• Journal of radiological science and technology
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• v.4 no.1
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• pp.63-71
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• 1981

The metal-plates(Aluminium. Copper, Lead) of change the variation thickness have been penetrated by the collimated beam($450mm{\times}4mm{\phi}$) of Gamma-ray from $^{192}Ir$. Then, the scattered $\gamma$-ray dose in variable angle and the directly transmitted $\gamma$-ray dose were measured using the electrometer of ionization chamber. The results were summarized as follows: 1. Obtained the mass attenuation coefficients of $Al;0.0937cm^2g^{-1},\;Cu;0.0937cm^2g^{-1},\;pb;0.244cm^2g^{-1}$. 2. Total intensity of front scattered $\gamma$-ray follow the order of Al>Cu>pb. 3. The scattered $\gamma$-ray intensity with the lager angle of scattering was saturated after increase rapidly, and the scattering angle of the more larger was decreased. 4. The scattered $\gamma$-ray intensity through plates of aluminium or copper was saturated after increase with thicker scatterer, and the intensity was decreased at the more thicker. But the variation of scattered $\gamma$-ray dose in the lead plate made the fewest than Al and Cu. 5. The ratio of the scattered $\gamma$-ray dose and the directly transmitted $\gamma$-ray dose was saturated after increase with the thicker scatterer, and the scatterer of the more thicker was decreased. Degree of total intensity in these ratios was followed the order of Cu>Al>Pb.

• The Korean Journal of Nuclear Medicine Technology
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• v.17 no.2
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• pp.10-14
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• 2013

Purpose: Due to developed simultaneous PET/MRI, it has become possible to obtain more anatomical image information better than conventional PET/CT. By the way, in the PET/CT, the linear absorption coefficient is measured by X-ray directly. However in case of PET/MRI, the value is not measured from MRI images directly, but is calculated by dividing as 4 segmentation ${\mu}-map$. Therefore, in this paper, we will evaluate the SUV's difference of attenuation correction PET images from PET/MRI and PET/CT. Materials and Methods: Biograph mCT40 (Siemens, Germany), Biograph mMR were used as a PET/CT, PET/MRI scanner. For a phantom study, we used a solid type $^{68}Ge$ source, and a liquid type $^{18}F$ uniformity phantom. By using VIBE-DIXON sequence of PET/MRI, human anatomical structure was divided into air-lung-fat-soft tissue for attenuation correction coefficient. In case of PET/CT, the hounsfield unit of CT was used. By setting the ROI at five places of each PET phantom images that is corrected attenuation, the maximum SUV was measured, evaluated %diff about PET/CT vs. PET/MRI. In clinical study, the 18 patients who underwent simultaneous PET/CT and PET/MRI was selected and set the ROI at background, lung, liver, brain, muscle, fat, bone from the each attenuation correction PET images, and then evaluated, compared by measuring the maximum SUV. Results: For solid $^{68}Ge$ source, SUV from PET/MRI is measured lower 88.55% compared to PET/CT. In case of liquid $^{18}F$ uniform phantom, SUV of PET/MRI as compared to PET/CT is measured low 70.17%. If the clinical study, the background SUV of PET/MRI is same with PET/CT's and the one of lung was higher 2.51%. However, it is measured lower about 32.50, 40.35, 23.92, 13.92, 5.00% at liver, brain, muscle, fat, femoral head. Conclusion: In the case of a CT image, because there is a linear relationship between 511 keV ${\gamma}-ray$ and linear absorption coefficient of X-ray, it is possible to correct directly the attenuation of 511 keV ${\gamma}-ray$ by creating a ${\mu}$map from the CT image. However, in the case of the MRI, because the MRI signal has no relationship at all with linear absorption coefficient of ${\gamma}-ray$, the anatomical structure of the human body is divided into four segmentations to correct the attenuation of ${\gamma}-rays$. Even a number of protons in a bone is too low to make MRI signal and to localize segmentation of ${\mu}-map$. Therefore, to develope a proper sequence for measuring more accurate attenuation coefficient is indeed necessary in the future PET/MRI.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.3051-3057
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• 2021

