Measurement and Monte Carlo simulation of γ-ray dose rate in high exposure building materials

Measurement and Monte Carlo simulation of γ-ray dose rate in high exposure building materials 1 first-author 1 corresp Corresponding Author, E-mail: m_hassanzadeh1354@yahoo.com, Corresponding Author, E-mail: m_hassanzadeh1354@yahoo.com, m_hassanzadeh1354@yahoo.com Reactor Research School, Nuclear Science and Technology Institute (NSTRI), P.O.Box:141551339, Tehran, Iran Reactor Research School, Nuclear Science and Technology Institute (NSTRI), P.O.Box:141551339, Tehran, Iran Natural radioactivity radionuclides in building materials, such as 226 Ra, 232 Th and 40 K, cause indoor exposure due to their gamma rays. In this research, in a standard dwelling room (5.0 m×4.0 m×2.8 m), with the floor covered by various granite stones, was set up to simulate the dose rates from the radionuclides using MCNP4C code. Using samples of granite building products in Iran, activities of the 226 Ra, 232 Th and 40 K were measured at 3.8–94.2, 6.5–172.2 and 556.9–1529.2 Bq·kg −1 , respectively. The simulated dose rates were 26.31–184.36 nGy·h −1 , while the measured dose rates were 27.70–204.17 nGy·h −1 . With the results in good agreement, the simulation is suitable for any kind of dwelling places. Radioactivity Building Materials Absorbed dose Experimental MCNP4C. 1. Introduction 1 Indoor exposure to gamma-rays from natural radionuclides in building materials is inherently greater than outdoor exposure [ 1 1 ]. To determine radiological hazards to human health, the natural radioactivity from building materials must be under control. Some researchers measured radioactivity in concrete, granites and sand [ 2 2 ]. Terrestrial origin building materials, such as concrete, cement, brick, sand, aggregate, marble, gypsum and granite, [ 3 3 ] usually contain the uranium and thorium decay series radionuclides, so the radiation exposure arises mainly from 238 U, 232 Th series and 40 K [ 4 4 ]. When the duration of occupancy is taken into account, indoor exposure becomes even more significant [ 5 5 ]. The distribution and concentrations of the parent radium radionuclides in bedrocks of various types vary greatly from type to type. In general, granites have relatively high radium content. Therefore, it is not only important but also feasible to assess the radiological hazard by calculating indoor external dose based on radioactivity measured for building materials [ 6 6 ]. Radium-226, as an alpha emitter with a half-life of 1622 years, is a natural decay product of 238 U series. The other gamma-emitting radionuclides from 226 Ra decay are 214 Pb and 214 Bi. The specific absorbed dose rate in air is mainly affected by parameters of position concerned in room, thickness and density of building materials [ 7 7 - 8 8 7-8 ]. The Monte Carlo code MCNP4C has been used to evaluate the absorbed doses in air. Developed for simulating transports of electrons, neutrons, photons etc ., the code gives an arbitrary three-dimensional configuration of materials in geometric cells. The model has been presented in term of an input file in this code. It encloses the geometry, material, source information and the type of output needed in the form of standard tallies already supplied [ 9 9 ]. In this paper, in a living room geometry, dose rates of 226 Ra, 232 Th and 40 K in high exposure building materials in Iran are measured and simulated using the MCNP4C code. The simulation and measurement results are compared. 