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Radiation Protection Dosimetry
, Volume Advance Article – Jan 25, 2018

7 pages

/lp/ou_press/determination-of-the-response-to-the-atmospheric-cosmic-radiation-of-a-km5ox7WLfZ

- Publisher
- Oxford University Press
- Copyright
- © The Author(s) 2018. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com
- ISSN
- 0144-8420
- eISSN
- 1742-3406
- D.O.I.
- 10.1093/rpd/ncx309
- Publisher site
- See Article on Publisher Site

Abstract A TLD-based dosimeter of polyethylene–lead–polyethylene, was developed and characterized with Monte Carlo simulations, using the MCNPX code. This passive system for the determination of the ambient dose equivalent (H*(10)) for neutrons over a wide energy range can be used for the dosimetry of neutrons from atmospheric cosmic radiation, on the ground, and onboard aircraft. A method assisted by Monte Carlo simulations that improves the calibration of fast neutron dosimeters based on moderation and thermalization of the incident fast flux and the measurement of the thermal flux by a sensor, which respond mainly to thermal neutrons, is presented in this work. The H*(10) energy response of this dosimeter was obtained from simulations for monoenergetic neutrons from 10−10 to 104 MeV. The validation of the modeling was done with irradiations for ISO standard neutron fields of 241Am–Be, 252Cf and 252Cf(D2O) at Instituto de Radioproteção e Dosimetria (IRD, Brazil) and at CERN-EU high-energy reference field (CERF). INTRODUCTION The ionizing radiation level control in the aeronautical environment and its effects on crews has become much more evident for the last 2 decades. Numerous works on occupational health, radiation protection, flight safety, among others, highlight the fact that aviation professionals are exposed to levels of radiation of the same order of magnitude that the radiation workers in medical and nuclear areas(1–12). Cosmic radiation (CR) in the atmosphere produces dose rates that increase considerably with altitude and, for aviation professionals, may exceed the annual dose limit for the public proposed by international organizations(12). Since 1990, the International Commission on Radiological Protection (ICRP) has recognized the need to control the exposure of flight professionals, such as pilots and crew considering such exposures comparable to those of nuclear workers(13) and recommending the dosimetric monitoring of crews subjected to doses >1 mSv ICRP 60. Due to the inherent characteristics of this professional activity, it is recommended that such dosimetric monitoring be carried out by means of computer programs to evaluate the doses on flights below 15 km(14) and occasional measurements should be made with active and/or passive devices(5) with the purpose of confirming the values obtained through the computational estimates. In Europe, radiological protection for crews has been regulated since 1996 by the European Council Directive 29/96/EURATOM(15), where one of the main requirements is to assess the crew dose(16). This type of control has also been implemented in other countries, such as Canada and the USA. The development of aircraft onboard radiation dosimetry methods, therefore, presents itself as a current necessity in order to ensure the fidelity of the doses assigned to aircraft crews. For the area monitoring of penetrating radiations in work environments, the operational quantity indicated by the ICRP is the ambient dose