Applied Radiation and Isotopes 95 (2015) 122–128
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Estimation of low-level neutron dose-equivalent rate by using extrapolation method for a curie level Am–Be neutron source Gang Li a, Jiayun Xu a,n, Jie Zhang b a Department of Nuclear Engineering and Technology, College of Physical Science and Technology, Sichuan University, Chengdu 610064, Sichuan Province, China b Key Laboratory for Neutron Physics, Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999, Sichuan Province, China
H I G H L I G H T S
The scope of the affected area for a curie-level Am–Be neutron source was measured. The low-level neutron dose-equivalent rates around the source increase exponentially with the increasing count rates when the source is in different shielding state. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter.
art ic l e i nf o
a b s t r a c t
Article history: Received 4 March 2014 Received in revised form 29 September 2014 Accepted 11 October 2014 Available online 22 October 2014
Neutron radiation protection is an important research area because of the strong radiation biological effect of neutron field. The radiation dose of neutron is closely related to the neutron energy, and the connected relationship is a complex function of energy. For the low-level neutron radiation field (e.g. the Am–Be source), the commonly used commercial neutron dosimeter cannot always reflect the low-level dose rate, which is restricted by its own sensitivity limit and measuring range. In this paper, the intensity distribution of neutron field caused by a curie level Am–Be neutron source was investigated by measuring the count rates obtained through a 3He proportional counter at different locations around the source. The results indicate that the count rates outside of the source room are negligible compared with the count rates measured in the source room. In the source room, 3He proportional counter and neutron dosimeter were used to measure the count rates and dose rates respectively at different distances to the source. The results indicate that both the count rates and dose rates decrease exponentially with the increasing distance, and the dose rates measured by a commercial dosimeter are in good agreement with the results calculated by the Geant4 simulation within the inherent errors recommended by ICRP and IEC. Further studies presented in this paper indicate that the low-level neutron dose equivalent rates in the source room increase exponentially with the increasing low-energy neutron count rates when the source is lifted from the shield with different radiation intensities. Based on this relationship as well as the count rates measured at larger distance to the source, the dose rates can be calculated approximately by the extrapolation method. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter. & 2014 Elsevier Ltd. All rights reserved.
Keywords: Low-level neutron dose value Am–Be neutron source 3 He proportional countertube BH3105 type neutron dose-equivalent meter Extrapolation method
1. Introduction Am–Be neutron sources are widely used in scientific research and practical teaching in the laboratories because of their low cost and
n
Corresponding author. Tel.: þ 86 28 85412258. E-mail addresses: [email protected] (G. Li), [email protected] (J. Xu).
http://dx.doi.org/10.1016/j.apradiso.2014.10.010 0969-8043/& 2014 Elsevier Ltd. All rights reserved.
convenience for operation. Nevertheless, as its strong penetrating power and strong radiation biological effect, people working in/ around the Am–Be source room are always worried about its radioactivity in the source room and its adjacent rooms (Ghassoun and Senhou, 2012; Wang et al., 2011). Generally, if the source was properly shielded with low atomic-number materials, then the dose values maintain a relatively lower level. The neutron dose-equivalent rate close to the source can be easily got by using a commercial
G. Li et al. / Applied Radiation and Isotopes 95 (2015) 122–128
neutron dosimeter. But for a farther distance around the source, the dosimeter cannot always reflect such a low-level dose rate, which is restricted by its own sensitivity limit and measuring range. Many studies have been done to explore how to measure neutron dose values more accurately, but these studies were mostly focused on the high neutron dose field. Up to now, few studies have been done to estimate the low-level neutron dose-equivalent rate caused by a shielded Am–Be neutron source, and further studies are still essential. This paper aims at investigating the low-level neutron doseequivalent rate in the working area around the source. As is widely known that the radiation effect of neutron action on human body is closely related to the neutron energy, meanwhile the connected relationship is a complex function of energy (Hunt and Kluge, 1985; Gómez et al., 2010; Hernandez-Davila et al., 2014). But for a shielded neutron source, this relationship can be simplified, because the vast majority of the neutrons escaped from the shield have very low energy (o10 keV). Consequently, this paper studied the distribution of neutrons by measuring count rates through a 3He gas proportional counter. To further known the dose-equivalent rates at larger distances to the source, two basic instruments including a 3He tube and a commercial dosimeter were used to obtain the internal relationships between count rates and dose-equivalent rates. Then according to the measured corresponding relationship as well as the count rates measured at larger distances to the source, the lowlevel neutron dose-equivalent rates in the source room can be calculated approximately by the extrapolation method. To sum up, the present work tentatively put forward a method of estimating the low-level neutron dose-equivalent rates caused by a shielded Am–Be neutron source, which cannot be measured directly by the dosimeter. This method is of great value for lowlevel neutron radioactivity assessment and radiation protection.
2. Experimental instruments and descriptions 2.1. BH3105 type neutron dose-equivalent meter (Ji et al., 2011) The neutron dosimeter used in this study is a commercially used BH3105 type neutron dose-equivalent meter, which is designed and manufactured by Beijing Nuclear Instrument Factory of China National Nuclear Corporation. As shown in Fig. 1, a polyethylene sphere of 20 cm in diameter is used as neutron moderator, at the center of which placed a spherical detector made of 6Li glass scintillator. Furthermore, about 100 Cadmium absorption sticks with different lengths are evenly distributed on the surface of the moderator to realize the absorption stick principle method. This
Polyethylene moderator Cadmium absorber rods Photomultiplier tube Emitter follower Signal High voltage Low voltage
6
Li glass scintillator
Fig. 1. Schematic diagram of the probe of the BH3105 type neutron doseequivalent meter.
