Development of a neutron detector for radiation protection monitoring

Nội dung text: Development of a neutron detector for radiation protection monitoring

1. Nuclear Science and Technology, Vol.10, No. 4 (2020), pp. 41-47 Development of a neutron detector for radiation protection monitoring Mai Van Dien*, Nguyen Duc Tuan, Nguyen Ngoc Quynh, Vu Trung Tan, Le Ngoc Thiem, Vu Van Tien Institute for Nuclear Science and Technology, VINATOM, Vietnam *Corresponding email: maivandien.inst@gmail.com (Received 28 October 2020, accepted 28 December 2020) Abstract: The paper presents the results of the development of a neutron detector for radiation protection purposes. Monte Carlo simulations, using MCNP5 code, were performed to optimize the configuration of the neutron detector. The developed detector consists of a 3He proportional counter embedded in a multi-layer moderator made of high-density polyethylene (HDPE) and Cadmium. The characteristics of the developed neutron detector including neutron fluence response and ambient dose equivalent response were calculated, analyzed and compared with those from other neutron survey meters. The simulation model and computed results were assessed through experimental measurements at the Secondary Standards Dosimetry Laboratory of the Institute for Nuclear Science and Technology (INST). A good agreement between the simulated and experimental results was observed within 9.3% for 241Am-Be source and four simulated workplace neutron fields. Keywords: developed neutron detector, neutron fluence response, ambient dose equivalent response. I. INTRODUCTION (ICRP) for area monitoring, the ambient dose equivalent, H*(10), is used as an In recent years, with the increase in the approximation of the protection quantity in use of radioactive sources, especially the radiation measurements of external exposure applications of neutrons in industry and [3]. The mensuration of the ambient dose medicine, radiation protection monitoring equivalent will furnish the necessary continues to be a subject of increasing general information for controlling the radiation at interest. Apart from the main sources of workplaces and definition of controlled or neutrons such as sealed radio-isotopic sources, forbidden areas. nuclear reactors, and neutron generators, neutrons are also encountered in the high Neutron survey meters are employed to energy particle accelerator systems. The measure the neutron ambient dose equivalent radiation field outside the shielding of rate. They are designed based on single or accelerators is frequently dominated by the multiple detectors enclosed by the single or neutron component, which has complex and multi-layer moderator to have a response non-uniform energy distributions [1]. Exposure function that is independent of the energy and to free neutrons can be hazardous since the direction of incident radiation [4, 5]. interaction of neutrons with molecules in the Depending on the measured energy range, body can lead to chromosome damage and neutron survey meters are classified in two adversely affects human health [2]. According main groups: conventional neutron survey to the recommendation of the International meters (e.g., LB 6411 [6], TPS 451C [5]) and Commission on Radiological Protection extended neutron survey meters (e.g., LINUS ©2020 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute
2. DEVELOPMENT OF A NEUTRON DETECTOR FOR RADIATION PROTECTION MONITORING [7], WENDI [8], LUPIN [9], PRESCILA [10]). configuration of the neutron detector. The In the first group, a combination of Cadmium fluence response function of the developed (or/and Boron) and high hydrogen neutron detector was calculated and validated concentration materials are used as filters and by experimental measurements at the moderators to measure neutrons in the range Secondary Standards Dosimetry Laboratory from thermal to 15 MeV. Meanwhile, by of the Institute for Nuclear Science and adding heavy metals such as lead, copper or Technology (INST). The H*(10) response tungsten to the moderator, extended neutron was also investigated in comparison with survey meters are capable of measuring the those from other neutron survey meters to ambient dose equivalent rate for a neutron field confirm the proper operation of the of energy up to several hundred MeV through developed neutron detector. spallation reactions (n, xn). Although in reality most of the instruments encounter many II. MATERIAL AND METHOD difficulties at energies between 1 keV and 200 A. Helium proportional counter keV, where over-estimations by up to a factor of ten are rather the normal case, however, they In this work, a cylindrical are still widely used for radiation monitoring proportional counter was used. The counter, due to their advantage capability of measuring supplied by the Centronic UK Ltd, has and displaying the neutron ambient dose dimensions of 2.54 cm in diameter and 15 equivalent rate in real-time. cm in length. It is filled with gas at 2 In this study, the neutron detector was atm, corresponding to 5.02 x 1019 atoms/cm3 designed by following the principle of at room temperature. The cathode shell is conventional neutron survey meters. Monte made of stainless steel with a thickness of 1 Carlo N-Particle simulations (using MCNP5 mm. Figure 1 shows the sketch of the 3He code [11]) were applied to optimize the proportional counter. Fig. 1. The structure of the 3He proportional counter (dimensions in mm) B. Design principle ∫ The conventional true value of neutron ambient dose equivalent rate, (in Where, is the incident spectral pSv/h) measured by neutron survey meters, is neutron fluence rate (in h-1.cm-2); is the expressed as Eq. (1). neutron fluence-to-dose equivalent 42