2. Materials and Methods 1 2.1 Experimental procedure 2 The collected samples were pulverized, sieved through a 0.2 mm mesh, sealed in standard 1 L Marinelli beakers, dry-weighed and stored for 4 weeks before counting in order to allow the attainment of equilibrium between 226 Ra and 222 Rn and its decay products. In equilibrium, the activity of each daughter is equal to that of the initial isotope of the series [ 8 8 ]. The gamma-ray spectra of the prepared samples were measured using a typical high-resolution gamma spectrometer based on a coaxial P-type shielded high-p urity germanium (HPGe) detector, with a relative photo peak efficiency of 80% and energy resolution of 1.80 keV (FWHM) at 1332 keV, coupled to a high count-rate Multi-Task 16 k MCA card. Commercial software Gamma Genie-2000 was used for data analysis [ 6 6 ]. The 226 Ra activities were calculated from the short-lived daughters 214 Pb (295.2 and 351.9 keV) and 214 Bi (609.3 keV). Similarly, 232 Th activities were measured by taking the mean activity of photo peaks of the daughter nuclides 228 Ac (338.40, 911.07 and 968.90 keV) and 212 Pb (238.63 keV). Activities of 40 K were determined directly from its 1460.83 keV gamma-ray. Quality assurance of the measurements was determined by using of Standard Reference Material IAEA Soil-375 [ 10 10 ]. The density of all the granite samples was estimated at 2580 kg·m −3 in average. The granite stones in markets are usually sized at 30 cm × 50 cm × 3 cm [ 11 11 ]. The granites differ from one type to another in their compositions. Table 1 Table 1 shows the composition of a typical granite stone. 1 1 Fe 2 O 3 1 1 FeO 1 1 CaO 1 1 Na 2 O 1 1 K 2 O 1 1 Al 2 O 3 1 1 SiO 2 1 1 1.22 1 1 1.68 1 1 1.82 1 1 3.69 1 1 4.12 1 1 14.42 1 1 72.04 Calculation of the dose rate conversion factors was done based on the point kernel integration method for floors covered with 3.0 cm thick granite. The free-in-air absorbed dose rate (nGy·h −1 ) in the room center can be expressed as Eq. (1) [ 4 4 ]: <math display="block" id="M1"><mrow><mtext>d</mtext><mi>D</mi><mo>/</mo><mtext>d</mtext><mi>t</mi><mo>=</mo><mn>0.04</mn><msub><mi>A</mi><mtext>K</mtext></msub><mo>+</mo><mn>0.46</mn><msub><mi>A</mi><mrow><mtext>Ra</mtext></mrow></msub><mo>+</mo><mn>0.60</mn><msub><mi>A</mi><mrow><mtext>Th</mtext></mrow></msub><mo>,</mo></mrow></math> where A K , A Ra and A Th are the activity concentrations (Bq·kg −1 ) of 40 K, 226 Ra and 232 Th, respectively. To calculation the effective annual dose rate (mSv), we consider the conversion factor 0.7 Sv·Gy −1 for adult categories. The indoor occupancy factor is 0.8 given by UNSCEAR report and the allowed indoor dose is 1 mSv·y −1 [ 5 5 ]. The effective annual dose rate is by Eq. (2): <math display="block" id="M2"><mrow><mtext>d</mtext><mi>E</mi><mtext>/d</mtext><mi>t</mi><mtext>=</mtext><mrow><mo>(</mo><mrow><mtext>d</mtext><mi>D</mi><mtext>/d</mtext><mi>t</mi></mrow><mo>)</mo></mrow><mtext>×0,7</mtext><mrow><mo>(</mo><mrow><msup><mrow><mtext>Sv·Gy</mtext></mrow><mrow><mtext>-1</mtext></mrow></msup></mrow><mo>)</mo></mrow><mtext>×24×365×0</mtext><mtext>.8 </mtext><mrow><mo>(</mo><mrow><msup><mrow><mtext>h·y</mtext></mrow><mrow><mtext>-1</mtext></mrow></msup></mrow><mo>)</mo></mrow><mo>.