123
method is used for thermal 14 MeV neutron equal dose-equivalent detection, which gives high neutron sensitivity and wide measurement range of 10 cps/(μSv h 1) and 0.1 μSv h 1–999.9 mSv h 1, respectively. The energy response characteristic of the dosimeter satisfies the ICRP standard. The gamma inhibition coefficient and the deviation of directional response are superior to the international standard. 2.2.
3
He proportional countertube
The 3He gas proportional counters are particularly sensitive to detect low-energy neutrons with high cross section of 5330 b for thermal neutron (Falahat et al., 2013), which makes it an ideal candidate for the measurement of neutrons emitted from a shielded source. The cylindrical 3He counter used in this study was manufactured by Centronic Company, which consists of a stainless steel tube that is filled with 3He under a pressure of 6 atm. The neutron sensitivity of the counter with a diameter of 25 mm and an effective length of 127 mm is 35 cps/(n/cm2 s) (Centronic Company), and the optimum operating voltage based on the experimental measured plateau curve is 1200 V. 2.3. Am–Be neutron source storage configuration The cylindrical Am–Be neutron source used in this study is manufactured by the China Institute of Atomic Energy. The source is surrounded by three layers of thin stainless steel with the total height and diameter of 36 mm and 30 mm, respectively, and the source activity is 7 109 Bq with the neutron emission rate of 4.6 106 s 1 (Li et al., 2011). The source and its shield are located in the corner of the neutron source room shown in Figs. 2 and 3. Based on Fig. 2, the source is located at the bottom of the PVC tube in normal shielding state, and the PVC tube with an open upper end and a sealed lower end is suspended in the center of the cylindrical steel pail. The steel pail is filled with deionized water and surrounded with dozens of centimeters thickness of paraffin to further slow down the escaped neutrons. In addition, for the purpose of shielding the neutrons emitted from the upper end of the PVC tube, a cylindrical paraffin rod is inserted into the PVC tube as well. 3. Distribution of low-energy neutrons measured by 3He tube In order to investigate the distribution of the neutron field, several positions in the source room and its adjacent rooms were selected as the measuring points, which were all located on the first floor of the laboratory building. These measuring points were selected at representative positions which were considered to be important. As shown in Fig. 3 with an X–Y coordinate axis, all the measuring points were chosen along the X axis and located in the ZE70 cm plane. Here, we made the assumption that all the shielding materials (including walls and other objects in the room) along other directions were considered to be similar as the X axis in the X–Y plane. The neutrons emitted from the shield were mostly low-energy neutrons with energy below 0.01 MeV, and then the emitted neutrons were promptly moderated through elastic scattering and inelastic scattering by walls and other low-atomic-number materials. Therefore, considering the moderating effect of the shield and walls, the count rates outside of the source room are much lower. So in order to reduce the measuring statistical errors, the count rates of point C–O were measured at least 3 times to get its average value with measuring time of 12 h. The measured count rates at different measuring points change with distance are summarized in Table 1. Table 1 presents the measured neutron count rates at different measuring points around the source. It can be seen that the count
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Fig. 2. Schematic diagram of the source and its shield; (a) front view; (b) top view. The neutron source lies at the bottom of the PVC tube in normal shielding state.
Measuring points Y
·F
Walls Shield
Source B room
Source
·
·A ·C
·K
·E ·H ·D
·G
·I
·J
·N ·L
·O
·M X
Fig. 3. The plan view of the measuring points around the source room (the thickness of the wall is 0.6 m and all the measurements were performed in the z ¼ 0.7 m plane).
Table 1 Neutron count rates measured by 3He tube versus positions of each measuring point. Measuring point
Coordinate (m, m)
Distance (m)
Count rate (h 1)
A B C D E F G H I J K L M N O
(0, 0.300) (0, 2.900) (2.650, 0.200) (4.500, 2.100) (4.950, 4.100) (6.600, 0.800) (7.050, 4.100) (9.400, 2.350) (11.800, 0.350) (13.150, 3.650) (16.150, 2.700) (19.950,2.700) (21.400, 4.650) (22.650, 5.350) (23.786, 5.769)
0.30 2.90 2.66 4.97 6.43 6.65 8.16 9.69 11.81 13.65 16.37 20.13 21.90 23.27 24.48
68,9747 214 5526 7 43 130.5 7 2.5 137.4 7 4.7 108.17 2.5 95.5 7 3.5 112.5 7 2.6 80.6 7 1.1 74.5 7 2.6 65.8 7 3.9 67.2 7 2.3 57.5 7 1.6 47.17 1.2 45.4 7 3.1 45.17 2.0
Note: Point A and point B were located in the source room.
rates of point C–O in the adjacent rooms are so small that can be ignored, compared to the count rates of point A–B in the source room. This result is in good agreement with the dose results measured by the BH3105 type dosimeter directly, where the dose values of point C–O are so small that cannot be measured. The measured count rate values n of point C–O as a function of distance x are presented in Fig. 4. These count rate values contain the contribution of the natural background produced by the cosmic-ray, and this contribution was subtracted from the measured experimental values according to Table 1. Fig. 4 clearly indicates that the neutron count rates decrease exponentially with the distances increase. An exponential fitting curve with the Adj. R-squared value of 0.9 has also been appended in Fig. 4. It can be
Fig. 4. Neutron count rates measured by the 3He tube versus distance outside of the source room.
also found that the values at some individual measuring points have large deviation compared with the fitted curve. This may be due to the neutron scattering phenomenon which causes a high neutron flux in some area (Eisenhauer, 1989).
