NaI Scintillation Detector Research Articles

What is a breakdown of continuous component hidden under 4.4-MeV gamma-ray peak from the AmBe source?

Americium-Beryllium (AmBe) is a typical source to be use for a calibration of neutron detection efficiency in underground astroparticle physics experiments. It has been considered to emit a neutron and a 4.4 MeV gamma ray simultaneously, which allows us to select events with a neutron by tagging the 4.4 MeV gamma-ray signal. This time, we focus on the component of high energy tail above 4.4 MeV peak in a deposited energy distribution, under the measurement with inorganic scintillators. In neutron tagging technique with AmBe surrounded by BGO for gamma tagging, there is a possibility of underestimating the tagging efficiency when there are neutrons stopped at BGO. We aim to understand the breakdown of component of the high energy tail from AmBe to contribute reduction of a systematic uncertainty related to the neutron detection efficiency for the astroparticle physics. We will present a result of measurements for neutrons and gamma rays using NaI(Tl) and liquid scintillation detectors.

The use of gamma rays in studying the homogeneity of the alloy (AL-Co-Ni) reinforced with chromium oxide

This research deals with the preparation number of composite materials by casting the base alloy (Al-Co-Ni) by adding chromium oxide (Cr2O3) as reinforcement material, by weight ratios of (5, 10, 15 wt%) for the base alloy, with heat treatment of the prepared alloys. The basic alloys and composite materials were prepared by using solution casting method and manual mixing method to disperse reinforcement grains in the base alloy floor and with pressure of 5 ton in purpose of forming. The materials were melted and poured into molds and suddenly cooled in cold water, after the molten samples were frozen. The attenuation factor of gamma rays was founded the prepared alloys and their homogeneity was examined by shining a narrow beam of single energy gamma rays emitted from the radiating source (137Cs) with different energies (511, 662, 1173, 1284, 1333) keV respectively, gamma ray system (UCS-20) was being used which bounded to NaI(Tl) scintillation detector. The homogeneity of base alloy and composite material was determined by studying the contrast of gamma ray intensity, the linear attenuation coefficient at seven different locations at any samples and the percentage standard deviation. The results showed that the highest value of the linear attenuation coefficient was 0.252 mm-1 for sample [5%Cr2O3+Al+Co+Ni] with the energy (511 keV) and that the lowest value for the ratio of linear attenuation coefficient was 0.062 mm-1 at the energy of sample [15%Cr2O3+Al+Co+Ni] with the energy (1333 keV) indicating that the homogeneity of the sample [15%Cr2O3+Al+Co+Ni] at energy (1333 KeV).

Neutron/gamma radiation shielding characteristics and physical properties of (97.3−x)Pb−xCd–2.7Ag alloys for nuclear radiation applications

In this work, the shielding performance of (97.3–x)Pb–xCd–2.7Ag (x = 10, 18, and 30) ternary alloys against neutrons and gamma rays has been investigated. The microstructure, thermal and mechanical properties of the ternary alloys were examined. The total mass attenuation coefficients, for prepared alloys were determined at 662, 1173, and 1332 keV photon energies using NaI (Tl) scintillation detector. The theoretical values of were calculated using WinXCom program depending on the mixture rule. The estimated values were compared with the measured values for all investigated alloys. Atomic cross-section, σ a, electronic cross-section, σ e, effective atomic number, Z eff , effective electron number, N eff , and GP fitting parameters (b, c, a, Xk, and d) were determined. The exposure buildup factor, EBF, have been also calculated. Fast neutron attenuation for the prepared samples have been investigated via the macroscopic effective removal cross-section () calculation. Also, thermal neutron attenuation has been evaluated via neutron scattering calculator. The results show that the alloys containing 10 and 30% Cd compromise between superior tensile strength and Young modulus, while the pasty range, heat of fusion and ductility decreased with increasing Cd content. Moreover, the prepared ternary alloys have a high attenuation ability for gamma rays as the standard Pb. The increase of Cd ratio also significantly enhances the thermal neutron attenuation by amazing way along with the increase in the attenuation rate of fast neutrons.