</mo></mrow></math> 2.2 Theoretical procedure 2 2.2.1 Room geometry 3 According to the standard living room, a room with dimensions of 4.0 m × 5.0 m × 2.8 m is defined for the MC simulation. Its floor is covered with granite stones of 3.0 cm thick, with the compositions listed in Table 1 Table 1 . The detector is at the room center. 2.2.2 Monte Carlo Simulation 3 By defining necessary parameters of high exposure building materials and room geometry, we can simulate the indoor external gamma-ray exposure and the dose distribution. The interaction of photons with walls and the air is simulated by MCNP4C code. To improve the calculation accuracy, 10–2000 keV gamma-rays of over 1% emission intensity from 226 Ra, 232 Th and 40 K, as summarized in Table 2 Table 2 , are used in the simulation. 1 1 232 Th 1 1 E (keV) 1 1 39 1 1 57.7 1 1 63.7 1 1 84.4 1 1 99.5 1 1 129 1 1 154 1 1 209 1 1 238 1 1 241 1 1 277 1 1 338 1 1 510 1 1 583 1 1 727 1 1 860 1 1 911 1 1 969 1 1 2614 1 1   1 1 Intensity (%) 1 1 1.9 1 1 4.8 1 1 3.8 1 1 1.2 1 1 1.2 1 1 2.4 1 1 6.2 1 1 3.9 1 1 2.9 1 1 4.1 1 1 6.8 1 1 11.4 1 1 21.6 1 1 84.2 1 1 11.8 1 1 12.5 1 1 27.7 1 1 16.6 1 1 100 1 1 226 Ra 1 1 E (keV) 1 1 46 1 1 63 1 1 92 1 1 295 1 1 352 1 1 609 1 1 1120 1 1 1238 1 1 1764 1 1 2204 1 1   1 1   1 1   1 1   1 1   3 1 40 K E (keV) 1 1 1461 1 1   1 1 Intensity (%) 1 1 3.9 1 1 3.8 1 1 5.4 1 1 19.2 1 1 37.2 1 1 46.3 1 1 15.1 1 1 5.9 1 1 15.8 1 1 5.0 1 1   1 1   1 1   1 1   1 1   3 1 Intensity (%) 1 1 10.7 The photon flux and energy deposition in the room center are calculated by using F 4 and F 6 tallies, respectively. This code reads the input file such as the geometry, materials, neutron source etc. Results of interest can be scored by using tallies. A tally is a specification of what should be included in the output. For example, the photons flux ( F 4) through a certain area or the number of photons in a particular energy interval can be calculated by [ 9 9 ]: <math display="block" id="M3"><mrow><mi>F</mi><mn>4</mn><mo>=</mo><mi>C</mi><mo>∫</mo><mi>Φ</mi><mrow><mo>(</mo><mi>E</mi><mo>)</mo></mrow><mtext>d</mtext><mi>E</mi><mo>,</mo></mrow></math> where C is an arbitrary scalar quantity for normalization, Φ ( E ) is the photon flux ( F 4) in cm −2 , and E is incident photon energy. The energy deposition or heating tally ( F 6) is the following track length estimates and can be calculated by: <math display="block" id="M4"><mrow><mi>F</mi><mn>6</mn><mo>=</mo><msub><mi>ρ</mi><mtext>a</mtext></msub><mo>/</mo><msub><mi>ρ</mi><mtext>g</mtext></msub><mo>∫</mo><mi>H</mi><mrow><mo>(</mo><mi>E</mi><mo>)</mo></mrow><mi>Φ</mi><mrow><mo>(</mo><mi>E</mi><mo>)</mo></mrow><mtext>d</mtext><mi>E</mi><mo>,</mo></mrow></math> where, ρ a is the atom density (atoms·barn −1 ·cm −1 ); ρ g is the mass density (g·cm −3 ); and H ( E ) is the heating response, having different meanings, depending upon context as follows: <math display="block" id="M5"><mrow><mi>H</mi><mrow><mo>(</mo><mi>E</mi><mo>)</mo></mrow><mo>=</mo><msub><mi>σ</mi><mi>T</mi></msub><mrow><mo>(</mo><mi>E</mi><mo>)</mo></mrow><msub><mi>H</mi><mrow><mtext>avg</mtext></mrow></msub><mrow><mo>(</mo><mi>E</mi><mo>)</mo></mrow><mo>,</mo></mrow></math> where σ T is total photon cross section, and H avg ( E ) is average energy of exiting gamma-rays for each reaction and for all energies is assumed to be deposited locally. The energy deposition tally ( F 6) is in MeV·g −1 . 