4. Relationship of dose-equivalent rates and count rates 4.1. Dose-equivalent rate and count rate versus distance Compared with the strength distribution of the neutron field, the low-level neutron dose-equivalent rates in the source room need to be paid more attention. Because the walls around the source room are thick enough, hence the dose rates outside of the source room are so low that can almost be ignored. But in the source room the dose rates are much higher, especially when the source is lifted from the bottom of the PVC tube for different heights. Most of the time the source is
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Fig. 5. Schematic diagram of the source and measuring points; (a) front view; (b) top view.
located at the bottom of the PVC tube in normal shielding state. But in practical applications, the source is likely to be lifted up for some distance to obtain a larger neutron flux density. In this way, the doseequivalent rates outside of the shield will be observably increased and needed to be paying more attention. For obtain these low level neutron dose-equivalent values, we tried to find the internal relationship between neutron count rates and neutron dose-equivalent rates measured respectively by two basic instruments, 3He proportional countertube and BH3105 type neutron dose-equivalent meter. In order to investigate the internal relationship at different shielding states to obtain a reliable and unified result, the neutron count rates and neutron dose-equivalent rates of measuring points U–Z were measured respectively when the source was lifted up for different heights. In this study, the source was lifted up equidistantly from the bottom of the PVC tube to 15 cm (Position 1), 22 cm (Position 2) and 29 cm (Position 3) shown in Fig. 5. The neutron count rates of U–Z points were measured 3 times with each measurement of 10 min, and the dose-equivalent rates of these points were measured 10 times with the default measuring time. All of these data were calculated for their average value. Furthermore, all the measuring points located equidistantly in a straight line paralleled to the upper bottom surface of the shield, and the space between the two adjacent measuring points was 35 cm. The measured neutron count rates and neutron dose-equivalent rates change with the distance increases are shown in Fig. 6, and the exponential fitting curves with the R-squared values of 0.99 have also been appended. 4.2. Simulation of dose-equivalent rate using a Monte Carlo method The neutron spectrum of Am–Be source is shown in Fig. 7 with the mean energy of 4.3 MeV (ISO 8529-1, International Standard, 2001). As is known to all that the neutron dose is closely related to the neutron energy, and this connected relationship is a complex
function of neutron energy. Table 2 summarizes part of the typical neutron dose conversion coefficient values per unit fluence of different energy based on the International Standard IEC 1322 and ICRU report (IEC-61322, International Standard, 1994; ICRU, International Commission on Radiation Units and Measurement, 1998); these values indicate that different neutron energies per unit fluence correspond to different dose conversion coefficients. Based on its energy and the conversion coefficient as well as the normalization factor, the neutrons with energy ranging from 10 8 MeV to 10 MeV were divided into 3 groups (Ji et al., 2011). The group T, represented by the thermal neutrons with neutron energy less than 104 eV, has the mean conversion coefficient of 13.1 pSv cm2; the group S, represented by neutrons with energy of 105 eV, has the mean conversion coefficient of 116.3; and the group F, represented by neutrons with energy more than 5 105 eV, has the mean conversion coefficient of 435.3. The mean normalization factors of the above three groups can be simplified to 0.025, 0.25 and 1. In the practical design process to develop a neutron dose-meter, the primary goal is to make the ratio of detection efficiency approximate at 0.025: 0.25: 1 for the above mentioned three groups to realize the measurement of bioequivalence. In order to identify the accuracy of the experimental results of BH3105, the Monte Carlo simulation toolkit GEANT4 (Geant4.10.00. p02) (Agostinelli et al., 2003; CERN, 2013) was used to calculate the neutron dose-equivalent values of measuring points U–Z when the source was lifted in different shielding states. The neutron dose is related to the neutron energy, and higher energy corresponds to higher dose values for a constant neutron fluence rate (Table 2). In the GEANT4 simulation process, the neutron dose-equivalent rates were calculated by the following steps. First, the simulation model with the same dimension as described in Figs. 2 and 5 was used in the simulation. Then, the neutron fluence rates for different energies or groups of point U–Z were recorded automatically by the program.
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Fig. 6. Neutron count rates and neutron dose-equivalent rates versus distance when the source is lifted from the bottom of the PVC tube to (a) 15 cm (position 1 in Fig. 5(a)), (b) 22 cm (position 2 in Fig. 5(a)) and (c) 29 cm (position 3 in Fig. 5(a)).
Table 2 Dose conversion coefficient values per unit fluence for different neutron energy.
Fig. 7. 241Am–Be neutron spectrum (B represents the source strength of different energies).
At last, the neutron dose-equivalent rates were calculated depending on the neutron sensitivity of the BH3105 and the dose conversion coefficient values shown in Table 2. All the neutron dose-equivalent
Neutron energy (eV)
Conversion coefficient (pSv cm2)
Normalization factor
Group
2.5 10 2 1 10 100 103 104 105 5 105 106 5 106 107
13.1 16.2 14.8 11.7 9.3 13.2 116.3 (mean) 367 439 472 463
0.03 0.04 0.03 0.02 0.02 0.03 0.26 0.84 1.00 1.08 1.06
T
S F
rates were calculated within the relative uncertainty of 3.0% with the neutrons generated more than 50,000,000. The simulation results are shown in Fig. 6.