The concrete is considered as an important radiation shielding material employed widely in nuclear reactors, particle accelerators, laboratory hot cells and other different radiation sources. The present research is dedicated to the shielding properties study of the ordinary concrete reinforced with different weight fractions of lead oxide micro/nano particles. Lead oxide particles were fabricated by chemical synthesis method and their properties including the average size, morphological structure, functional groups and thermal properties were characterized by XRD, FESEM-EDS, FTIR and TGA analysis. The gamma ray mass attenuation coefficient of concrete composites has been calculated and measured by means of the Monte Carlo simulation and experimental methods. The simulation process was based on the use of MCNP Monte Carlo code where the mass attenuation coefficient (μ/ρ) has been calculated as a function of different particle sizes and filler weight fractions. The simulation results showed that the employment of the lead oxide filler particles enhances the mass attenuation coefficient of the ordinary concrete, drastically. On the other hand, there are approximately no differences between micro and nano sized particles. The mass attenuation coefficient was increased by increasing the weight fraction of nanoparticles. However, a semi-saturation effect was observed at concentrations more than 10 wt%. The experimental process was based on the fabrication of concrete slabs filled by different weight fractions of nano lead oxide particles. The mass attenuation coefficients of these slabs were determined at different gamma ray energies using ²²Na, ¹³⁷Cs and ⁶⁰Co sources and NaI (Tl) scintillation detector. The experimental results showed that the HVL parameter of the ordinary concrete reinforced with 5 wt% of nano PbO particles was reduced by 64% at 511 keV and 48% at 1332 keV. Reasonable agreement was obtained between simulation and experimental results and showed that the employment of nano PbO particles is more efficient at low gamma energies up to 1Mev. The proposed concrete is less toxic and could be prepared in block form instead of toxic lead blocks.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.216-223
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• 2017

In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and $10^{th}$ value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2570-2575
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• 2024

The study aims to develop a novel, lead-free, flexible and lightweight composite shielding material against ionizing radiation. For this, it was used bismuth oxide (Bi₂O₃) in RTV-2 silicon matrix. The shielding tests were carried out in both diagnostic X-ray energies and intermediate gamma-ray energy range of up to 662 keV to determine the radiation attenuation properties of this material in terms of attenuation ratio, half value layer, tenth value layer, mean free path and lead equivalency of samples in weight of 30%, 40%, 50% in Bi₂O₃. In the diagnostic X-ray energy range, half value layer, tenth value layer and lead equivalency (in mm Pb) of the produced samples were measured at 80 and 100 kVp narrow beam conditions according to the requirements of EN IEC 61331-1 standard. The results show that lead equivalent values of the produced novel sheets was measured to be 0.16 mm Pb, corresponding to a 6 mm thickness of the flexible sample when it contains 30% wt. Bi₂O₃ in RTV matrix. The experimental findings for durability and flexibility also indicated that this new RTV-based flexible, lead -free shielding composite can be used safely for especially for manufacturing aprons, garments and thyroid guards used in mammography, radiology, nuclear medicine and dental applications in practice.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4093-4097
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• 2021

This study presents a new dead-time measurement method using the gamma attenuation law and generalized dead-time models for nuclear gamma-ray detectors. The dead-time of the NaI(Tl) detection system was obtained to validate the new dead-time determination method using very thin lead and polyethylene absorbers. Non-paralyzing dead-time was found to be 8.39 ㎲, and paralyzing dead-time was found to be 8.35 ㎲ using lead absorber for NaI(Tl) scintillator detection system. These dead-time values are consistent with the previously reported dead-time values for scintillator detection systems. The gamma build-up factor's contribution to the dead-time was neglected because a very thin material was used.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1421-1430
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• 2022