Synergistic effects on gamma-ray shielding by novel light-weight nanocomposite materials of bentonite containing nano Bi2O3 additive

In recent years, nanocomposites have gained prominence as potential materials in the γ-radiation shielding and protection areas compared to other conventional materials. In view of this, the present study is devoted to fabricate new nanocomposites, based on raw bentonite clay (RB), as a natural, abundant and low-cost matrix. This was then intercalated with bismuth oxide nanoparticles (Bi2O3 NPs), as a non-toxic nano-additive for the formation of RB/(Bi2O3)x nanocomposites using Bi2O3(x) NPs + raw bentonite clay(1- x), where x = 0, 0.1, 0.2, 0.3 and 0.4 fractional weight. Conventional pressing method at 200 bar was used to prepare nanocomposite disks with different compositions. Fourier-transform infrared spectroscopy (FTIR) and X-ray diffraction (XRD) confirmed the successful preparation of various assembled nanocomposites. Moreover, the as-prepared nanocomposites as well as the Bi2O3 NPs were examined using scanning electron microscopy (SEM) and thermogravimetric analysis (TGA). The SEM images confirmed that the Bi2O3 NPs were successfully incorporated into the bentonite clay layers and its average particle size was in the range 20.74–25.61 nm. Bulk density, effective porosity (%), and heaviness (H) were also determined by different nanofiller loading percentages. Measurements of the γ-ray shielding performance were studied and determined at various γ-ray photon energies emitted from various standard point sources of γ-rays (133Ba,137Cs and 60Co) using NaI(Tl) scintillation detector. Thus, some shielding parameters such as mass attenuation coefficients (μm), effective atomic number (Zeff) and effective electron density (Nel) were measured at 356.01, 662.66, 1173.23 keV, and 1332.50 keV. It was confirmed that the γ-ray shielding properties were enhanced by increasing the loading percentage of Bi2O3 NPs. Therefore, the 40Bi nanocomposite specimen, containing 40.0 wt% Bi2O3 NPs, was characterized to exhibit the highest values of μm (0.1712 at 356.01 keV), Zeff (22.78 at 356.01 keV), and Nel (4.32 at 356.01 keV). It is evident that the addition of Bi2O3 NPs into RB clay leads to vast improvements in γ-ray shielding properties. Such materials would be good candidates for use in construction and civil engineering, as well as in the fabrication of building materials used in medical and nuclear establishments.

Experimental and simulated methods to characterize the response of a scintillator detector

The NaI(Tl) detector is one of the most used in gamma ray spectrometry because it presents high counting efficiency for gamma rays in a wide energy range. This study presents the modeling of a real 1¼ x ¾" NaI(Tl) scintillation detector + photomultiplier using computational simulations carried out by the MCNP6 code, which considers energy resolution. In the mathematical model, the sensitive volume of the detector was adjusted using 241Am and 137Cs radiation sources by means of the absolute photopeak efficiency. Moreover, the MCNP6 code was used to calculate the effective solid angle of the detector in order to obtain the intrinsic efficiency response function. The mathematical model was experimentally validated using calibrated radioactive sources (241Am, 133Ba, 137Cs and 60Co) and the response functions - efficiency curve and energy resolution - were obtained. In addition, an experimental evaluation of the crystal homogeneity of the detector was made using a137Cs radiation source and a 3D printed device developed without stepper motors. The proposed methodology is able to obtain absolute photopeak efficiency in agreement with experimental data with maximum relative error of 5.64%. The gamma scanning procedure indicated that the crystal of the detector remained homogeneous.

Effect of natural radionuclide's in the environment along the Jwalamukhi thrust of Himachal Pradesh, North West Himalayas, India.

This research paper is devoted to measure the activity contents of natural radionuclide, like, radium (226Ra), thorium (232Th) and potassium (40K) in the soil gathered along the Jwalamukhi thrust of Himachal Pradesh, North Western Himalayas, India. NaI(Tl) Scintillator detector was utilized for the estimation of activity content. The activity concentration of 226Ra, 232Th and 40K for some of the soil samples have been observed to be above the global normal mean values. The outcomes acquired for indoor and outdoor effective dosage are well below the normal international and national proposed results. The determined values of external hazard (Hex) for studied locations are less than unity, therefore; samples assembled from these regions are safe from a health hazard point of view and can be utilized as a construction purposes without producing any radio-logical hazard to human beings. The average estimations of radium equivalent activity were found to be within the limits suggested by Organization for Economic Cooperation and Development (OECD). Radon (222Rn) and thoron (220Rn) exhalation rates have also been calculated and discussed.