G. Li et al. / Applied Radiation and Isotopes 95 (2015) 122–128
127
Fig. 8. Calculated neutron spectrum of measuring points U–Z when the source is in normal shielding state (a) and lifted to 15 (b), 22 (c) and 29 cm (d). The neutron energy was divided into 3 groups according to Table 2.
The calculated neutron spectrum of measuring points U–Z is also appended in Fig. 8 when the source was in different shielding states. The spectrum was simplified and represented by the proportion of the above mentioned three groups. For a certain shielding state, the proportion of groups T, S and F remains almost the same for each point, and can be approximated to 65%, 20% and 15% respectively. The simulation results indicate that the overwhelming majority of neutrons escaped from the shield belong to group T with energy below 104 eV.
4.3. Neutron dose-equivalent rates versus neutron count rates As discussed in the above sections that the neutron dose value is closely related to the neutron energy and different energy types of neutrons per unit fluence have different dose conversion coefficient values. Hence, for the low-energy neutron radiation field in our study, the low-level neutron dose equivalent rate values and the neutron count rate values may have some internal connections which can be used to estimate the low level neutron dose values that cannot be measured by the commercial dosimeter directly. For this purpose, the measured neutron dose equivalent rate values as a function of measured neutron count rate values when the source was lifted to 15, 22 and 29 cm are presented in Fig. 9, and the fitting curves are also appended in Fig. 9 by using exponential form. The results indicate that the low level neutron dose equivalent rate values change exponentially with the neutron count rate values when the source is lifted to different heights. According to these results, the low level neutron dose equivalent rate values can be expressed by the neutron count rate values by the exponential formulas shown in Fig. 9. The items of H15, H22, H29 and n15, n22, n29 are the neutron dose equivalent rate values and neutron count rate values when the source is lifted to 15, 22
and 29 cm, respectively. The terms of R-Squared are the goodness of fit of the above formulas which are all above 0.99. It is evident in Figs. 6 and 9 that the simulation results have large deviation with the experimental results. For the dose values, the deviation is mainly contributed by the inherent error of the BH3105 type neutron dose-equivalent meter, where the inherent error is due to the complexity of the measurement itself. It is difficult for a commercial neutron dose-meter to obtain the dose values absolutely accurate. In order to evaluate the reliability and energy response of a neutron dosimeter, the inherent error of 50% to þ100% was recommended as the reference threshold to design the dosimeter by the ICRP organization (ICRP, 2007) and the international standard IEC (IEC-61322, International Standard, 1994; IEC-61005, International Standard, 2014). Therefore, the deviation of neutron dose values between experimental and simulative results is always subsistent and inevitable. For all that, the experimental results are still in quantitative agreement with the results calculated by means of Geant4 simulation within the inherent errors recommended by ICRP and IEC.
5. Conclusions From the intensity distribution of the neutron field measured by He proportional countertube at different measuring points in the source room and its adjacent rooms, the scope of the affected area was obtained. Based on the intensity distribution, the neutrons with lower energy decrease exponentially with the increasing distance approximately. Further, according to the measured count rates and dose equivalent rates in the source room when the source is lifted to different heights, the phenomenon that the dose rate values change exponentially with the count rates is found, and the fitted exponential curve
3
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G. Li et al. / Applied Radiation and Isotopes 95 (2015) 122–128
Based on the results, if the relationship between dose rates and count rates near to the source was obtained, then the dose rates at larger distances to the source can be calculated approximately according to the fitted curve and the measured count rates, where the count rates can be easily acquired by the 3He proportional countertube. This principle can be used to estimate the low-level neutron dose equivalent values at larger distance to the source where the dose rate cannot be measured directly by commercial dosimeters. The results are of great value for low-level neutron radioactivity assessment and radiation protection of curie level Am–Be neutron source.
References
Fig. 9. Neutron dose-equivalent rates versus neutron count rates in the source room outside of the shield; the source is lifted to 15 cm (a), 22 cm (b), and 29 cm (c).
fits very well with the original measured data. Furthermore, the experimental results have a good agreement with the simulative results within the inherent errors recommended by the ICRP and IEC.