3. Results and Discussion 1 3.1 Experimental and Simulation Results 2 The measured activities of 232 Th, 226 Ra and 40 K in the granite samples are 6.5–172.2, 3.8–94.2 and 556.9–1529.2 Bq·kg −1 , respectively. They are shown in Table 3 Table 3 , together with measurement and simulation results of the absorbed dose rate (d D /d t ) and effective annual dose rate (d E /d t ) due to 226 Ra, 232 Th and 40 K in air. * Uncertainties are given within 1 standard deviation. ** Calculation error is under 5%. 1 1 Sample code 1 1 Commercial name 3 1 Activity (Bq·kg −1 ) 2 1 d D /d t (nGy·h −1 ) 2 1 d E /d t (mSv·y −1 ) 1 1   1 1   1 1 232 Th 1 1 226 Ra 1 1 40 K 1 1 Measured * 1 1 Simulated ** 1 1 Measured 1 1 Simulated 1 1 G 1 1 1 Chayan sable 1 1 7.2 1 1 4.8 1 1 561.4 1 1 28.7 ± 1.1 1 1 33.7±0.4 1 1 0.14 1 1 0.17 1 1 G 2 1 1 Tekab 1 1 75.9 1 1 39.2 1 1 1017.8 1 1 101.9 ± 2.4 1 1 99.6±0.3 1 1 0.50 1 1 0.50 1 1 G 3 1 1 Nehbndan birjand 1 1 172.2 1 1 99.2 1 1 1529.2 1 1 204.2 ± 2.9 1 1 175.2±0.8 1 1 1.00 1 1 0.88 1 1 G 4 1 1 Peranshahr 1 1 82.1 1 1 53.3 1 1 964.7 1 1 109.2 ± 1.8 1 1 81.6±0.6 1 1 0.54 1 1 0.41 1 1 G 5 1 1 Torbat hydaryeh 1 1 9.1 1 1 10.8 1 1 647.5 1 1 35.7 ± 1.4 1 1 38.1±0.2 1 1 0.18 1 1 0.19 1 1 G 6 1 1 Natanz 1 1 87.1 1 1 58.3 1 1 1124.7 1 1 120.9 ±2.6 1 1 131.7±0.7 1 1 0.59 1 1 0.66 1 1 G 7 1 1 Morvared mashhad 1 1 65.6 1 1 39.7 1 1 941.6 1 1 92.9 ± 4.1 1 1 47.3±0.5 1 1 0.46 1 1 0.24 1 1 G 8 1 1 Akbatan hamedan 1 1 65.8 1 1 49.6 1 1 1144.3 1 1 105.1 ± 2.2 1 1 71.6±0.3 1 1 0.52 1 1 0.36 1 1 G 9 1 1 Sangeh alamot 1 1 79.6 1 1 43.2 1 1 1101.7 1 1 109.1 ± 2.1 1 1 68.5±0.6 1 1 0.54 1 1 0.34 1 1 G 10 1 1 Garmez yazd 1 1 59.2 1 1 29.9 1 1 1047.2 1 1 89.4 ± 1.7 1 1 96.7±0.4 1 1 0.44 1 1 0.48 1 1 G 11 1 1 Balloch zahedan 1 1 84.9 1 1 41.8 1 1 1121.4 1 1 112.5 ± 1.6 1 1 121.7±0.3 1 1 0.55 1 1 0.61 1 1 G 12 1 1 Morvared sabz 1 1 73.1 1 1 36.8 1 1 1002.3 1 1 98.7 ± 3.1 1 1 70.4±0.5 1 1 0.48 1 1 0.35 1 1 G 13 1 1 khoramdareh 1 1 171.3 1 1 96.7 1 1 1362.7 1 1 196.0 ± 5.8 1 1 184.4±0.8 1 1 0.96 1 1 0.92 1 1 G 14 1 1 Chayan sable 1 1 55.9 1 1 25.3 1 1 811.8 1 1 76.1 ± 1.5 1 1 86.1±0.4 1 1 0.37 1 1 0.43 1 1 G 15 1 1 Tekab 1 1 68.1 1 1 44.3 1 1 1101.4 1 1 102.6 ± 2.3 1 1 67.6±0.3 1 1 0.50 1 1 0.34 1 1 G 16 1 1 Trasheh sfed 1 1 148.9 1 1 89.3 1 1 1341.2 1 1 178.7 ± 3.9 1 1 160.2±0.6 1 1 0.88 1 1 0.80 1 1 G 17 1 1 Morvared sabz 1 1 6.5 1 1 3.8 1 1 556.9 1 1 27.7 ± 1.4 1 1 26.9±0.2 1 1 0.14 1 1 0.13 1 1 G 18 1 1 Hekmtaneh 1 1 58.9 1 1 39.4 1 1 1131.2 1 1 96.4 ± 2.6 1 1 83.4±0.6 1 1 0.47 1 1 0.42 1 1 G 19 1 1 Sangeh lorestan 1 1 9.5 1 1 5.8 1 1 601.8 1 1 32.1 ± 1.1 1 1 26.3±0.4 1 1 0.16 1 1 0.13 1 1 G 20 1 1 Alborz 1 1 166.4 1 1 79.2 1 1 1234.5 1 1 180.9 ± 3.6 1 1 142.6±0.8 1 1 0.89 1 1 0.71 1 1 Average 1 1   1 1 77.4 1 1 44.5 1 1 1017.2 1 1 104.9± 2.5 1 1 90.7 1 1 0.51 1 1 0.45 1 1 Minimum value 1 1   1 1 6.5 1 1 3.8 1 1 556.9 1 1 27.7 1 1 26.3 1 1 0.14 1 1 0.13 1 1 Maximum value 1 1   1 1 172.2 1 1 99.2 1 1 1529.2 1 1 204.2 1 1 184.4 1 1 1.00 1 1 0.92 The absorbed dose rate is calculated by using DFn (Dose Function), DEn (Dose Energy) cards and tally F 4 operations. Suppose one wanted to compute a dose rate of some type associated with a flux tally, either total or by energy group. This feature allows one to enter a point-wise response function (such as flux-to-dose conversion factors) as a function of energy to modify a regular tally. The energy points are specified on the DEn card and the corresponding values of the dose function are identified on the DFn card. The calculated dose rate distributions due to 232 Th, 226 Ra and 