S5, first use and Calculate according to formula ⑺ r , then according to a and b and r The corresponding relationship is determined a and b The value of and and r The corresponding relationship is determined and The value of A = × , B = × .

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