Method for measuring spectrum and accumulated dose of neutrons by utilizing passive detector

A technique of accumulating dose and detector, which is applied in the field of measuring neutron energy spectrum and accumulating dose by using passive detectors, which can solve the problem that the measurement results of unknown energy neutrons have large uncertainty and cannot obtain incident neutron energy information, and the measurement method It can achieve the effect of light weight, strong adaptability and convenient use

Active Publication Date: 2013-02-13
CHINA INST FOR RADIATION PROTECTION
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AI-Extracted Technical Summary

Problems solved by technology

[0008] However, although the existing active neutron dose measurement methods and instruments (such as the energy spectrum type single-sphere multi-detector neutron measurement system developed by the China Radiation Institute) can obtain incident neutron energy information, they lack power supply Or in the case of some serious nuclear accidents, this measurement method and instrument can not or is difficult t...
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Abstract

The invention relates to a method for measuring the spectrum and accumulated dose of neutrons by utilizing a passive detector. According to the method, a moderator probe is used for measuring a neutron dose in a radiation field; and when data are processed, the energy of incident neutrons is calculated on the basis of different moderation distances experienced by the neutrons incident to each detector, a count measured by each detector is converted into the accumulated fluence of the neutrons, spectrum unscrambling software is used for obtaining the actual fluence spectrum of the incident neutrons, and the accumulated dose of the neutrons is calculated according to the actual fluence of the incident neutrons. According to the measurement method, the supply of power is not required during measurement; the method is high in environmental adaptability, low in cost and convenient to use; and in addition, the energy information of the neutrons can be acquired by the method, uncertainty about measurement results is lower than that of the conventional passive neutron measurement method, and the method can be widely applied to the neutron dose monitoring work of nuclear accident emergencies and environments.

Application Domain

Neutron radiation measurement

Technology Topic

PhysicsRadiation field +7

Image

  • Method for measuring spectrum and accumulated dose of neutrons by utilizing passive detector
  • Method for measuring spectrum and accumulated dose of neutrons by utilizing passive detector
  • Method for measuring spectrum and accumulated dose of neutrons by utilizing passive detector

Examples

  • Experimental program(1)