40 K in the room floor are shown in Fig 1 Fig 1 . As can be seen, the dose rate in the floor center is the maximum. The dose rate decreases with increasing height (the Z direction, not shown). The simulated dose rates are 26.3–184.4 nGy·h −1 , averaged at 90.7 nGy·h −1 , while the measured dose rates are 27.7–204.2 nGy·h −1 , averaged at 104.9 nGy·h −1 .The two sets of minimum-maximum range and average value overlap closely. Thus the simulation method is suitable to any kind of dwelling places with the use of granite stones. The uncertainties calculation are calculated within 1 σ (standard deviation) and simulation uncertainties calculated fewer than 5 % errors by code. The uncertainty is calculated by: <math display="block" id="M6"><mrow><msub><mi>X</mi><mrow><mtext>mean</mtext></mrow></msub><mo>=</mo><mo stretchy="false">(</mo><mo>∑</mo><msub><mi>X</mi><mi>i</mi></msub><mo stretchy="false">)</mo><mo>/</mo><mi>N</mi><mo>;</mo><mo> </mo><msub><mi>σ</mi><mi>X</mi></msub><mo>=</mo><msup><mrow><mo stretchy="false">[</mo><mo>∑</mo><msup><mrow><mo stretchy="false">(</mo><msub><mi>X</mi><mi>i</mi></msub><mo>−</mo><msub><mi>X</mi><mrow><mtext>mean</mtext></mrow></msub><mo stretchy="false">)</mo></mrow><mn>2</mn></msup><mo>/</mo><mi>N</mi><mo stretchy="false">]</mo></mrow><mrow><mn>1</mn><mo>/</mo><mn>2</mn></mrow></msup><mo>;</mo><mi>X</mi><mo>=</mo><msub><mi>X</mi><mrow><mtext>mean</mtext></mrow></msub><mo>±</mo><msub><mi>σ</mi><mi>X</mi></msub><mi>，</mi></mrow></math> where X mean is the mean value, N is number of data and X i is data. ( i= 1,2,3…, N ). The approximations errors are expressed by: <math display="block" id="M7"><mrow><mi>R</mi><mi>e</mi><mi>l</mi><mrow><mo>(</mo><mrow><msub><mi>X</mi><mi>A</mi></msub></mrow><mo>)</mo></mrow><mo>=</mo><mo>|</mo><mo stretchy="false">(</mo><msub><mi>X</mi><mi>T</mi></msub><mo>−</mo><msub><mi>X</mi><mi>A</mi></msub><mo stretchy="false">)</mo><mo>/</mo><msub><mi>X</mi><mtext>T</mtext></msub><mo>|</mo><mo> </mo><msub><mi>X</mi><mi>T</mi></msub><mo>≠</mo><mn>0</mn><mo>,</mo></mrow></math> where, X T is the true value and X A is the approximate value. The measured 232 Th, 226 Ra and 40 K activities of the samples are used to calculate the effective annual dose rates, which are. 0.14–1.00 mSv·y −1 , while the simulated effective annual dose rates are 0.13 – 0.92 mSv·y −1 . 4. Conclusion 1 The amount of radiation from natural radionuclides of 226 Ra, 232 Th and 40 K in building materials are determined with a standard room of 5.0 m × 4.0 m × 2.8 m with its floor covered by various granite stones produced in Iran. The dose rates are measured by a gamma-ray spectral system with an HPGe detector, and simulated with the MCNP4C code. The two sets of minimum-maximum dose rate range and averaged dose rate show good agreement.. References UNSCEAR , United Nations Scientific Committee on the Effects of Atomic Radiation . Sources and effects of ionizing radiation. Report to the General Assembly with annexes . Volume I , United Nations Publications , ( 2000 ). United Nations Scientific Committee on the Effects of Atomic Radiation European Commission , Radiation Protection 112-Radiological Protection Principles Concerning the Natural Radioactivity of Building Materials . In: Directorate- General: Environment, Nuclear Safety and Civil Protection . Office for Official Publications of the European Communities , ( 1999 ). Radiation Protection 112-Radiological Protection Principles Concerning the Natural Radioactivity of Building Materials E. Cetin , N. Altinsoy , Y. Örgün , Natural radioactivity levels of granites used in Turkey . Radiation protection dosimetry . 