equivalent, H*(10), which is the dose equivalent produced by an expanded and aligned radiation field at a depth of 10 mm in the ICRU sphere, which is obtained by the product of the absorbed dose at the reference point times the corresponding radiation quality factor(17). The determination of H*(10) on board of aircrafts can be performed by active instruments, involving electronic systems or passive, using passive dosimeters (solid state, track dosimeters, etc.). Passive dosimeters have the advantage of not emitting electromagnetic fields that could interfere with aircraft systems, making their use much easier and safer. In this work it is presented the calibration, assisted by Monte Carlo simulation, of a fast neutron dosimeter with the extended response to the neutron spectrum of the atmospheric CR. THE IEAv NEUTRON DOSIMETER The fast neutron dosimeter used in this work consists of a set of concentric spheres of high density polyethylene and lead, as shown in Figure 1, inside which a pair of Harshaw’s thermoluminescent dosimeters, 6LiF:Mg,Ti and 7LiF:Mg,Ti (TLD-600 and TLD-700, respectively), was placed in its geometric center. The TLD-600 dosimeter, enriched in the 6Li isotope (95.6%), has higher thermal neutron sensitivity than the TLD-700 dosimeter, while both have approximately the same sensitivity to photons and other charged particles. This particularity allows the use of the subtraction technique of the TLD-700 response from the TLD-600 response, in order to discriminate the neutrons response from the photons and other directly ionizing particles response. This technique is known by dosimetric pair technique. In this case, the dosimetric pair is employed as a thermal neutron detector. Figure 1. View largeDownload slide HDPE/Pb sphere showing the material layers and the central cavity for the accommodation of thermoluminescent dosimeters. The other holes are for fixing pins. Extracted and adapted from Federico(1). Figure 1. View largeDownload slide HDPE/Pb sphere showing the material layers and the central cavity for the accommodation of thermoluminescent dosimeters. The other holes are for fixing pins. Extracted and adapted from Federico(1). This passive dosimeter, proposed by Federico(1) aims to amplify the production of thermal neutrons in the center of the sphere by means of the combined processes of spallation and thermalization of the high-energy neutrons that compose the atmospheric CR (up to 400 MeV)(18). Given an incident neutron spectrum ϕ(E), the total neutron fluence is given by the following equation: ϕTOT=∫EiEfϕ(E)dE (1) The ambient dose equivalent, H*(10), is determined through of the product of the total neutron fluence ϕTOT and the averaged fluence-to-ambient dose equivalent conversion coefficient hϕ⁎(10)_ for this spectrum: H*(10)=hϕ⁎(10)_×φTOT (2) This averaged fluence-to-ambient dose equivalent conversion coefficient for spectrum ϕ(E) is given by the following equation: hϕ⁎(10)_=∫EiEfhϕ⁎(10;E).