Example Embodiment

[0024] The present invention will be further described below in conjunction with the drawings and specific embodiments.
[0025] Such as figure 1 As shown, the moderator probe used in the measurement method of the present invention includes a moderator sphere 2, and the detector is installed in the moderator sphere 2. Specifically, the detector is first encapsulated in the stainless steel tube 1, and then the three stainless steel tubes 1 with the detector are inserted into the moderating sphere 2 in a two-by-two vertical manner to obtain the moderation used in the measurement method of the present invention. Body probe.
[0026] The detector used in the measurement method of the present invention can be a thermoluminescence (TLD) detector, a bubble chamber detector or a CR-39 solid nuclear track detector. The most widely used LiF (Mg, Cu, P) thermal The luminescence detector is taken as an example to illustrate the method for measuring neutron energy spectrum and cumulative dose provided by the present invention, where LiF (Mg, Cu, P) is abbreviated as LiF.
[0027] This embodiment adopts 6 LiF and 7 Each piece of LiF forms a detector pair to deduct the γ background, which is determined by 6 LiF measures the total fluence of gamma and neutrons, which is determined by 7 LiF measures the fluence caused by γ, and subtracts the two to obtain the cumulative fluence of neutrons; since the thickness of the moderated sphere covered by each detector pair is different, the neutron transport program (such as MCNP, etc.) can be used to calculate the neutron fluence. The fluence energy spectrum of the son. Since bubble chamber detectors and nuclear track detectors are not sensitive to gamma rays, they can be used as detectors without deduction of gamma background.
[0028] Such as figure 2 As shown, the detectors 3 are packaged in the stainless steel tube 1 at equal intervals, and the specific number of the detector pairs can be adjusted according to the diameter of the moderating sphere 2.
[0029] When using the method provided by the present invention to measure the neutron energy spectrum and the cumulative dose, the figure 1 The moderator probe shown is placed in the radiation field to be measured. After a measurement period, the probe is retrieved and read, and then the data is processed according to the following method provided by the present invention:
[0030] First of all, the moderation distance (that is, the thickness of the neutron through the moderating sphere) experienced by the incident neutron before hitting the detector 3 is different, such as image 3 As shown, the moderation distance of a certain neutron is L1, and the moderation distance of another neutron is L2. According to the theory of interaction between neutrons and matter, the corresponding incident neutron energy can be calculated.
[0031] For a certain neutron field, the measurement result can be described by the following equation:
[0032] A i = ∫ 0 ∞ R i ( E ) Φ ( E ) dE - ϵ i , i = 1 , . . . . . . , n - - - ( 1 )
[0033] n is the number of detectors, A i Is the count (rate) of the i-th detector (unit: s -1 ), R i (E) is the fluence energy response function of the i-th detector (unit: cm 2 ), Φ(E) is the fluence (rate) of neutron energy E (unit is cm -2 s -1 ), ε i Is the measurement uncertainty of the i-th detector.
[0034] From a limited number of measurements A i It is impossible to determine the unique continuous function Φ(E). To solve this problem, a physically relevant energy spectrum can only be obtained by solving the spectrum, that is, the neutrons in the energy range are discretized into small energy intervals (box ), using the discrete function Φ j (E) Replace the continuous function Φ(E).
[0035] A i +ε i =∑R ij Φ j (E)j=1,...,m (2)
[0036] In equations (1) and (2), the known number, that is, the number of multi-sphere spectrometer detectors (n), usually does not exceed 20, while the unknown number, that is, the number of neutron energy boxes (m), is relatively large. Therefore, equations (1) and (2) are actually a kind of ill-conditioned equations, that is, the unknown number is greater than the number of equations, and the solution methods are divided into three categories, namely iterative method, direct method and Monka method. The Monte Carlo method, or Monte Carlo method, is widely used in nuclear physics. It is based on the specific environment and experimental conditions of the experimental site, based on the principle of interaction between rays and matter, and using computer programs to simulate the actual situation under physical experimental conditions. The required amount. The present invention uses the existing Monte Carlo calculation software MCNP-4B to solve the above problems.
[0037] followed by 6 LiF and 7 The counts measured by the LiF detector are converted into cumulative fluence of neutrons, and the actual fluence energy spectrum of incident neutrons is obtained by using the despectrum software.
[0038] Finally, according to the actual fluence energy spectrum of incident neutrons, the cumulative neutron dose calculation is completed by the computer.
[0039] In addition, since the measured counts of each detector are converted into neutron fluence, and this fluence is distributed according to energy, the value of the neutron fluence is multiplied by the fluence-dose conversion of the corresponding energy The coefficient (given in Publication No. 74 of the International Radiological Commission) can be used to obtain the required neutron cumulative dose equivalent. Calculated as follows:
[0040] E=H p (d)=∑Φ(E)*k(E) (3)
[0041] In the formula, E is the cumulative neutron dose equivalent (unit: pSv), and Φ(E) is the cumulative neutron fluence (unit: cm -2 ), k(E) is the neutron fluence dose conversion coefficient distributed with energy (unit: pSv*cm -2 ).
[0042] According to the method of the present invention, the cumulative neutron fluence, the neutron cumulative dose equivalent, the actual fluence energy spectrum of the possible incident neutrons, and the neutron cumulative dose obtained by de-spectrum are not more than the reference data. ±30%, indicating that the method of the present invention improves the uncertainty of the measurement result.
[0043] In the measurement method of the present invention, the moderator probe used uses spherical, 20-25cm in diameter, hydrogen-rich polyethylene material as the moderator to slow down fast neutrons, and other moderator materials can also be used , It can also achieve the purpose of the present invention.
[0044] Using the measurement method of the present invention, the passive detector has the advantages of passiveness, strong environmental adaptability, low price, convenient use, etc., and can obtain neutron energy information, and the uncertainty of the measurement result is much better than that of conventional The passive neutron measurement method has broad application prospects in the field of nuclear accident emergency and environmental monitoring.
[0045] The method of the present invention is not limited to the examples described in the specific implementation manners, and those skilled in the art can obtain other implementation manners based on the technical solutions of the present invention, which also belong to the technical innovation scope of the present invention.
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