Agostinelli, S., Allison, J., Amako, K., Apostolakis, J., Araujo, H., et al., 2003. Geant4—a simulation toolkit. Nucl. Instrum. Methods Phys. Res. A 506, 250–303. CERN, 2013. Geant4 User's Guides. Available at: 〈http://geant4.cern.ch/support/ userdocuments.shtml〉. Centronic Company. 3He neutron proportional counters. Available at: 〈http://www. centronic.co.uk/helium.htm〉. Eisenhauer, C.M., 1989. Review of scattering corrections for calibration of neutron instruments. Radiat. Prot. Dosim. 28 (4), 253–262. Falahat, Sascha, Best, Andreas, Couder, Manoel, et al., 2013. A 3He neutron detector for the measurement of (α, n) reactions. Nucl. Instrum. Methods Phys. Res. A 700, 53–58. Ghassoun, J., Senhou, N., 2012. The evaluation of neutron and gamma ray dose equivalent distributions in patients and the effectiveness of shield materials for high energy photons radiotherapy facilities. Appl. Radiat. Isot. 70, 620–624. Gómez, F., Iglesias, A., Sánchez-Doblado, F., 2010. A new active method for the measurement of slow-neutronfluence in modern radiotherapy treatment rooms. Phys. Med. Biol. 55, 1025–1039. Hernandez-Davila, Victor M., Soto-Bernal, Tzinnia G., Vega-Carrillo, Hector Rene, 2014. Determination of neutron fluence-to-dose conversion coefficients by means of artificial neural networks. Appl. Radiat. Isot. 83, 249–251. Hunt, J.B., Kluge, H., 1985. Different relative responses as a function of neutron energy of a number of neutron dose equivalent rate meters of the same type and a possible explanation. Radiat. Prot. Dosim. 12 (1), 29–32. ICRP, 2007. The 2007 Recommendations of the International Commission on Radiological Protection. ICRP Publication 103. Annals of the ICRP. Approved by the commission in March 2007. Conversion Coefficients for Use in Radiological Protection Against External Radiation. ICRU, Bethesda, Maryland (Report 57). IEC-61005, International Standard, 2014. Radiation Protection Instrumentation— Neutron Ambient Dose Equivalent (Rate) Meters (Edition 3.0). IEC-61322, International Standard, 1994. Radiation Protection Instrumentation— Installed Dose Equivalent Rate Meters, Warning Assemblies and Monitors for Neutron Radiation of Energy from Thermal to 15 MeV (First Edition). ISO 8529-1, International Standard, 2001. Reference Neutron Radiations-Part 1: Characteristic and Methods of Production (First Edition). Ji, Changsong, Wang, Tingting, Zhang, Shuheng, et al., 2011. The development of BH3105E type neutron dose-equivalent meter. (in Chinese). Nucl. Electron. Detect. Technol. 31 (7), 728–731. Li, Chunfang, Xu, Jiayun, Deng, Di, 2011. Distribution of thermal neutron flux density produced by curie class Am–Be neutron source (in Chinese). Nucl. Electron. Detect. Technol. 31 (9), 983–985. Wang, J.N., Huang, C.K., Tsai, W.C., Liu, Y.H., Jiang, S.H., 2011. Effective dose evaluation for BNCT treatment in the epithermal neutron beam at THOR. Appl. Radiat. Isot. 69, 1850–1853.
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Estimation of low-level neutron dose-equivalent rate by using extrapolation method for a curie level Am–Be neutron source Gang Li a, Jiayun Xu a,n, Jie Zhang b a Department of Nuclear Engineering and Technology, College of Physical Science and Technology, Sichuan University, Chengdu 610064, Sichuan Province, China b Key Laboratory for Neutron Physics, Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999, Sichuan Province, China
H I G H L I G H T S
The scope of the affected area for a curie-level Am–Be neutron source was measured. The low-level neutron dose-equivalent rates around the source increase exponentially with the increasing count rates when the source is in different shielding state. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter.
art ic l e i nf o
a b s t r a c t
Article history: Received 4 March 2014 Received in revised form 29 September 2014 Accepted 11 October 2014 Available online 22 October 2014
Neutron radiation protection is an important research area because of the strong radiation biological effect of neutron field. The radiation dose of neutron is closely related to the neutron energy, and the connected relationship is a complex function of energy. For the low-level neutron radiation field (e.g. the Am–Be source), the commonly used commercial neutron dosimeter cannot always reflect the low-level dose rate, which is restricted by its own sensitivity limit and measuring range. In this paper, the intensity distribution of neutron field caused by a curie level Am–Be neutron source was investigated by measuring the count rates obtained through a 3He proportional counter at different locations around the source. The results indicate that the count rates outside of the source room are negligible compared with the count rates measured in the source room. In the source room, 3He proportional counter and neutron dosimeter were used to measure the count rates and dose rates respectively at different distances to the source. The results indicate that both the count rates and dose rates decrease exponentially with the increasing distance, and the dose rates measured by a commercial dosimeter are in good agreement with the results calculated by the Geant4 simulation within the inherent errors recommended by ICRP and IEC. Further studies presented in this paper indicate that the low-level neutron dose equivalent rates in the source room increase exponentially with the increasing low-energy neutron count rates when the source is lifted from the shield with different radiation intensities. Based on this relationship as well as the count rates measured at larger distance to the source, the dose rates can be calculated approximately by the extrapolation method. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter. & 2014 Elsevier Ltd. All rights reserved.
Keywords: Low-level neutron dose value Am–Be neutron source 3 He proportional countertube BH3105 type neutron dose-equivalent meter Extrapolation method
1. Introduction Am–Be neutron sources are widely used in scientific research and practical teaching in the laboratories because of their low cost and
n
Corresponding author. Tel.: þ 86 28 85412258. E-mail addresses: [email protected] (G. Li), [email protected] (J. Xu).
http://dx.doi.org/10.1016/j.apradiso.2014.10.010 0969-8043/& 2014 Elsevier Ltd. All rights reserved.
convenience for operation. Nevertheless, as its strong penetrating power and strong radiation biological effect, people working in/ around the Am–Be source room are always worried about its radioactivity in the source room and its adjacent rooms (Ghassoun and Senhou, 2012; Wang et al., 2011). Generally, if the source was properly shielded with low atomic-number materials, then the dose values maintain a relatively lower level. The neutron dose-equivalent rate close to the source can be easily got by using a commercial
G. Li et al. / Applied Radiation and Isotopes 95 (2015) 122–128
neutron dosimeter. But for a farther distance around the source, the dosimeter cannot always reflect such a low-level dose rate, which is restricted by its own sensitivity limit and measuring range. Many studies have been done to explore how to measure neutron dose values more accurately, but these studies were mostly focused on the high neutron dose field. Up to now, few studies have been done to estimate the low-level neutron dose-equivalent rate caused by a shielded Am–Be neutron source, and further studies are still essential. This paper aims at investigating the low-level neutron doseequivalent rate in the working area around the source. As is widely known that the radiation effect of neutron action on human body is closely related to the neutron energy, meanwhile the connected relationship is a complex function of energy (Hunt and Kluge, 1985; Gómez et al., 2010; Hernandez-Davila et al., 2014). But for a shielded neutron source, this relationship can be simplified, because the vast majority of the neutrons escaped from the shield have very low energy (o10 keV). Consequently, this paper studied the distribution of neutrons by measuring count rates through a 3He gas proportional counter. To further known the dose-equivalent rates at larger distances to the source, two basic instruments including a 3He tube and a commercial dosimeter were used to obtain the internal relationships between count rates and dose-equivalent rates. Then according to the measured corresponding relationship as well as the count rates measured at larger distances to the source, the lowlevel neutron dose-equivalent rates in the source room can be calculated approximately by the extrapolation method. To sum up, the present work tentatively put forward a method of estimating the low-level neutron dose-equivalent rates caused by a shielded Am–Be neutron source, which cannot be measured directly by the dosimeter. This method is of great value for lowlevel neutron radioactivity assessment and radiation protection.