ϕ(E)dE∫EiEfϕ(E)dE (3)where hϕ⁎(10;E) represent the fluence-to-ambient dose equivalent conversion coefficients as a function of neutron energy and Ei and Ef are the lower and upper limits, respectively, of the energy range of the spectrum ϕ(E). In this work, were used the coefficients obtained from ICRP 74(17) for energies up to 19 MeV and from Pelliccioni(19) for energies above 19 MeV, as shown in Figure 2. Figure 2. View largeDownload slide Fluence-to-ambient dose equivalent conversion coefficients (H*(10)/ϕ) as a function of neutron energy. Figure 2. View largeDownload slide Fluence-to-ambient dose equivalent conversion coefficients (H*(10)/ϕ) as a function of neutron energy. The total neutron fluence incident on the sphere ϕTOT measured by the dosimetric responses of the pair of dosemeters TLD600–TLD700, is given by the following equation: ϕTOT=F-T×FC×(M600−M700Fγ700Fγ600) (4)where, F-T is the average thermalization factor of the sphere for the incident fast neutron spectrum; FC is the conversion factor (in cm−2 pC−1) of the TL signal of the dosimetric pair (expression into brackets in pC) for thermal neutron fluence (in cm−2); M600 and M700 are the thermoluminescent signal measurements of the TLD-600 and TLD-700, respectively, obtained in the thermoluminescent dosimeters (in pC) and Fγ600 and Fγ700 are the calibration factors for gamma radiation of the TLD-600 and TLD-700, respectively (in μGy pC−1). The relationship between the fluence of neutrons with energy E incident on the sphere and the fluence of thermal neutrons produced within this arrangement is determined by means of the thermalization factor, FT(E), which is defined as follows: FT(E)=ϕ(E)ϕTH (5)where, ϕ(E) is the total neutron fluence incident at the point of interest with energy between E and E+ΔE and ϕTH is the thermal neutron fluence at the point of interest, thermalized by the dosimetric system. This factor depends on the energy of the incident neutron and thus depends on the shape of the incident energy spectrum ϕ(E), which can be calculated by the following equation: F-T=∫EiEfFT(E)ϕ(E)dE∫EiEfϕ(E)dE (6)where FT(E) is the thermalization factor for neutrons with energy E and Ei and Ef are the lower and upper limits, respectively, of the energy range of the spectrum ϕ(E). this factor can be obtained experimentally or through computer simulation, such as: F-T=ϕTOTϕTH (7) Therefore, by combining equations (2) and (4), the ambient dose equivalent can be calculated by the following equation: H*(10)=hϕ⁎(10)_.FT_.FC.(M600−M700Fγ700Fγ600)=FH*(10).MPair (8)where FH*(10) is the general calibration factor of the dosimeter, which relates the ambient dose equivalent to the dosimetric response of the thermoluminescent pair MPair (the term described into brackets). In the approach of the present work, this factor is the product of three factors related to distinct phenomenological origins and different dosimetric variables. The first factor, hϕ⁎(10)_, independent of the dosimeter, is calculated from the integration of the product of the interpolation curve of the monoenergetic factors tabulated by the normalized spectrum of the incident neutrons, as shown in equation (3). The second factor, FT, depends only on the geometry and nature of the moderator used to thermalize the incident neutrons and can be obtained through an appropriate Monte Carlo simulation, either to obtain the FT(E) point-to-point values for several neutron energy values E and applying equation (6) to the spectrum of interest, or for the mean value FT_ for the spectrum ϕ(E) in question by the ratio between the incident total neutron fluence, ϕTOT and the simulated thermal neutron fluence, ϕTH, at the point of interest (equation (7)). The third factor, FC, independent of the neutron spectrum incident on the dosimeter as well as of the dosimeter form and geometry, must be obtained by a conventional dosimetric pair calibration procedure, subjecting it to a known thermal neutron flux. Thus, the requirement for