On-site neutron energy spectrum and dose parameterized real-time measurement method

Abstract

The invention provides an on-site neutron energy spectrum and dose parameterization real-time measurement method. The method comprises the following steps: representing an on-site neutron energy spectrum by using a three-section type function model; constructing simulated field neutron energy spectrums of various moderated levels, measuring by using a detector, and calculating a thermal fast ratio; simulating and calculating a simulated field neutron energy spectrum of the test point through a Monte Carlo model, and obtaining model parameters; establishing a corresponding relation between the model parameters and the thermal fast ratio; a detector is used for measuring the fast neutron pulse counting rate and the thermal neutron pulse counting rate on site, and according to the corresponding relation between the model parameters and the thermal fast ratio, the values of the model parameters are determined in a table look-up or interpolation mode; and calculating a field neutron energy spectrum according to the model parameters, and calculating the field neutron dose by using the field neutron energy spectrum and the neutron fluence-dose conversion coefficient.

Description

Technical Field

[0001] The invention belongs to the technical field of nuclear radiation monitoring, and in particular relates to a parametric real-time measurement method for on-site neutron energy spectrum and dose. Background Art

[0002] Affected by the nuclear fission chain reaction, the reactor will produce a large number of neutrons and gamma rays. Measuring the neutron and gamma ray radiation levels in the environment around the reactor is of great significance for monitoring the operating status of the reactor. The radiation level measurement around the reactor includes fixed monitoring and patrol measurement in terms of methods, and includes dose measurement and energy spectrum measurement in terms of content. Fixed monitoring has lower requirements on the volume and weight of the instrument, while patrol measurement requires the instrument to be easy to carry. Dose is the core parameter for evaluating radiation levels, and energy spectrum can be used to identify the type of radioactive nuclides and analyze the cause and scale of nuclear leakage.

[0003] Both neutrons and gamma rays are indirect ionizing radiation, and their mechanisms and types of interaction with matter are different. The commonly used detector sensitive materials are also different. Therefore, different instruments are usually used to measure neutrons and gamma rays respectively. At present, there are many types of gamma ray measurement instruments with relatively mature technology, such as high-pressure ionization chambers for fixed monitoring, portable gamma radiation meters and portable gamma spectrometers for patrol, etc. However, neutron measurement instruments are relatively scarce, mainly Rem meters for measuring neutron ambient dose equivalent, time-of-flight spectrometers, organic scintillators and 3 Neutron spectrum measurement instruments such as He proportional counters are affected by factors such as method principles, spectrum inversion and detection efficiency, and are not suitable for on-site neutron spectrum measurement. In addition, due to the presence of polyethylene moderators, the size and weight of the Rem instrument are much larger than portable gamma radiation instruments, which is not affected by fixed monitoring, but difficult to be used in a portable manner. It can be seen that the miniaturization and integrated measurement of on-site neutron spectrum and dose has not yet been solved. Summary of the invention

[0004] 1. Technical issues to be resolved

[0005] The present invention proposes a parametric real-time measurement method for on-site neutron energy spectrum and dose, so as to solve the technical problem of realizing portable integrated real-time measurement of on-site neutron energy spectrum and dose.

[0006] (II) Technical solution

[0007] In order to solve the above technical problems, the present invention proposes a parametric real-time measurement method for on-site neutron energy spectrum and dose, which comprises the following steps:

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

[0009] The on-site neutron energy spectrum is divided into thermal energy zone, medium energy zone and fast energy zone. The on-site neutron energy spectrum Φ(E) is described by formula (1):

[0010]

[0011] In the formula, Φ t (E), Φ m (E) and Φ f (E) are the neutron energy spectra in the thermal energy region, the medium energy region and the fast energy region, respectively. 1 is the boundary energy point between the thermal energy zone and the medium energy zone, E 2 It is the boundary energy point between the medium energy zone and the fast energy zone; Φ t (E), Φ m (E) and Φ f (E) are respectively as formula (2) to (4):

[0012]