2. Experimental instruments and descriptions 2.1. BH3105 type neutron dose-equivalent meter (Ji et al., 2011) The neutron dosimeter used in this study is a commercially used BH3105 type neutron dose-equivalent meter, which is designed and manufactured by Beijing Nuclear Instrument Factory of China National Nuclear Corporation. As shown in Fig. 1, a polyethylene sphere of 20 cm in diameter is used as neutron moderator, at the center of which placed a spherical detector made of 6Li glass scintillator. Furthermore, about 100 Cadmium absorption sticks with different lengths are evenly distributed on the surface of the moderator to realize the absorption stick principle method. This
Polyethylene moderator Cadmium absorber rods Photomultiplier tube Emitter follower Signal High voltage Low voltage
6
Li glass scintillator
Fig. 1. Schematic diagram of the probe of the BH3105 type neutron doseequivalent meter.
123
method is used for thermal 14 MeV neutron equal dose-equivalent detection, which gives high neutron sensitivity and wide measurement range of 10 cps/(μSv h 1) and 0.1 μSv h 1–999.9 mSv h 1, respectively. The energy response characteristic of the dosimeter satisfies the ICRP standard. The gamma inhibition coefficient and the deviation of directional response are superior to the international standard. 2.2.
3
He proportional countertube
The 3He gas proportional counters are particularly sensitive to detect low-energy neutrons with high cross section of 5330 b for thermal neutron (Falahat et al., 2013), which makes it an ideal candidate for the measurement of neutrons emitted from a shielded source. The cylindrical 3He counter used in this study was manufactured by Centronic Company, which consists of a stainless steel tube that is filled with 3He under a pressure of 6 atm. The neutron sensitivity of the counter with a diameter of 25 mm and an effective length of 127 mm is 35 cps/(n/cm2 s) (Centronic Company), and the optimum operating voltage based on the experimental measured plateau curve is 1200 V. 2.3. Am–Be neutron source storage configuration The cylindrical Am–Be neutron source used in this study is manufactured by the China Institute of Atomic Energy. The source is surrounded by three layers of thin stainless steel with the total height and diameter of 36 mm and 30 mm, respectively, and the source activity is 7 109 Bq with the neutron emission rate of 4.6 106 s 1 (Li et al., 2011). The source and its shield are located in the corner of the neutron source room shown in Figs. 2 and 3. Based on Fig. 2, the source is located at the bottom of the PVC tube in normal shielding state, and the PVC tube with an open upper end and a sealed lower end is suspended in the center of the cylindrical steel pail. The steel pail is filled with deionized water and surrounded with dozens of centimeters thickness of paraffin to further slow down the escaped neutrons. In addition, for the purpose of shielding the neutrons emitted from the upper end of the PVC tube, a cylindrical paraffin rod is inserted into the PVC tube as well. 3. Distribution of low-energy neutrons measured by 3He tube In order to investigate the distribution of the neutron field, several positions in the source room and its adjacent rooms were selected as the measuring points, which were all located on the first floor of the laboratory building. These measuring points were selected at representative positions which were considered to be important. As shown in Fig. 3 with an X–Y coordinate axis, all the measuring points were chosen along the X axis and located in the ZE70 cm plane. Here, we made the assumption that all the shielding materials (including walls and other objects in the room) along other directions were considered to be similar as the X axis in the X–Y plane. The neutrons emitted from the shield were mostly low-energy neutrons with energy below 0.01 MeV, and then the emitted neutrons were promptly moderated through elastic scattering and inelastic scattering by walls and other low-atomic-number materials. Therefore, considering the moderating effect of the shield and walls, the count rates outside of the source room are much lower. So in order to reduce the measuring statistical errors, the count rates of point C–O were measured at least 3 times to get its average value with measuring time of 12 h. The measured count rates at different measuring points change with distance are summarized in Table 1. Table 1 presents the measured neutron count rates at different measuring points around the source. It can be seen that the count
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Fig. 2. Schematic diagram of the source and its shield; (a) front view; (b) top view. The neutron source lies at the bottom of the PVC tube in normal shielding state.
Measuring points Y
·F
Walls Shield
Source B room
Source
·
·A ·C
·K
·E ·H ·D
·G
·I
·J
·N ·L
·O
·M X
Fig. 3. The plan view of the measuring points around the source room (the thickness of the wall is 0.6 m and all the measurements were performed in the z ¼ 0.7 m plane).