an accurate determination of the ambient dose equivalent is the best possible knowledge of the incident neutron spectrum shape on the dosimetric system. The advantage of this procedure is that it does not require several calibrations in known and certified fields for each neutron spectrum characteristic of the environment to be monitored. It requires only the calibration of the dosimetric pair with a certified thermal neutron flux and the calculation of the mean thermalization factor by integration of monoenergetic thermalization factors over the normalized neutron spectrum of interest, where the monoenergetic thermalization factors was already obtained by suitable Monte Carlo simulations. MONTE CARLO MODELING The modeling of the dosimetric arrangement and its exposure to neutron fields was performed using the transport code MCNPX 2.7(20). Firstly, were performed simulations for monoenergetic neutron spectra in the range of 1 × 10−10 to 1 × 104 MeV and were calculated FT(E) for 59 neutron energies from equation (7), in order to obtain a smooth interpolation curve in log–log scale that adequately fits a large peak in high-energy region. Then, simulations were performed for reference fields from ISO standard sources of 241Am–Be, 252Cf and 252Cf (D2O) from the Laboratório Nacional de Metrologia das Radiações Ionizantes of the Instituto de Radioproteção e Dosimetria (LMN/LNMRI/IRD) and from CERN-EU high-energy reference field (CERF/CERN)(21, 22). The simulations for monoenergetic neutrons as well for the LMN/LNMRI/IRD ISO standard sources consist of positioning the center of the dosimeter, in air, at 1 m from a point source of neutrons, which produces a conical beam directed to the dosimeter within a small solid angle, as shown in Figure 3. Figure 3. View largeDownload slide Scheme of the HDPE/Pb/HDPE sphere modeling with ISO standard sources. Figure 3. View largeDownload slide Scheme of the HDPE/Pb/HDPE sphere modeling with ISO standard sources. For the high-energy reference field (CERF/CERN), the simulation consists of positioning the center of the array in the air at a distance of 25 cm above the surface of the concrete slab from which a known flux of neutrons emerge in a broad spectrum of energy(21) as shown in Figure 4. A planar source at the superior surface of concrete roof that emits neutrons with angular distribution according to the cosine law was employed in the simulations. Figure 4. View largeDownload slide Schematization modeling of the sphere in CERF/CERN. Figure 4. View largeDownload slide Schematization modeling of the sphere in CERF/CERN. The materials composition and densities described in the simulations are presented in Table 1. The TLDs used were described as LiF with their 6Li and 7Li proportions with their dopants. Table 1. Composition and density (g cm−3) of thermoluminescent materials and dosimeters used in the simulations. HDPE Lead Air TLD600 TLD700 Density 0 .927 11.3 0.001225 2.4 2.49 1H 1.44E-01 — — — — 6Li — — — 2.30E-01 2.40E-04 7Li — — — 1.19E-02 2.69E-01 12C 8.56E-01 — — — — 14N — — 7.56E-01 — — 16O — — 2.31E-01 — — 19F — — — 7.58E-01 7.30E-01 24Mg — — — 9.48E-05 9.48E-05 25Mg — — — 1.20E-05 1.20E-05 26Mg — — — 1.32E-05 1.32E-05 40Ar — — 1.29E-02 — — 46Ti — — — 1.07E-06 1.07E-06 47Ti — — — 9.67E-07 9.67E-07 48Ti — — — 9.58E-06 9.58E-06 49Ti — — — 7.03E-07 7.03E-07 50Ti — — — 6.73E-07 6.73E-07 204Pb — 1.40E-02 — — — 205Pb — 2.41E-01 — — — 207Pb — 2.21E-01 — — — 208Pb — 5.24E-01 — — — HDPE Lead