The computation of the solid angle and the detector efficiency is considering to be one of the most important factors during the measuring process for the radioactivity, especially the cylindrical γ-ray NaI(Tl) detectors nowadays have applications in several fields such as industry, hazardous for health, the gamma-ray radiation detectors grow to be the main essential instruments in radiation protection sector. In the present work, a generic numerical simulation method (NSM) for calculating the efficiency of the γ-ray spectrometry setup is established. The formulas are suitable for any type of source-to-detector shape and can be valuable to determine the full-energy peak and the total efficiencies and P/T ratio of cylindrical γ-ray NaI(Tl) detector setup concerning the truncated conical radioactive source. This methodology is based on estimate the path length of γ-ray radiation inside the detector active medium, inside the source itself, and the self-attenuation correction factors, which typically use to correct the sample attenuation of the original geometry source. The calculations can be completed in general by using extra reasonable and complicate analytical and numerical techniques than the standard models; especially the effective solid angle, and the detector efficiency have to be calculated in case of the truncated conical radioactive source studied condition. Moreover, the (NSM) can be used for the straight calculations of the γ-ray detector efficiency after the computation of improvement that need in the case of γ-γ coincidence summing (CS). The (NSM) confirmation of the development created by the efficiency transfer method has been achieved by comparing the results of the measuring truncated conical radioactive source with certified nuclide activities with the γ-ray NaI(Tl) detector, and a good agreement was obtained after corrections of (CS). The methodology can be unlimited to find the theoretical efficiencies and modifications equivalent to any geometry by essential sufficiently the physical selective considered situation.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4142-4149
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• 2021

In the radiation protection application, the metal-polymer composites have been developed for their radiation shielding properties. In this research, the elastomer composites doped by 10 ㎛ and 100nm size of lead, bismuth and tungsten particles as filler with 30 and 60 wt percentages were prepared. To survey the shielding properties of the polymer composites using gamma-ray emitted from 152Eu and 137Cs sources, the gamma flux was measured by using NaI(Tl) detector, then the linear attenuation coefficient was calculated. Also, the Monte Carlo simulation (MCs) method was used. The results showed a direct relationship between the linear attenuation coefficients of the absorbent and filler ratio. Also, the decrease in the particle size of the shielding material in each weight percentage improved the radiation shielding features. When the dimension of the particles was in the order of nano-size, more attenuation was achieved. At low energies used for medical diagnostic X-ray applications due to the predominance of the photoelectric effect, bismuth and lead were suitable selection as filler.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.957-962
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• 2018

The gamma radiation shielding capability (GRSC) of two clay-materials (Ball clay and Kaolin)of Southwestern Nigeria ($7.49^{\circ}N$, $4.55^{\circ}E$) have been investigated by determine theoretically and experimentally the mass attenuation coefficient, ${\mu}/{\rho}(cm^2g^{-1})$ of the clay materials at photon energies of 609.31, 1120.29, 1173.20, 1238.11, 1332.50 and 1764.49 keV emitted from $^{214}Bi$ ore and $^{60}Co$ point source. The mass attenuation coefficients were theoretically evaluated using the elemental compositions of the clay-materials obtained by Particle-Induced X-ray Emission (PIXE) elemental analysis technique as input data for WinXCom software. While gamma ray transmission experiment using Hyper Pure Germanium (HPGe) spectrometer detector to experimentally determine the mass attenuation coefficients, ${\mu}/{\rho}(cm^2g^{-1})$ of the samples. The experimental results are in good agreement with the theoretical calculations of WinXCom software. Linear attenuation coefficient (${\mu}$), half value layer (HVL) and mean free path (MFP) were also evaluated using the obtained ${\mu}/{\rho}$ values for the investigated samples. The GRSC of the selected clay-materials have been compared with other studied shielding materials. The cognizance of various factors such as availability, thermo-chemical stability and water retaining ability by the clay-samples can be analyzed for efficacy of the material for their GRSC.

• Journal of Radiation Protection and Research
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• v.28 no.1
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• pp.19-24
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• 2003

The self-attenuation factor for an $^{198}Au$ sample and the 0.412 MeV gamma-ray penetration ratio in the circular Al-cover of the radiation detector have been determined using an analytical solution and MCNP code. The results show that the self-attenuation factors obtained from the analytical solution coincide with those of MCNP code for all but the Au sample with the relatively larger radius. Then the maximum difference between the two methods appears to be 9 % in the Au sample of 1.5 mm radius. It also is revealed that the analytical solutions of the 0.412 MeV gamma-ray penetration ratio in the Al-cover of 7.62 cm radius are consistent with those of the MCNP code within the standard deviation.