3. MAI VAN DIEN et al. conversion factor (in pSv.cm2), given by polyethylene (HDPE). The thickness of the ICRP 74 [3]. moderator was gradually increased from 1 The ambient dose equivalent rate can be cm to 14 cm. Obviously, with the increase of calculated by using the Eq. (1) if the neutron polyethylene thickness, the fluence response energy spectrum is known. A typical method of the neutron detector is significantly using the Bonner sphere spectrometer can be improved. However, it should be noted that applied for measuring the neutron spectrum, the large thickness of the moderator makes it but it requires long measuring time and a difficult to use in the workplace. In addition, complex deconvolution algorithm [12]. An the helium proportional counter is extremely alternative approach to be applied to most sensitive to thermal neutrons, leading to a neutron survey meters is to design the neutron drastic over-estimation of the H*(10) at detector to have a fluence response function intermediate energies. To reduce their effect, with a similar form as the neutron fluence-to- the helium proportional counter must be dose equivalent conversion function given by wrapped by a perforated cadmium shell. The ICRP 74. This method allows determining the parameters including the diameter of the neutron ambient dose equivalent rate through moderator, position and perforated fraction the reading number recorded without regard to of the cadmium shell were adjusted the energy of incident radiation. iteratively until a best fluence response function is achieved. To enhance the Suppose that: and are the accuracy and reliability of simulation results, reading number (count rate) and the fluence measurements were also carried out to check 2 response (in cm ) of the neutron detector, the density and the purity of materials. The respectively. The interrelation between these densities of polyethylene and cadmium were quantities and the incident spectral neutron determined to be 0.95 g/cm3 and 8.72 g/cm3, fluence rate is expressed by Eq. (2): respectively. These values have been obtained from measurements of cylindrical ∫ samples taken from the same batch of polyethylene and cadmium used for design. From Eq. (1) and Eq. (2), it is obvious The purity of the materials is confirmed by that if the fluence responses, , have the the X-Ray fluorescence (XRF) method, same energy dependence as the neutron giving a result greater than 99 % as stated by fluence-to-dose equivalent conversion the manufacturer. factors, , then the reading of a neutron detector can be calibrated based on a In MCNP5 simulations, the fluence conventional true value of with a response, , is determined through the calibration factor of as given by Eq. (3). number of capture reactions, , generated in the active volume of the proportional counter. This number is obtained To achieve detector configuration from F4 tally and the associated FM card by according to the above principle, first of all, using the appropriate multiplication factor. The simulations were performed to evaluate the neutron cross-section was taken from the fluence response of the 3He proportional Evaluated Nuclear Data Files library (i.e., counter with various moderator ENDF/B-VI [13]). The S(α,β) treatment, configurations using the high-density poly.01t [11], was used to account for the 43
4. DEVELOPMENT OF A NEUTRON DETECTOR FOR RADIATION PROTECTION MONITORING carbon and hydrogen chemical binding at room appropriate response function as the temperature. For each configuration of the recommendation of IEC 61005-2014 [15] was neutron detector, simulations were performed selected for designing the neutron detector. with 29 mono-energetic sources ranging from 10- Figure 2 depicts the structure of the developed 9 MeV to 15 MeV. In order to minimize neutron detector, in which the 3He statistical errors of MCNP5 outputs, the variance proportional counter is surrounded by a 6 cm reduction techniques based on the geometry thick outer polyethylene moderator, a 2.7 cm splitting method were applied [14]. The source thick inner polyethylene moderator, and a 3 particle histories were also considered to keep mm thick cadmium shell. In order to extend statistical uncertainties below 3%. the response to thermal neutrons, the cadmium shell is perforated so that the III. RESULT AND DISCUSSION opening accounts for 10% of the surface area of the 3He counter. The self-developed A. The developed neutron detector neutron detector is about 6 kg weigh with the Based on the simulation results, the outer dimensions of 20.5 cm in diameter and configuration of the neutron detector with an 24.5 cm in length. Fig. 2. The developed neutron detector configuration (from the left to the right): (1) side view, (2) top view, and (3) outer view - Neutron fluence response range from 50 keV to 10 MeV. At energies The simulation results are given in between 10 eV and 50 keV and above 10 MeV, Figure 3, which shows the variation of the there is a significant difference between the fluence response as a function of incident shape of the neutron fluence response function neutron energy. In general, fluence responses and the recommended ICRP 74 curve. of the developed detector closely follow the However, it should be noted that this behavior shape of the neutron fluence-to-dose equivalent is a common feature of conventional neutron conversion curve, [3, 15], in the energy survey meters. 44