Air TLD600 TLD700 Density 0 .927 11.3 0.001225 2.4 2.49 1H 1.44E-01 — — — — 6Li — — — 2.30E-01 2.40E-04 7Li — — — 1.19E-02 2.69E-01 12C 8.56E-01 — — — — 14N — — 7.56E-01 — — 16O — — 2.31E-01 — — 19F — — — 7.58E-01 7.30E-01 24Mg — — — 9.48E-05 9.48E-05 25Mg — — — 1.20E-05 1.20E-05 26Mg — — — 1.32E-05 1.32E-05 40Ar — — 1.29E-02 — — 46Ti — — — 1.07E-06 1.07E-06 47Ti — — — 9.67E-07 9.67E-07 48Ti — — — 9.58E-06 9.58E-06 49Ti — — — 7.03E-07 7.03E-07 50Ti — — — 6.73E-07 6.73E-07 204Pb — 1.40E-02 — — — 205Pb — 2.41E-01 — — — 207Pb — 2.21E-01 — — — 208Pb — 5.24E-01 — — — Table 1. Composition and density (g cm−3) of thermoluminescent materials and dosimeters used in the simulations. HDPE Lead Air TLD600 TLD700 Density 0 .927 11.3 0.001225 2.4 2.49 1H 1.44E-01 — — — — 6Li — — — 2.30E-01 2.40E-04 7Li — — — 1.19E-02 2.69E-01 12C 8.56E-01 — — — — 14N — — 7.56E-01 — — 16O — — 2.31E-01 — — 19F — — — 7.58E-01 7.30E-01 24Mg — — — 9.48E-05 9.48E-05 25Mg — — — 1.20E-05 1.20E-05 26Mg — — — 1.32E-05 1.32E-05 40Ar — — 1.29E-02 — — 46Ti — — — 1.07E-06 1.07E-06 47Ti — — — 9.67E-07 9.67E-07 48Ti — — — 9.58E-06 9.58E-06 49Ti — — — 7.03E-07 7.03E-07 50Ti — — — 6.73E-07 6.73E-07 204Pb — 1.40E-02 — — — 205Pb — 2.41E-01 — — — 207Pb — 2.21E-01 — — — 208Pb — 5.24E-01 — — — HDPE Lead Air TLD600 TLD700 Density 0 .927 11.3 0.001225 2.4 2.49 1H 1.44E-01 — — — — 6Li — — — 2.30E-01 2.40E-04 7Li — — — 1.19E-02 2.69E-01 12C 8.56E-01 — — — — 14N — — 7.56E-01 — — 16O — — 2.31E-01 — — 19F — — — 7.58E-01 7.30E-01 24Mg — — — 9.48E-05 9.48E-05 25Mg — — — 1.20E-05 1.20E-05 26Mg — — — 1.32E-05 1.32E-05 40Ar — — 1.29E-02 — — 46Ti — — — 1.07E-06 1.07E-06 47Ti — — — 9.67E-07 9.67E-07 48Ti — — — 9.58E-06 9.58E-06 49Ti — — — 7.03E-07 7.03E-07 50Ti — — — 6.73E-07 6.73E-07 204Pb — 1.40E-02 — — — 205Pb — 2.41E-01 — — — 207Pb — 2.21E-01 — — — 208Pb — 5.24E-01 — — — The cross sections for the nuclear reactions up to 20 MeV used in the simulations were obtained from ENDF/B-VI.6 data library and for the hydrogenated materials the scattering matrix S(α,β) was also applied. For energies above 20 MeV, we used physical models that consider the elastic scattering of neutrons and protons, pre-equilibrium after the intranuclear cascade, Bertini for nucleons and Isabel for other particle types and Coulomb barrier for charged particles(20). In all simulations, F4 tally(20) was used to calculate the fluences at the point of interest, with and without the presence of the arrangement and thus determine the thermalization factor, FT, as defined in equation (7). RESULTS The response curve of the FT thermalization factor of the dosimeter obtained by simulating the arrangement exposure to monoenergetic neutron fluxes in the range of 1×10−10 to 1×104 MeV is shown in Figure 5. Figure 5. View largeDownload slide Thermalization response factor, FT, to the dosimetric arrangement. Figure 5. View largeDownload slide Thermalization response factor, FT, to the dosimetric arrangement. The averaged thermalization factor was obtained by two paths: (a) by integrating the FT(E) curve, produced from monoenergetic neutron simulations, multiplied by the normalized neutron fluence spectrum for each of the standard ISO source and CERF/CERN neutron field, by the application of equation (6) and (b) by the ratio between the total neutron fluence ϕTOT and thermal neutron fluence ϕTH determined from the simulations of the ISO standard sources and CERF/CERN neutron field, by applying equation (7). The first one, shown in Figure 5, was used to obtain the instrumental measurement of H*(10) and the second one was applied to calculate H*(10) from simulations data of each reference neutron field. The comparison of these values is shown in Table 2 and H*(10) calculated with them is shown in Table 3. Table 2. Comparison between the spectrum averaged thermalization factors determined from: (a) monoenergetic factors curve (equation (6)) and (b) Monte Carlo simulation of the fluence for each source spectrum (equation (7)). 