[0013] In the formula, A and B are Φ t (E) and Φ f (E) is the flux, which characterizes the amplitude of the function; a and b are the energies corresponding to the temperature, which characterize the spectral shape of the function; k 1 and k 2 are the intercept and slope of the log-linear function respectively; k 1 and k 2 By Φ t (E) and Φ f (E) respectively in E 1 and E 2 The function value at is calculated as (5) and (6):

[0014]

[0015] k 1 =Ε 2 Φ f (E 2 )-k 2 logE 2 (6)

[0016] S2. Use shielding materials of different thicknesses to construct simulated on-site neutron energy spectra at various slowing-down levels, and use detectors to measure and obtain the fast neutron pulse count rate η f and thermal neutron pulse count rate η t , and use formula (7) to calculate the thermal speed ratio r:

[0017]

[0018] S3. Calculate the simulated on-site neutron energy spectrum Φ(E) of the test point by Monte Carlo model simulation, and fit it with formula (1) to obtain the model parameters A, a, B and b in Φ(E);

[0019] S4. Establish the corresponding relationship between model parameters A, a, B and b and r

[0020] S5. Measure the fast neutron pulse count rate η using a detector on site f and thermal neutron pulse count rate η t , according to the corresponding relationship between model parameters A, a, B and b and r, the values ​​of A, a, B and b are determined by table lookup or interpolation;

[0021] S6. According to the model parameters A, a, B and b, the on-site neutron energy spectrum Φ(E) is calculated according to formula (1), and then the on-site neutron energy spectrum Φ(E) and the neutron injection-dose conversion coefficient h(E) are used to calculate the on-site neutron dose H according to formula (8):

[0022] H=c∫Φ(E)h(E)dE (8)

[0023] Where c is the calibration factor, which is used to correct the overall deviation of H. Different conversion coefficients h(E) are selected to calculate different neutron doses. The unit of Φ(E) 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.

[0024] 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 moderating material, and the energy spectrum distribution is obtained through experimental measurement or Monte Carlo simulation calculation.

[0025] Further, 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 used for measurement.

[0026] Further, in step S4, the fluence parameter A and the thermal neutron pulse count rate η are first calculated. t The ratio v A , and the fluence parameter B and the fast neutron pulse count rate η f The ratio v B , then build v A and v B The corresponding relationship with r.

[0027] Further, in step S5, firstly use n f and η t Calculate r according to formula (7), and then determine the values ​​of a and b according to the corresponding relationship between a and b and r; then determine the values ​​of a and b according to v A and v B The corresponding relationship with r determines v A and v B The value of , then A=v A ×η t , B = v B ×η f .

[0028] (III) Beneficial effects

[0029] The present invention proposes a parameterized real-time measurement method for on-site neutron energy spectrum and dose, which comprises the following steps: using a three-segment function model to represent the on-site neutron energy spectrum; constructing simulated on-site neutron energy spectra of multiple moderation levels, measuring them using a detector, and calculating a thermal-fast ratio; obtaining model parameters by simulating and calculating the simulated on-site neutron energy spectrum of a test point using a Monte Carlo model; establishing a corresponding relationship between the model parameters and the thermal-fast ratio; measuring a fast neutron pulse count rate and a thermal neutron pulse count rate using a detector on-site, and determining the values ​​of the model parameters by table lookup or interpolation according to the corresponding relationship between the model parameters and the thermal-fast ratio; calculating the on-site neutron energy spectrum according to the model parameters, and then calculating the on-site neutron dose by using the on-site neutron energy spectrum and a neutron injection-dose conversion coefficient.

[0030] Compared with the prior art, the present invention has the following beneficial effects:

[0031] 1. The present invention can realize the measurement of on-site neutron energy spectrum only by using the pulse counting rate, without the need for spectrum decomposition calculation, with small amount of calculation and low difficulty of use. It is significantly superior to neutron spectrometers such as organic scintillator and 3He proportional counter tube, and can realize online real-time measurement through a single-chip microcomputer or FPGA.

[0032] 2. The method proposed in the present invention can realize the integrated real-time measurement of neutron energy spectrum and dose on site, without the need for polyethylene moderator, and is significantly superior to the Rem meter in terms of volume, weight and types of measurement parameters, and is more suitable for on-site use around reactors.