Calculation of the Effective Energy Release Centerʼs Position of Inorganic Scintillation Detectors for Calibration at Small “Source–Detector” Distances

Inorganic scintillation detectors are widely used to measure of dose rate in the environment due to their high sensitivity to photon radiation. A distinctive feature when using such detectors is the need to take into account of the position of the effective energy release center. This peculiarity is actual when using measuring instruments with inorganic scintillation detectors as working standards during calibration at short “source–detector” distances in conditions of low-background shield or using a facility with protection from external gamma radiation background in the dose rate range from 0.03 to 0.3 μSv/h (μGy/h). The purpose of this work was to calculate the position of the effective energy release center of NaI(Tl) scintillation detectors and to take it into account when working at short “source–detector” distances.An original method of determining the position of the effective energy release center when irradiating the side and end surfaces of inorganic scintillation detector with parallel gamma radiation flux and point gamma radiation sources at small “source–detector” distances using Monte Carlo methods is proposed. The results of calculations of the position of the effective energy release center of NaI(Tl) based detectors of “popular” sizes for the cases of parallel gamma radiation flux and point sources of gamma radiation at small “source–detector” distances are presented. The functional dependences of the position of the effective energy release center of NaI(Tl) based detectors on the distance to the point gamma radiation sources and the energy of gamma radiation sources are presented.As a result of the study it was found that for scintillation NaI(Tl) detectors of medium size (for example, Ø25×40 mm or Ø40×40 mm) the point gamma radiation source located at a distance of 1 m or more, creates a radiation field which does not differ in characteristics from the radiation field created by a parallel flux of gamma radiation. It is shown that approaching the point gamma radiation source to the surface of scintillation detector leads to displacement of the position of the effective energy release center to the surface of the detector.

Variation in cosmic ray flux during Solar Eclipse on 21 June, 2020 at Jaipur, India

Variation in gamma ray flux was recorded during the Annular Solar Eclipse on 21 June, 2020 at Jaipur, India where 88.3% of the solar disk was eclipsed. Two Sodium Iodide NaI (Tl) scintillation detectors were employed for the experiment and flux was recorded in the energy range 300 keV to 2 MeV. The first detector employed to record data for the duration of 3 h and 30 min records a sizeable enhancement at the characteristic energy 1460 keV and the corresponding single escape peak at 949 keV. No substantial variation in the radiations was noticed by the detector employed for recording data for 20 critical minutes during the peak of the eclipse.

Gamma Radiation Shielding Properties of Some Binary Tellurite Glasses

The radiation shielding properties of some binary tellurite glasses have been investigated both theoretically and experimentally. A series of tellurite glasses modified with different types and amounts of modifiers (Nb2O5, B2O3, ZnO, GeO2, WO3, MoO3, and V2O5) have been synthesized using the traditional melting and quenching method. In order to confirm the vitreous nature of the as-synthesized samples, XRD measurements were performed. As for the gamma radiation shielding purposes, linear attenuation coefficient, mass attenuation coefficient, half-value layer and effective atomic number values of the samples were experimentally measured for 662, 1173, and 1332 keV gamma-ray energies emitted from 137Cs, and 60Co radioactive sources. Measurements were performed at narrow-beam transmission geometry by employing a NaI(Tl) scintillation detector. In addition, all these parameters were theoretically calculated using the WinXCOM program in the energy region of 0.015 to 15 MeV. A good agreement was observed between theoretical and experimental values. In order to estimate the gamma-ray shielding effectiveness of the glass samples, all investigated protection parameters were compared with those of commercial RS 253 G-18 glass and some shielding concretes that are commonly used in the nuclear application. It was seen that our synthesized tellurite-based glasses have better shielding properties than standard shielding materials. The results revealed that TeW-3 glass containing 80 mole% TeO2 and 20 mole% WO3 is the most effective shield among the synthesized glasses.

Radioactivity Investigation In Water of Tigris River in Salah Al-Din Governorate, Iraq

The spacific activity of naturally radionuclides (226Ra, 212Pb, and 40K) in water samples of Tigris river was tested using a NaI(Tl) scintillation detector at 30 separate locations in tigris river in Iraq’s Salah Al-Din Governorate. The activity concentrations of activity of 226Ra, 212Pb, 137Cs, and 40K ranged from ( 0.056, 0.702, 0.15, 105.12 Bq/L ) to ( 805.6, 75.12, 27.6, 1149.4 Bq/L ), with an average of ( 315.6, 30.8, 12.04, and 509.2 Bq/L) respectively. The average The radium equivalent activity was found to be (397.4 Bg/kg). The absorbed dose rate in air for the samples was also investigated, and its averagr of (185.6 nGyh-1). For one year, the outdoor annual effective doses with an average of (0.23mSv/y), and the indoor annual effective doses with an average of (0.911mSv/y). The external hazard indices, as well as the internal hazard indices, were found to be more than one higher than the global limit, with average values of 2.04 and 2.89, respectively, higher than the global limit. The International Commission on Radiological Safety and Ionizing Radiation’s Biological Effects Radiation risk factors is used to measure the excess lifetime cancer risk.