241Am–Be 252Cf 252Cf(D2O) CERF FT(a) 1.129 ± 0.012 0.857 ± 0.010 0.867 ± 0.035 1.056 ± 0.028 FT_ (b) 1.215 ± 0.023 0.942 ± 0.018 0.901 ± 0.017 1.133 ± 0.021 a/b 0.93 ± 0.02 0.91 ± 0.02 0.96 ± 0.03 0.89 ± 0.11 241Am–Be 252Cf 252Cf(D2O) CERF FT(a) 1.129 ± 0.012 0.857 ± 0.010 0.867 ± 0.035 1.056 ± 0.028 FT_ (b) 1.215 ± 0.023 0.942 ± 0.018 0.901 ± 0.017 1.133 ± 0.021 a/b 0.93 ± 0.02 0.91 ± 0.02 0.96 ± 0.03 0.89 ± 0.11 Table 2. Comparison between the spectrum averaged thermalization factors determined from: (a) monoenergetic factors curve (equation (6)) and (b) Monte Carlo simulation of the fluence for each source spectrum (equation (7)). 241Am–Be 252Cf 252Cf(D2O) CERF FT(a) 1.129 ± 0.012 0.857 ± 0.010 0.867 ± 0.035 1.056 ± 0.028 FT_ (b) 1.215 ± 0.023 0.942 ± 0.018 0.901 ± 0.017 1.133 ± 0.021 a/b 0.93 ± 0.02 0.91 ± 0.02 0.96 ± 0.03 0.89 ± 0.11 241Am–Be 252Cf 252Cf(D2O) CERF FT(a) 1.129 ± 0.012 0.857 ± 0.010 0.867 ± 0.035 1.056 ± 0.028 FT_ (b) 1.215 ± 0.023 0.942 ± 0.018 0.901 ± 0.017 1.133 ± 0.021 a/b 0.93 ± 0.02 0.91 ± 0.02 0.96 ± 0.03 0.89 ± 0.11 Table 3. Comparison between ambient dose equivalent measured, simulated results and from reference neutron fields. 241Am–Be 252Cf 252Cf(D2O) CERF hϕ⁎(10)_a [pSv cm2] 391 385 105 228 H*(10) (Reference) [mSv] 2.137 ± 0.016 1.173 ± 0.009 0.447 ± 0.003 0.116 ± 0.012 FC [cm−2 pC−1] 5.65 ± 0.43 5.65 ± 0.43 5.65 ± 0.43 5.65 ± 0.43 FT_ (Spectrum averaged) 1.215 ± 0.023 0.942 ± 0.018 0.901 ± 0.017 1.133 ± 0.021 M [pC] (7.785 ± 0.013) × 105 (5.471 ± 0.011) × 105 (8.211 ± 0.013) × 105 (7.604 ± 0.060) × 104 H*(10) (Measurement) [mSv] 2.096 ± 0.164 1.123 ± 0.088 0.440 ± 0.034 0.111 ± 0.009 Meas./Ref. 0.98 ± 0.08 0.96 ± 0.08 0.98 ± 0.08 0.96 ± 0.12 FT (Simulated) 1.129 ± 0.012 0.857 ± 0.010 0.867 ± 0.035 1.056 ± 0.028 H*(10) (Simulated) [mSv] 1.947 ± 0.150 1.023 ± 0.079 0.423 ± 0.036 0.103 ± 0.008 Sim/Ref 0.91 ± 0.07 0.87 ± 0.07 0.95 ± 0.08 0.89 ± 0.11 241Am–Be 252Cf 252Cf(D2O) CERF hϕ⁎(10)_a [pSv cm2] 391 385 105 228 H*(10) (Reference) [mSv] 2.137 ± 0.016 1.173 ± 0.009 0.447 ± 0.003 0.116 ± 0.012 FC [cm−2 pC−1] 5.65 ± 0.43 5.65 ± 0.43 5.65 ± 0.43 5.65 ± 0.43 FT_ (Spectrum averaged) 1.215 ± 0.023 0.942 ± 0.018 0.901 ± 0.017 1.133 ± 0.021 M [pC] (7.785 ± 0.013) × 105 (5.471 ± 0.011) × 105 (8.211 ± 0.013) × 105 (7.604 ± 0.060) × 104 H*(10) (Measurement) [mSv] 2.096 ± 0.164 1.123 ± 0.088 0.440 ± 0.034 0.111 ± 0.009 Meas./Ref. 0.98 ± 0.08 0.96 ± 0.08 0.98 ± 0.08 0.96 ± 0.12 FT (Simulated) 1.129 ± 0.012 0.857 ± 0.010 0.867 ± 0.035 1.056 ± 0.028 H*(10) (Simulated) [mSv] 1.947 ± 0.150 1.023 ± 0.079 0.423 ± 0.036 0.103 ± 0.008 Sim/Ref 0.91 ± 0.07 0.87 ± 0.07 0.95 ± 0.08 0.89 ± 0.11 aObtained from reference(23,) for ISO conventional neutron sources and calculated from reference(24) for CERN/CERF neutron field. Table 3. Comparison between ambient dose equivalent measured, simulated results and from reference neutron fields. 