[0033] 3. The method proposed in the present invention uses the energy spectrum to calculate the on-site neutron dose, does not involve the energy response problem, can integrate multiple conversion factors to calculate multiple doses at the same time, and when changing the practical amount of neutron dose, only the neutron-dose conversion factor needs to be replaced without changing the instrument structure. BRIEF DESCRIPTION OF THE DRAWINGS

[0034] Figure 1 Comparison between neutron simulation spectrum and fitting spectrum under different polyethylene thickness;

[0035] Figure 2 is the variation of model parameters with heat-fast ratio;

[0036] Figure 3 The measured value of neutron ambient dose equivalent rate, the theoretical value and the relative deviation vary with the thickness of polyethylene. DETAILED DESCRIPTION

[0037] In order to make the purpose, content and advantages of the present invention more clear, the specific implementation methods of the present invention are further described in detail below in conjunction with the drawings and examples.

[0038] This embodiment provides a parametric real-time measurement method for on-site neutron spectrum and dose, which specifically includes the following steps:

[0039] S1. Using a three-segment function model to represent the on-site neutron energy spectrum

[0040] Reactor nuclear fission mainly produces fast neutrons, which conforms to the fission spectrum distribution. When fast neutrons are transported in the shielding material, a part of them is scattered and slowed down. As the thickness of the shielding material increases, the proportion of slowed neutrons increases, and the proportion of fast neutrons decreases accordingly. Some slowed neutrons will be captured by the surrounding matter, resulting in a decrease in the total neutron injection. After multiple scatterings, the neutrons reach thermal equilibrium with the surrounding matter, and their energy basically no longer changes, thus forming a thermal neutron peak. Scattered neutrons that have not reached thermal equilibrium are between fast neutrons and thermal neutrons.

[0041] According to ISO12789:1-2008, the on-site neutron energy spectrum can be divided into thermal energy zone, medium energy zone and fast energy zone. The on-site neutron energy spectrum Φ(E) can be described by formula (1):

[0042]

[0043] In the formula, Φ t (E), Φ m (E) and Φ f (E) are the neutron energy spectra in the thermal energy region, the medium energy region and the fast energy region, respectively. 1 is the boundary energy point between the thermal energy zone and the medium energy zone, E 2 is the boundary energy point between the medium energy zone and the fast energy zone. Equations (2), (3) and (4) are respectively Φ t (E), Φ m (E) and Φ f (E) A form of expression.

[0044]

[0045] In the formula, A and B are Φ t (E) and Φ f(E) is the flux, which characterizes the amplitude of the function; a and b are the energies corresponding to the temperature, which characterize the spectral shape of the function; k 1 and k 2 are the intercept and slope of the log-linear function respectively; k 1 and k 2 By Φ t (E) and Φ f (E) respectively in E 1 and E 2 The function value at is calculated as (5) and (6) to reduce the number of parameters:

[0046]

[0047] k 1 =Ε 2 Φ f (E 2 )-k 2 logE 2 (6)

[0048] S2. Use shielding materials of different thicknesses to construct simulated on-site neutron energy spectra at various slowing-down levels, and use detectors to measure and obtain the fast neutron pulse count rate η f and thermal neutron pulse count rate η t , and use formula (7) to calculate the thermal speed ratio r:

[0049]

[0050] The slowing down level of the shielding material will affect Φ t (E) and Φ f (E) is the ratio of the amount of radiation. The thermal neutron pulse count is mainly produced by neutrons in the thermal energy region, while the fast neutron pulse count is only produced by neutrons in the fast energy region. Therefore, r can represent the moderation level of the shielding material.

[0051] Simulating the on-site neutron spectrum can be done by 252 Isotope neutron sources such as Cf or accelerator monoenergetic neutron sources are constructed in conjunction with moderating materials, and their energy spectrum distribution can be calculated through Monte Carlo simulation.

[0052] 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 counter tubes or diamond detectors) and thermal neutron detectors (such as 3 He proportional counter tube, 10 BF 3 Proportional counter 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.