Gamma Radiation Shielding and Thermal Properties of the polystyrene /Tungsten (VI) Oxide Composites

In the current study, it was shown that polystyrene (PS) composites holding tungsten (VI) oxide were formed in groups (10% to 50%) by radical polymerization. For this goal, the styrene reacted with BPO for 2.5 hours at 70 °C in tetrahydrofuran (THF) to give the consistent polymer. Synthesized PS were defined by FTIR, GPC (Mn = 7900 g/mol PDI = 1.66), SEM and 1H NMR spectroscopy. PS-tungsten (VI) oxide composites formed between 80-85°C. The gamma-ray of composites was calculated using the NaI(Tl) scintillation detector. Linear attenuation coefficients were also estimated hypothetically by the XCOM platform, taking into account the fundamental analysis of composites and comparing them with empirical outcomes. Among the composites studied, the best protective material was PS -50% W03 with lower absorption thicknesses and a higher linear slimming coefficient. Thermal degradation of unirradiated and irradiated polymer composites has also been examined

Enhancement of Bentonite Materials with Cement for Gamma-Ray Shielding Capability.

The gamma-ray shielding ability of various Bentonite–Cement mixed materials from northeast Egypt have been examined by determining their theoretical and experimental mass attenuation coefficients, μm (cm2g−1), at photon energies of 59.6, 121.78, 344.28, 661.66, 964.13, 1173.23, 1332.5 and 1408.01 keV emitted from 241Am, 137Cs, 152Eu and 60Co point sources. The μm was theoretically calculated using the chemical compositions obtained by Energy Dispersive X-ray Analysis (EDX), while a NaI (Tl) scintillation detector was used to experimentally determine the μm (cm2g−1) of the mixed samples. The theoretical values are in acceptable agreement with the experimental calculations of the XCom software. The linear attenuation coefficient (μ), mean free path (MFP), half-value layer (HVL) and the exposure buildup factor (EBF) were also calculated by knowing the μm values of the examined samples. The gamma-radiation shielding ability of the selected Bentonite–Cement mixed samples have been studied against other puplished shielding materials. Knowledge of various factors such as thermo-chemical stability, availability and water holding capacity of the bentonite–cement mixed samples can be analyzed to determine the effectiveness of the materials to shield gamma rays.

Semi-empirical gamma-ray response function of BGO, NaI(Tl) and LaBr[formula omitted](Ce) scintillation detectors

The response function of the BGO, NaI (Tl) and LaBr3(Ce) scintillation detectors to monoenergetic gamma quanta was built on the basis of Monte Carlo simulations using the GEANT4 toolkit and calibration measurements with gamma radiation sources of different energies. The response function consists of seven components: the first six constitute the detector’s response to a direct hit of monoenergetic gamma radiation and depend on the properties of the detector (size, material, energy resolution, etc.), and the seventh component represents the effect of surrounding materials on the measured gamma spectrum. For each component of the function, the analytical form of the energy dependence is determined and its parameters are found when registering gamma quanta with energies in the range from 0.3 MeV to 10 MeV.

Investigations of various gamma radiation interaction parameters of human tissues and their tissue substitute materials for dosimetric applications

The aim of present work is to examine the radiation interaction properties of muscle, cortical bone, lungs and their tissue equivalent materials in variety of radiation based dosimetric applications. The radiation interaction parameters like photon fluence rate, energy fluence rate, kerma, relative kerma, effective atomic number (Zeff) and computed tomography (CT) number for tissues and their various equivalent materials for energies 60-, 81-, 122-, 184-, 356-,511-, 662-, 834- and 1120-keV respectively are investigated. Effective atomic numbers (Zeff) of tissue equivalent materials have been assigned using anon-destructive experimental technique of multiple back-scattering of 662 keV photons. The gamma radiation flux of 662 keV photons is detected by 3″x 3″ NaI(Tl) scintillation detector. The numbers of multiply backscattered events, for all the tissue equivalent materials (TEM), are found to be increasing with thickness of TEM and finally saturates. The saturation thickness helps in assigning a precise value of effective atomic number Zeff for tissue equivalent materials. Monte Carlo simulations support the results of present experiments. Kerma and relative kerma for tissue equivalent materials are evaluated for various gamma photon energies available from standard radioisotope sources. CT number of these tissues and their equivalent materials are evaluated for the first time for which no data are available for comparison with present results.