241Am–Be 252Cf 252Cf(D2O) CERF hϕ⁎(10)_a [pSv cm2] 391 385 105 228 H*(10) (Reference) [mSv] 2.137 ± 0.016 1.173 ± 0.009 0.447 ± 0.003 0.116 ± 0.012 FC [cm−2 pC−1] 5.65 ± 0.43 5.65 ± 0.43 5.65 ± 0.43 5.65 ± 0.43 FT_ (Spectrum averaged) 1.215 ± 0.023 0.942 ± 0.018 0.901 ± 0.017 1.133 ± 0.021 M [pC] (7.785 ± 0.013) × 105 (5.471 ± 0.011) × 105 (8.211 ± 0.013) × 105 (7.604 ± 0.060) × 104 H*(10) (Measurement) [mSv] 2.096 ± 0.164 1.123 ± 0.088 0.440 ± 0.034 0.111 ± 0.009 Meas./Ref. 0.98 ± 0.08 0.96 ± 0.08 0.98 ± 0.08 0.96 ± 0.12 FT (Simulated) 1.129 ± 0.012 0.857 ± 0.010 0.867 ± 0.035 1.056 ± 0.028 H*(10) (Simulated) [mSv] 1.947 ± 0.150 1.023 ± 0.079 0.423 ± 0.036 0.103 ± 0.008 Sim/Ref 0.91 ± 0.07 0.87 ± 0.07 0.95 ± 0.08 0.89 ± 0.11 241Am–Be 252Cf 252Cf(D2O) CERF hϕ⁎(10)_a [pSv cm2] 391 385 105 228 H*(10) (Reference) [mSv] 2.137 ± 0.016 1.173 ± 0.009 0.447 ± 0.003 0.116 ± 0.012 FC [cm−2 pC−1] 5.65 ± 0.43 5.65 ± 0.43 5.65 ± 0.43 5.65 ± 0.43 FT_ (Spectrum averaged) 1.215 ± 0.023 0.942 ± 0.018 0.901 ± 0.017 1.133 ± 0.021 M [pC] (7.785 ± 0.013) × 105 (5.471 ± 0.011) × 105 (8.211 ± 0.013) × 105 (7.604 ± 0.060) × 104 H*(10) (Measurement) [mSv] 2.096 ± 0.164 1.123 ± 0.088 0.440 ± 0.034 0.111 ± 0.009 Meas./Ref. 0.98 ± 0.08 0.96 ± 0.08 0.98 ± 0.08 0.96 ± 0.12 FT (Simulated) 1.129 ± 0.012 0.857 ± 0.010 0.867 ± 0.035 1.056 ± 0.028 H*(10) (Simulated) [mSv] 1.947 ± 0.150 1.023 ± 0.079 0.423 ± 0.036 0.103 ± 0.008 Sim/Ref 0.91 ± 0.07 0.87 ± 0.07 0.95 ± 0.08 0.89 ± 0.11 aObtained from reference(23,) for ISO conventional neutron sources and calculated from reference(24) for CERN/CERF neutron field. For the calculation of the averaged fluence-to-ambient dose equivalent conversion coefficient hϕ⁎(10)_ applying equation (3), the fluence-to-ambient dose equivalent conversion coefficients obtained from the ICRP 74(17) for energies up to 19 MeV and Pelliccioni(19) for energies above 19 MeV were used. The TLDs were read by the Laboratório de Dosimetria Aeroespacial of Instituto de Estudos Avançados (LDA/IEAv) and the measured ambient equivalent dose was obtained using equation (8), where the FC calibration factor, which correlates the TL signal with the thermal neutron fluence, was provided by the LDA/IEAv. Table 3 shows the comparison between the measured and simulated ambient dose equivalent values and the respective values determined from the neutron fields. The results of the experimental measurements are compatible with the reference values within one standard deviation, although they would appear to be systematically slightly below these, in 2–4% for ISO sources and 4% for the CERF. It may be pointed that this is a good agreement for CERN/CERF neutron reference field if we consider that an intercomparison of personal dosimeters from several US accelerator facilities conducted at CERF presented deviations from 25 to 65%, as reported by Mitaroff and Silari(21). The results obtained by means of MCNPX simulation for conventional neutron sources approximate the reference values within a discrepancy from 5 to 13%, but are consistently little smaller than these reference values. For the CERF case, the discrepancies found between the simulation result and the reference value was 11%. Just as the experimental value, which is also dependent on the simulated FT factor, is just below the reference value in 4%. Such discrepancy may be due to the greater imprecision of high-energy cross section experimental data, which were estimated in MCNPX using nuclear models and, also, from some possible perturbation of the flow by the detector, that was positioned