[0053] In this embodiment, use 252 Cf source, lead bricks and polyethylene plates were used to construct a simulated on-site neutron spectrum. 252 The Cf source releases neutrons through spontaneous fission, which is similar to the neutron spectrum of reactor nuclear fission. In order to simulate the shielding material's slowing absorption effect on gamma rays and polyethylene sheets, 252 A lead brick was placed behind the Cf source, followed by polyethylene sheets of different thicknesses to construct simulated in-situ neutron energy spectra at various slowing-down levels. A composite scintillator was placed in the radiation field, and the fast neutron pulse count rate η was measured for each polyethylene sheet thickness. f and thermal neutron pulse count rate η t , and calculate r.

[0054] S3. The simulated on-site neutron energy spectrum Φ(E) of the test point is calculated by Monte Carlo model simulation, and is fitted using equation (1) to obtain the model parameters A, a, B and b in Φ(E).

[0055] The moderation level is the main factor affecting the on-site neutron energy spectrum. When the thickness of the shielding material increases, the moderation level of the shielding material for neutrons increases, the total on-site neutron injection decreases, the probability of the source term neutrons not taking effect decreases, the share of the fast energy zone decreases, and the share of the thermal energy zone increases. Therefore, different moderation levels correspond to different model parameters.

[0056] In this embodiment, a Monte Carlo model of the entire laboratory is established, and the simulated on-site neutron energy spectrum of the test point is calculated by Monte Carlo simulation, and the model parameters A, a, B and b for each polyethylene plate thickness are obtained by fitting using formula (1).

[0057] S4. Establish the corresponding relationship between model parameters A, a, B and b and r

[0058] The spectral shape parameters a and b are determined by the moderation level, so the corresponding relationship between a and b and r can be directly established. In addition to being affected by the moderation level, the fluence parameters A and B are also related to the source term intensity. Therefore, we first calculate the fluence parameter A and the thermal neutron pulse count rate η t The ratio v A , and the fluence parameter B and the fast neutron pulse count rate η f The ratio v B , then build v A and v B The corresponding relationship with r.

[0059] In this embodiment, the corresponding relationship between the model parameters A, a, B and b and r is established, and the result is as follows Figure 2 As shown. Figure 2 It can be seen that a, b, v A and v BThere is a clear correspondence with r.

[0060] S5. Measure the fast neutron pulse count rate η using a detector on site f and thermal neutron pulse count rate η t , 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.

[0061] First, using η f and η t Calculate r according to formula (7), and then determine the values ​​of a and b according to the corresponding relationship between a and b and r; then determine the values ​​of a and b according to v A and v B The corresponding relationship with r determines v A and v B The value of , then A=v A ×η t , B = v B ×η f .

[0062] In this embodiment, the simulated on-site neutron energy spectrum constructed in step S2 is used as the on-site neutron energy spectrum, and the fast neutron pulse count rate η is used as the on-site neutron energy spectrum. f and thermal neutron pulse count rate η t , using the corresponding relationship between the model parameters and r, the values ​​of A, a, B and b are determined by cubic spline interpolation.

[0063] S6. According to the model parameters A, a, B and b, the on-site neutron energy spectrum Φ(E) is calculated according to formula (1), and then the on-site neutron energy spectrum Φ(E) and the neutron injection-dose conversion coefficient h(E) are used to calculate the on-site neutron dose H according to formula (8):

[0064] H=c∫Φ(E)h(E)dE (8)

[0065] Where c is the calibration factor, which is used to correct the overall deviation of H. Different conversion factors h(E) can be selected to calculate different neutron doses, such as ambient dose equivalent, effective dose and ambient dose. The unit of Φ(E) can be either fluence rate or fluence. The former is calculated in real time based on the fast neutron pulse count rate and the thermal neutron pulse count rate. In this case, formula (8) calculates the on-site neutron dose rate; the latter is calculated based on the total fast neutron pulse count and the total thermal neutron pulse count. In this case, formula (8) calculates the on-site neutron cumulative dose.