Table 1 Neutron count rates measured by 3He tube versus positions of each measuring point. Measuring point
Coordinate (m, m)
Distance (m)
Count rate (h 1)
A B C D E F G H I J K L M N O
(0, 0.300) (0, 2.900) (2.650, 0.200) (4.500, 2.100) (4.950, 4.100) (6.600, 0.800) (7.050, 4.100) (9.400, 2.350) (11.800, 0.350) (13.150, 3.650) (16.150, 2.700) (19.950,2.700) (21.400, 4.650) (22.650, 5.350) (23.786, 5.769)
0.30 2.90 2.66 4.97 6.43 6.65 8.16 9.69 11.81 13.65 16.37 20.13 21.90 23.27 24.48
68,9747 214 5526 7 43 130.5 7 2.5 137.4 7 4.7 108.17 2.5 95.5 7 3.5 112.5 7 2.6 80.6 7 1.1 74.5 7 2.6 65.8 7 3.9 67.2 7 2.3 57.5 7 1.6 47.17 1.2 45.4 7 3.1 45.17 2.0
Note: Point A and point B were located in the source room.
rates of point C–O in the adjacent rooms are so small that can be ignored, compared to the count rates of point A–B in the source room. This result is in good agreement with the dose results measured by the BH3105 type dosimeter directly, where the dose values of point C–O are so small that cannot be measured. The measured count rate values n of point C–O as a function of distance x are presented in Fig. 4. These count rate values contain the contribution of the natural background produced by the cosmic-ray, and this contribution was subtracted from the measured experimental values according to Table 1. Fig. 4 clearly indicates that the neutron count rates decrease exponentially with the distances increase. An exponential fitting curve with the Adj. R-squared value of 0.9 has also been appended in Fig. 4. It can be
Fig. 4. Neutron count rates measured by the 3He tube versus distance outside of the source room.
also found that the values at some individual measuring points have large deviation compared with the fitted curve. This may be due to the neutron scattering phenomenon which causes a high neutron flux in some area (Eisenhauer, 1989).
4. Relationship of dose-equivalent rates and count rates 4.1. Dose-equivalent rate and count rate versus distance Compared with the strength distribution of the neutron field, the low-level neutron dose-equivalent rates in the source room need to be paid more attention. Because the walls around the source room are thick enough, hence the dose rates outside of the source room are so low that can almost be ignored. But in the source room the dose rates are much higher, especially when the source is lifted from the bottom of the PVC tube for different heights. Most of the time the source is
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Fig. 5. Schematic diagram of the source and measuring points; (a) front view; (b) top view.
located at the bottom of the PVC tube in normal shielding state. But in practical applications, the source is likely to be lifted up for some distance to obtain a larger neutron flux density. In this way, the doseequivalent rates outside of the shield will be observably increased and needed to be paying more attention. For obtain these low level neutron dose-equivalent values, we tried to find the internal relationship between neutron count rates and neutron dose-equivalent rates measured respectively by two basic instruments, 3He proportional countertube and BH3105 type neutron dose-equivalent meter. In order to investigate the internal relationship at different shielding states to obtain a reliable and unified result, the neutron count rates and neutron dose-equivalent rates of measuring points U–Z were measured respectively when the source was lifted up for different heights. In this study, the source was lifted up equidistantly from the bottom of the PVC tube to 15 cm (Position 1), 22 cm (Position 2) and 29 cm (Position 3) shown in Fig. 5. The neutron count rates of U–Z points were measured 3 times with each measurement of 10 min, and the dose-equivalent rates of these points were measured 10 times with the default measuring time. All of these data were calculated for their average value. Furthermore, all the measuring points located equidistantly in a straight line paralleled to the upper bottom surface of the shield, and the space between the two adjacent measuring points was 35 cm. The measured neutron count rates and neutron dose-equivalent rates change with the distance increases are shown in Fig. 6, and the exponential fitting curves with the R-squared values of 0.99 have also been appended. 4.2. Simulation of dose-equivalent rate using a Monte Carlo method The neutron spectrum of Am–Be source is shown in Fig. 7 with the mean energy of 4.3 MeV (ISO 8529-1, International Standard, 2001). As is known to all that the neutron dose is closely related to the neutron energy, and this connected relationship is a complex
function of neutron energy. Table 2 summarizes part of the typical neutron dose conversion coefficient values per unit fluence of different energy based on the International Standard IEC 1322 and ICRU report (IEC-61322, International Standard, 1994; ICRU, International Commission on Radiation Units and Measurement, 1998); these values indicate that different neutron energies per unit fluence correspond to different dose conversion coefficients. Based on its energy and the conversion coefficient as well as the normalization factor, the neutrons with energy ranging from 10 8 MeV to 10 MeV were divided into 3 groups (Ji et al., 2011). The group T, represented by the thermal neutrons with neutron energy less than 104 eV, has the mean conversion coefficient of 13.1 pSv cm2; the group S, represented by neutrons with energy of 105 eV, has the mean conversion coefficient of 116.3; and the group F, represented by neutrons with energy more than 5 105 eV, has the mean conversion coefficient of 435.3. The mean normalization factors of the above three groups can be simplified to 0.025, 0.25 and 1. In the practical design process to develop a neutron dose-meter, the primary goal is to make the ratio of detection efficiency approximate at 0.025: 0.25: 1 for the above mentioned three groups to realize the measurement of bioequivalence. In order to identify the accuracy of the experimental results of BH3105, the Monte Carlo simulation toolkit GEANT4 (Geant4.10.00. p02) (Agostinelli et al., 2003; CERN, 2013) was used to calculate the neutron dose-equivalent values of measuring points U–Z when the source was lifted in different shielding states. The neutron dose is related to the neutron energy, and higher energy corresponds to higher dose values for a constant neutron fluence rate (Table 2). In the GEANT4 simulation process, the neutron dose-equivalent rates were calculated by the following steps. First, the simulation model with the same dimension as described in Figs. 2 and 5 was used in the simulation. Then, the neutron fluence rates for different energies or groups of point U–Z were recorded automatically by the program.