very close to the concrete roof in CERF measurements. We can expect that small adjustments in the simulation model could improve the simulation results, since the comparison with experimental values, and these with the reference values, indicates the viability of the method. CONCLUSION A method assisted by Monte Carlo simulation to improve the calibration of fast neutron dosimeters based on the process of moderation and thermalization of the incident fast flux and the measurement of the thermal flux by a sensor which responds mainly to thermal neutrons was presented in this work. This method is based on decomposition of the dosimeter calibration factor to ambient dose equivalent into three factors of distinct nature: fluence-to-ambient dose equivalent conversion factors (provided by ISO standard(23)), fast flux thermalization factor (characteristic only of the moderator body building material and the incident spectrum) and the calibration factor of the thermal neutron sensor response. The latter one is the only factor that depends on a calibration in a known neutron field (Maxwellian thermal spectrum). The essence of this method is the calculation of the thermalization factor, FT, through the Monte Carlo simulation. The FT simulations for various incident neutrons energies on the dosimeter allow us to build the response function of the dosimeter, FT(E), as a function of the incident neutron energy E, which, by a simple integration procedure over a given neutron energy spectrum, allowing to obtain the calibration factor of the monitor with better precision for the spectrum of interest. The results presented in the present work indicate that an accuracy of up to 10% can be obtained in measurements of unknown neutronic fields. This method can be applied to any neutron detector that is based on the process of thermalization of the incident fast neutrons and the resulting thermal neutron counting. For practical purposes, the dosimeter (or detector) response can be specified only by the calibration factor of the thermal neutron sensor, FC, and by the thermal factor FT(E) curve as a function of the energy E of the incident neutron. ACKNOWLEDGEMENTS The authors wish to thank CERN/CERF and Instituto de Radioproteção e Dosimetria (IRD) by making possible the opportunity to perform the neutron irradiations. REFERENCES 1 Federico, C. A. Dosimetria da radiação cósmica no interior de aeronaves no espaço aéreo brasileiro (cosmic radiation dosimetry onboard aircrafts at the brazilian air-space). PhD thesis, Instituto de Pesquisas Energéticas e Nucleares, São Paulo, São Paulo, Brazil, 2011. Available on http://www.teses.usp.br/teses/disponiveis/85/85131/tde-26012012–104504/en.php [in Portuguese]. 2 Federico, C. A., Pereira, M. A., Pereira, H. H. C., Mendes, A. C., Pazianotto, M. T. and Gonçalez, O. L. Cosmic radiation effects on aircrew and avionics in the Brazilian airspace. In: Brazilian Aerospacial Symposium. SAB 2012, São José dos Campos, São Paulo, Brazil ( 2012). 3 Federico, C. A., Gonçalez, O. L., Sordi, G. M. A. A. and Caldas, L. V. E. 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Characteristics and methods of production. ISO8529:2011(E), Part 1 ( 2011). 24 http://tis-div-rp-cerf.web.cern.ch/tis-div-rp-cerf/ © The Author(s) 2018. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com

Radiation Protection Dosimetry – Oxford University Press

**Published: ** Jan 25, 2018

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