[0066] In this embodiment, the on-site neutron energy spectrum Φ(E) is calculated according to the model parameters A, a, B and b according to formula (1), and then the on-site neutron energy spectrum Φ(E) and the neutron injection-ambient dose equivalent conversion coefficient h are used. *(E), the ambient dose equivalent is calculated according to formula (8) and compared with the theoretical value calculated by Monte Carlo simulation spectrum. The results are as follows Figure 3 As shown, it can be seen that the measured values ​​are in good agreement with the theoretical values, proving the feasibility of this method.

[0067] The above is only a preferred embodiment of the present invention. It should be pointed out that for ordinary technicians in this technical field, several improvements and modifications can be made without departing from the technical principles of the present invention. These improvements and modifications should also be regarded as the scope of protection of the present invention.

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 The on-site neutron energy spectrum is divided into thermal energy zone, medium energy zone and fast energy zone. The on-site neutron energy spectrum Φ(E) is described by formula (1): In the formula, Φ t (E), Φ m (E) and Φ f (E) are the neutron energy spectra of the thermal energy zone, the intermediate energy zone and the fast energy zone, respectively; E1 is the boundary energy point between the thermal energy zone and the intermediate energy zone, and E2 is the boundary energy point between the intermediate energy zone and the fast energy zone; Φ t (E), Φ m (E) and Φ f (E) are respectively as formula (2) to (4): In the formula, A and B are Φ t (E) and Φ f (E) is the flux, which characterizes the amplitude of the function; a and b are the energies corresponding to the temperature, which characterize the spectral shape of the function; k1 and k2 are the intercept and slope of the logarithmic linear function, respectively; k1 and k2 are given by Φ t (E) and Φ f (E) The function values ​​at E1 and E2 are calculated as (5) and (6): k1=Ε2Φ f (E2)-k2 logE2 (6) S2. Use shielding materials of different thicknesses to construct simulated on-site neutron energy spectra at various slowing-down levels, and use detectors to measure and obtain the fast neutron pulse count rate η f and thermal neutron pulse count rate η t , and use formula (7) to calculate the thermal speed ratio r: S3. Calculate the simulated on-site neutron energy spectrum Φ(E) of the test point by Monte Carlo model simulation, and fit it with formula (1) to obtain the model parameters A, a, B and b in Φ(E); S4. Establish the corresponding relationship between model parameters A, a, B and b and r S5. Measure the fast neutron pulse count rate η using a detector on site f and thermal neutron pulse count rate η t , according to the corresponding relationship between model parameters A, a, B and b and r, the values ​​of A, a, B and b are determined by table lookup or interpolation; S6. According to the model parameters A, a, B and b, the on-site neutron energy spectrum Φ(E) is calculated according to formula (1), and then the on-site neutron energy spectrum Φ(E) and the neutron injection-dose conversion coefficient h(E) are used to calculate the on-site neutron dose H according to formula (8): H=c∫Φ(E)h(E)dE (8) Where c is the calibration factor, which is used to correct the overall deviation of H. Different conversion coefficients h(E) are selected to calculate different neutron doses. The unit of Φ(E) 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 of on-site neutron spectrum and dose according to claim 1, characterized in that: In step S2, the simulated on-site neutron spectrum is 252 A Cf isotope neutron source or an accelerator monoenergetic neutron source is constructed in conjunction with a moderating material, and the energy spectrum distribution is obtained through experimental measurement or Monte Carlo simulation calculation.

3. The parametric real-time measurement method of 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 used for measurement.

4. The parametric real-time measurement method of on-site neutron energy spectrum and dose according to claim 1, characterized in that: In step S4, the fluence parameter A and the thermal neutron pulse count rate η are first calculated. t The ratio v A , and the fluence parameter B and the fast neutron pulse count rate η f The ratio v B , then build v A and v B The corresponding relationship with r.

5. The parametric real-time measurement method of on-site neutron spectrum and dose according to claim 1, characterized in that: In step S5, first use n f and η t Calculate r according to formula (7), and then determine the values ​​of a and b according to the corresponding relationship between a and b and r; then determine the values ​​of a and b according to v A and v B The corresponding relationship with r determines v A and v B The value of , then A=v A ×η t , B = v B ×η f .

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