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Fig. 6. Neutron count rates and neutron dose-equivalent rates versus distance when the source is lifted from the bottom of the PVC tube to (a) 15 cm (position 1 in Fig. 5(a)), (b) 22 cm (position 2 in Fig. 5(a)) and (c) 29 cm (position 3 in Fig. 5(a)).
Table 2 Dose conversion coefficient values per unit fluence for different neutron energy.
Fig. 7. 241Am–Be neutron spectrum (B represents the source strength of different energies).
At last, the neutron dose-equivalent rates were calculated depending on the neutron sensitivity of the BH3105 and the dose conversion coefficient values shown in Table 2. All the neutron dose-equivalent
Neutron energy (eV)
Conversion coefficient (pSv cm2)
Normalization factor
Group
2.5 10 2 1 10 100 103 104 105 5 105 106 5 106 107
13.1 16.2 14.8 11.7 9.3 13.2 116.3 (mean) 367 439 472 463
0.03 0.04 0.03 0.02 0.02 0.03 0.26 0.84 1.00 1.08 1.06
T
S F
rates were calculated within the relative uncertainty of 3.0% with the neutrons generated more than 50,000,000. The simulation results are shown in Fig. 6.
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Fig. 8. Calculated neutron spectrum of measuring points U–Z when the source is in normal shielding state (a) and lifted to 15 (b), 22 (c) and 29 cm (d). The neutron energy was divided into 3 groups according to Table 2.
The calculated neutron spectrum of measuring points U–Z is also appended in Fig. 8 when the source was in different shielding states. The spectrum was simplified and represented by the proportion of the above mentioned three groups. For a certain shielding state, the proportion of groups T, S and F remains almost the same for each point, and can be approximated to 65%, 20% and 15% respectively. The simulation results indicate that the overwhelming majority of neutrons escaped from the shield belong to group T with energy below 104 eV.
4.3. Neutron dose-equivalent rates versus neutron count rates As discussed in the above sections that the neutron dose value is closely related to the neutron energy and different energy types of neutrons per unit fluence have different dose conversion coefficient values. Hence, for the low-energy neutron radiation field in our study, the low-level neutron dose equivalent rate values and the neutron count rate values may have some internal connections which can be used to estimate the low level neutron dose values that cannot be measured by the commercial dosimeter directly. For this purpose, the measured neutron dose equivalent rate values as a function of measured neutron count rate values when the source was lifted to 15, 22 and 29 cm are presented in Fig. 9, and the fitting curves are also appended in Fig. 9 by using exponential form. The results indicate that the low level neutron dose equivalent rate values change exponentially with the neutron count rate values when the source is lifted to different heights. According to these results, the low level neutron dose equivalent rate values can be expressed by the neutron count rate values by the exponential formulas shown in Fig. 9. The items of H15, H22, H29 and n15, n22, n29 are the neutron dose equivalent rate values and neutron count rate values when the source is lifted to 15, 22
and 29 cm, respectively. The terms of R-Squared are the goodness of fit of the above formulas which are all above 0.99. It is evident in Figs. 6 and 9 that the simulation results have large deviation with the experimental results. For the dose values, the deviation is mainly contributed by the inherent error of the BH3105 type neutron dose-equivalent meter, where the inherent error is due to the complexity of the measurement itself. It is difficult for a commercial neutron dose-meter to obtain the dose values absolutely accurate. In order to evaluate the reliability and energy response of a neutron dosimeter, the inherent error of 50% to þ100% was recommended as the reference threshold to design the dosimeter by the ICRP organization (ICRP, 2007) and the international standard IEC (IEC-61322, International Standard, 1994; IEC-61005, International Standard, 2014). Therefore, the deviation of neutron dose values between experimental and simulative results is always subsistent and inevitable. For all that, the experimental results are still in quantitative agreement with the results calculated by means of Geant4 simulation within the inherent errors recommended by ICRP and IEC.
5. Conclusions From the intensity distribution of the neutron field measured by He proportional countertube at different measuring points in the source room and its adjacent rooms, the scope of the affected area was obtained. Based on the intensity distribution, the neutrons with lower energy decrease exponentially with the increasing distance approximately. Further, according to the measured count rates and dose equivalent rates in the source room when the source is lifted to different heights, the phenomenon that the dose rate values change exponentially with the count rates is found, and the fitted exponential curve
3
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Based on the results, if the relationship between dose rates and count rates near to the source was obtained, then the dose rates at larger distances to the source can be calculated approximately according to the fitted curve and the measured count rates, where the count rates can be easily acquired by the 3He proportional countertube. This principle can be used to estimate the low-level neutron dose equivalent values at larger distance to the source where the dose rate cannot be measured directly by commercial dosimeters. The results are of great value for low-level neutron radioactivity assessment and radiation protection of curie level Am–Be neutron source.
References
Fig. 9. Neutron dose-equivalent rates versus neutron count rates in the source room outside of the shield; the source is lifted to 15 cm (a), 22 cm (b), and 29 cm (c).
fits very well with the original measured data. Furthermore, the experimental results have a good agreement with the simulative results within the inherent errors recommended by the ICRP and IEC.
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