A New DCC Software for 4πβ(LS) – γ Coincidence Counting

Coincidence counting (CC) is a widely applied method for radionuclide activity measurement. The 4πβ( LS) -γ coincidence counting is a type of CC that is applied to determine the activity of radionuclides that undergo β-γ decay. A liquid scintillation (LS) counter has the advantages of no self-absorption and simple sample preparation. Hence, this study focuses on the 4πβ( LS) -γ method to measure the activity of LS sources with a standard triple-to-double coincidence ratio (TDCR) system for LS measurements in the β channel and a NaI(Tl) scintillation detector in the γ channel. In this study, high-accuracy digital coincidence counting (DCC) software was developed to determine the activity of radionuclides using this 4πβ( LS) -γ coincidence counting method. The software is characterized by the application of the living time method during true event discrimination and the TDCR technology to estimate the effect of system noise and after-pulses in the β channel. The developed software was validated by comparing the results obtained using the software with standardized <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">60</sup> Co activity measurement at the National Institute of Metrology (NIM) applying previously validated methods. The results showed that the activity measured by the developed DCC software had a tolerable error. It has good potential activity measurement, and it is expected to be applied to calibrate radionuclides with more complex decay diagrams in the future.

Study of radioactive particle tracking using MCNP-X code and artificial neural network

Agitators or mixers are highly used in the chemical, food, pharmaceutical and cosmetic industries. During the fabrication process, the equipment may fail and compromise the appropriate stirring or mixing procedure. Besides that, it is also important to determine the right point of homogeneity of the mixture. Thus, it is very important to have a diagnosis tool for these industrial units to assure the quality of the product and to keep the market competitiveness. The radioactive particle tracking (RPT) technique is widely used in the nuclear field. In this paper, a method based on the principles of RPT is presented. Counts obtained by an array of detectors properly positioned around the unit will be correlated to predict the instantaneous positions occupied by the radioactive particle by means of an appropriate mathematical search location algorithm. Detection geometry developed employs eight NaI(Tl) scintillator detectors and a Cs-137 (662 keV) source with isotropic emission of gamma-rays. The modeling of the detection system is performed using the Monte Carlo Method, by means of the MCNP-X code. In this work, a methodology is presented to predict the position of a radioactive particle to evaluate the performance of agitators in industrial units by means of an Artificial Neural Network.

The shielding properties of the ordinary concrete reinforced with innovative nano polymer particles containing PbO–H3BO3 for dual protection against gamma and neutron radiations

Most of the neutron sources such as nuclear reactors and reaction based sources are considered as a mixed field of radiation where both of neutron and gamma particles are emitted directly from the source. In these cases some special techniques are required for an efficient radiation shield design. Usually different neutron and gamma absorbers are employed via successive layers leading to a massive bulky design. In this study, an innovative nano polymer particle containing simultaneous gamma and neutron absorbers has been proposed to simplify and compact the shield design. The synthesized polymer is based on the H3BO3 as the neutron absorber and the PbO as the gamma ray attenuator, leading to PbO–H3BO3 nano polymer particles. The average size, morphological structure, functional groups and thermal properties of the synthesized polymer were characterized by instrumental techniques. The Monte Carlo simulations (MCNP4C) as well as the experimental methods have been employed to investigate the shielding properties of the ordinary concrete reinforced with PbO–H3BO3 filler particles. The simulation results showed that the employment of nano sized particles has a superior advantage over the micro ones for the neutron shielding. However, for the gamma shielding there are no important differences between the nano and micro sized particles and both of them increase the mass attenuation coefficient, significantly. A semi-saturation effect was observed for weight fractions more than 5 wt% of filler particles. Dozens of concrete slab samples reinforced with nano sized PbO–H3BO3 filler particles have been prepared for experimental studies. The mass attenuation coefficients as well as HVL values of samples were determined using some well-known gamma sources and NaI(Tl) scintillation detector. The results showed that the employment of 5 wt% PbO–H3BO3 reduces the HVL by 38% at 511 keV and 24% at 1332 keV. The experimental results endorsed the simulation study on the effectiveness of proposed innovative nano polymer particles for concrete shielding reinforcement.