151 ( 2 ): 299 - 305 ( 2012 ). doi: 10.1093/rpd/ncs007 http://doi.org/10.1093/rpd/ncs007 Natural radioactivity levels of granites used in Turkey A. Faanu , H. Lawluvi , D.O. Kpeglo et al ., Assessment of natural and anthropogenic radioactivity levels in soils, rocks and water in the vicinity of Chirano Gold Mine in Ghana . Radiation protection dosimetry . 158 ( 1 ): 87 - 99 ( 2014 ). doi: 10.1093/rpd/nct197 http://doi.org/10.1093/rpd/nct197 Assessment of natural and anthropogenic radioactivity levels in soils, rocks and water in the vicinity of Chirano Gold Mine in Ghana S. Dziri , A. Nachab , A. Nourreddine et al ., Experimental and simulated effective dose for some building materials in France . World Journal of Nuclear Science and Technology . 30;33 ( 02 ): 41 ( 2013 ). doi: 10.4236/wjnst.2013.32007 http://doi.org/10.4236/wjnst.2013.32007 Experimental and simulated effective dose for some building materials in France A. Abbasi , Calculation of gamma radiation dose rate and radon concentration due to granites used as building materials in Iran . Radiation protection dosimetry . 155 ( 3 ): 335 - 342 ( 2013 ). doi: 10.1093/rpd/nct003 http://doi.org/10.1093/rpd/nct003 Calculation of gamma radiation dose rate and radon concentration due to granites used as building materials in Iran S. Risica , C. Bolzan , C. Nuccetelli , Radioactivity in building materials: room model analysis and experimental methods . Science of the total environment . 14;272 ( 1 ): 119 - 126 ( 2001 ). doi: 10.1016/S0048-9697(01)00675-1 http://doi.org/10.1016/S0048-9697(01)00675-1 Radioactivity in building materials: room model analysis and experimental methods UNSCEAR , United Nations Scientific Committee on the Effects of Atomic Radiation . Sources and effects of ionizing radiation, Report to the General Assembly with annexes . Volume I , NY : United Nations , ( 2010 ). United Nations Scientific Committee on the Effects of Atomic Radiation J.F. Briesmeister , MCNP4C: Monte Carlo-N-Particle Transport Code System, LA-13709 , ( 2000 ). MCNP4C: Monte Carlo-N-Particle Transport Code System, LA-13709 Canberra, Genie-2000 Version 3.2 , Canberra Industries, Inc , ( 2013 ). Canberra, Genie-2000 N.W El-Dine , A. El-Shershaby , F. Ahmed et al ., Measurement of radioactivity and radon exhalation rate in different kinds of marbles and granites . Applied Radiation and Isotopes , 30;55 ( 6 ): 853 - 60 ( 2001 ). doi: 10.1016/S0969-8043(01)00107-5 http://doi.org/10.1016/S0969-8043(01)00107-5 Measurement of radioactivity and radon exhalation rate in different kinds of marbles and granites The authors would like to thank the simulation group in (NSTRI) of Iran. 10.1007/s41365-016-0171-x 2016-05-03 2016-08-28 2016-08-29 2016-12-26 2017-02-01 2021-04-11 © China Science Publishing & Media Ltd. (Science Press), Shanghai Institute of Applied Physics, the Chinese Academy of Sciences, ChineseNuclear Society and Springer Nature Singapore Pte Ltd. 2017 This is a post-peer-review, pre-copyedit version of an article published in Nuclear Science and Techniques. The final authenticated version is available online at: http://dx.doi.org/10.1007/s41365-016-0171-x http://dx.doi.org/10.1007/s41365-016-0171-x . 2017 Nuclear Science and Techniques 02 28