5. MAI VAN DIEN et al. - Ambient dose equivalent response The developed neutron detector over- estimates the neutron ambient dose To investigate the variation of ambient equivalent rate by a factor between 1.2 and dose equivalent due to incident neutron energy, 5.3 in the energy region from 10 eV to 50 computed fluence responses were converted to keV and under-estimates the neutron ambient responses and then normalized to the dose equivalent rate from 30% to 50% at 241Am-Be source. The response is energies above 10 MeV. It can be seen that defined as the ratio between the fluence the relative response of the response and the corresponding neutron fluence- developed neutron detector has a similar to-ambient dose equivalent conversion factor. tendency with those of other commercial As a result, Figure 4 shows the relative H*(10) neutron survey meters. These values also response function of the developed neutron comply with the recommendations of the detector (normalized to ) together international standard IEC 61005 [15]. That with those obtained from other neutron survey means the performance of the developed meters (i.e. the Aloka TPS-451C, Hitachi Co, neutron detector over the range from thermal Ltd.; the NDN1, Fuji Electric Co, Ltd; and the to 15 MeV is comparable to that of NSN1, Fuji Electric Co, Ltd [5]). commercial neutron survey meters. 10 100 Aloka TPS 451-C ) NSN1 2 ) NDN1 cm self-designed neutron det. ( 1 Am-Be 10 241 0.1 1 0.01 normalized to ( Relativeresponse H*(10) Fluence response response Fluence 1E-3 0.1 1E-8 1E-6 1E-4 0.01 1 100 1E-8 1E-6 1E-4 0.01 1 100 Energy (MeV) Energy (MeV) Fig 3. The variation of the fluence response as a Fig. 4. H*(10) responses of the developed neutron function of neutron energy detector and three commercial neutron survey meters (J.Saegusa et al., [5]). The results are normalized at 241Am-Be response Table I. The fluence response obtained from MCNP calculations and measurements for 241Am-Be source and four simulated workplace neutron fields 2 Neutron fluence response (cm ) Relative deviation Neutron source MCNP Calculation Measurement (%) 241Am-Be 2.55 ± 0.07 2.81 ± 0.09 9.25 20 cm Mod 1.87 ± 0.03 2.04 ± 0.06 8.33 25 cm Mod. 1.71 ± 0.04 1.88 ± 0.05 9.04 30 cm Mod. 1.66 ± 0.03 1.81 ± 0.06 8.29 35 cm Mod. 1.63 ± 0.02 1.72 ± 0.06 5.23 45
6. DEVELOPMENT OF A NEUTRON DETECTOR FOR RADIATION PROTECTION MONITORING B. Experimental validation calculation model and simulated results given in section III.1 are reliable. In order to validate the simulated results, measurements are usually performed in IV. CONCLUSIONS standardized mono-energetic neutron fields as recommendations of IEC 61005-2014 [15]. A neutron detector, consists of a However, it presents many difficulties for proportional counter and a multi-layer testing in a wide range of energy due to the moderator, has been developed to measure limitation of experimental conditions in neutrons in the energy range from thermal to Vietnam. To achieve meaningful results, the 15 MeV. The instrument has dimensions of simulation model was corroborated via 20.5 x 24.5 cm and weighs about 6 kg. The experimental measurements at the Secondary characteristics of the neutron fluence Standards Dosimetry Laboratory of INST. A response and the ambient dose equivalent comparison of the neutron fluence response response were investigated using MCNP5 was performed between MCNP5 simulations simulations and validated by experiments. 241 and experiments for Am-Be source and four The results show that the neutron detector is simulated workplace neutron fields. The suitable for neutron measurements and the 241 emission rate of the Am-Be source was radiation safety assessment. 1.287 x 107 (n/s) with an expanded uncertainty of about 2.9% (k=2). Simulated workplace ACKNOWLEDGEMENT neutron fields are formed by moderating 241Am-Be source with 4 polyethylene spheres The authors of this article gratefully with diameters of 20 cm, 25 cm, 30 cm, and 35 acknowledge the financial support of the cm. The characteristics of neutron calibration Institute for Nuclear Science and fields were described in detail in the Ref. [16, Technology (INST) and Vietnam Atomic 17]. In experiments, the developed neutron Energy Institute (VINATOM) under grant detector was placed at 150 cm from the source number CS/20/04-02. and connected to a read-out system including REFERENCES pre-amplifier, shaping amplifier, and single- channel analyzer (SCA). In order to eliminate [1]. K. Polaczek-Grelik, P. Gandor, et al., “Neutron the gamma interference and overall electronic Field Outside the Room of Linear Medical noise, the lower-level discrimination of SCA is Accelerator”, Inter.l Symp. on Bio. Eng. and applied by an appropriate value of about 300 Med. Phy, Vol.38, 139-141, 2012. mV. The measurement data are transmitted to a [2]. Isheeta Seth, L. Jeffrey et al., “Neutron personal computer via RS-232 communication. Exposures in Human Cells: Bystander Effect The fluence response of the neutron and Relative Biological Effectiveness”, PloS One, 2014. detector is determined by the ratio between the count rate and the incident neutron fluence [3]. ICRP, “The International Commission on rate, as given by Eq. (2). The obtained results Radiological Protection”, Publication 74, 1996. are presented in Table I, which shows the [4]. T.M. Oakes, S.L. Bellinger et al., “An accurate relative deviation between simulations and and portable solid state neutron rem meter”, experiments within ± 9.3 %. The consistency Nuclear Instruments and Methods in Physics between these results implies that the Research A, Vol 719, 6–12, 2013. 46
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