Monte Carlo modeling of scintillation detectors for continuous underground radon monitoring

Nuclear simulation methods were applied to two systems that investigate radon transport within geological porous media: a) a laboratory system built to test, under controlled climate conditions, the effect of temperature on radon transport, and b) a field monitoring system comprising gamma and alpha detectors in an abandoned water well. The use of Monte Carlo simulations of NaI and BGO scintillation detectors in continuous underground radon measurements by gamma counting, to estimate the photon flux in the detector volume, is presented.The advantages of shielding side-view NaI detectors were demonstrated for a laboratory system containing ground phosphate rock, including avoiding high counting rates and reducing the effective source volume in radon transport studies. The gross gamma counting procedure was shown to result in a lower uncertainty than spectrometric measurement, by at least a factor of two, despite it being a simpler and more suitable procedure for field measurements.The calculation of simulated source volumes for a BGO detector in a borehole and the measurements in the field support the assumption that the gamma signal comes primarily from radon flowing in the bedrock's air-filled pores. As a practical outcome of this study, positioning the detector a few cm off-center from the borehole's axis increased the gamma counting efficiency; however, measurements in groundwater taken too close to the iron casing had a lower detection efficiency.The conversion factor from the scintillator signal to the radon activity concentration, for the laboratory system, was calculated. Monte Carlo simulations demonstrated the advantages of the gross counting procedures using gamma scintillation detectors in field underground high-frequency radon monitoring.

Gamma-ray spectroscopy with anode pulses of NaI(Tl) detector using a low-cost digitizer system

Recently, digital gamma-ray spectroscopy employing low-cost and publicly available (Commercial off the shelf) digitizers has been frequently used in different studies worldwide. In this paper, we considered the digital methods for gamma-ray spectroscopy in which the anode pulses of the photomultiplier tube (PMT) output in a NaI(Tl) scintillation detector were immediately digitized by a PC sound card. We introduced and developed the methods for gamma-ray spectroscopy of microCurie gamma-ray sources by a sampling rate of 96 kHz. First, at low count rates, the pulse arrival time was determined directly by the raw waveform, and the gamma-ray spectrum was obtained by summing the corresponding values in the samples per pulse. In addition, the gamma-ray spectrum was obtained by an enhanced sampling rate waveform and the pulse arrival time was determined by employing the digital constant fraction discrimination (DCFD) method, where each pulse area was achievable by summing the corresponding values of pulse samples. On the other hand, fitting the appropriate model function on the pulses and obtaining the fitted pulse area were undertaken for gamma-ray spectroscopy. To this end, a non-iterative algorithm to fast fit the Gaussian model function was improved. Moreover, the pile-up correction was performed at different count rates employing the Maximum Likelihood Estimation (MLE) method and Gaussian model function. Also, an approximate method for solving the high run time challenge was identified in the MLE method for long-time waveforms. To reject the pile-up events, a method was introduced based on the calculation of the full-width at half maximum pulses. By applying the proposed rejection method, we achieved an energy resolution of 6.2% at 663 keV gamma-rays and a count rate of 5.3 kcps.

The assessment of usage of epoxy based micro and nano-structured composites enriched with Bi2O3 and WO3 particles for radiation shielding

This study has been devoted to investigate the radiation shielding performance of epoxy-based micro and nano Bi2O3 and WO3 reinforced composites. While the research has been carried out by using experimental data and MCNP6 simulation for the micro-structured composites; the effect of nano Bi2O3 and WO3 particles addition on the radiation shielding property of epoxy has been discussed by using experimental data. While the gamma-ray spectroscopy experiments have been carried by using NaI(Tl) scintillation detector and Ba-133, Cs-137, and Co-60 sources, the MCNP version 6.2 code has been utilized for the simulation. The experimental and MCNP6 simulation results showed good consistency for both particle reinforcement and different incoming photon energies. Particle size effect investigation also indicated that for both kinds of dopant particles, the nano dopant is more successful in attenuating the intensity of incoming photons than the micro dopant even if they are used in the same percentages. In conclusion, it can be suggested that the micro and nano-structured composites having the highest Bi2O3 and WO3 contents have a considerable potential to shield the harmful effects of some of the diagnostic radionuclides utilized in nuclear medicine as well as the radiation imaging machines such as roentgen, mammography or PET scans.