• Journal of the Korea institute for structural maintenance and inspection
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• v.23 no.7
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• pp.172-180
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• 2019

In this study, the microstructure and gamma-ray shielding efficiency of CRT glass mortar specimen were evaluated with replacement ratio and material properties. The results show that as the replacement ratio of CRT waste glass increases, the volume of pores with diameters below 50 nm and above 400 nm is increased. Also, the half-value layer of CRT glass mortar decreased with the increasing of linear attenuation coefficient. In addition, compressive and flexural strength were reduced when CRT waste glass was replaced as the fine aggregate, but the mechanical performance of CRT mortar specimen could be obtained by substitution of the mineral admixture.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.792-800
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• 2017

The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma $({\gamma})-rays$. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, ${\gamma}-ray$ shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV-15 MeV) and optical thickness (OT). The study was performed by computing various ${\gamma}-ray$ shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1932-1940
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• 2024

The current work deals with investigation of the gamma ray shielding properties for various natural stones from Vietnam to be applied in the radiation shielding applications. The physical and chemical properties affecting the γ-ray shielding performance were examined. The MH-300A density meter was utilized to measure the density (ρ, g/cm³) of stone samples, as well as the chemical composition of Vietnamese natural stones was measured using the X-ray fluorescence analyzer (Olympus X-5000). The study shows that the increase in Fe + Ca concentrations within the stone samples increases their density (from 2.48 to 2.86 g/cm³) accompanied with a reduction in the porosity (Φ, %) (from 8.23 to 0.15%) and water absorption (K, %) (from 3.42 to 0.05%) factors. Additionally, the increase in Fe + Ca concentrations increases the linear attenuation coefficient (μ, cm^-1) of the studied stones, where the Vietnamese marble stone (M 3.1) with the highest Fe + Ca concentration (65.97 wt%) has the highest linear attenuation coefficient which varied between 3.781 and 0.155 cm^-1 with raising the gamma ray energy between 0.040 and 1.332 MeV.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1633-1637
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• 2019

While considering the photon attenuation coefficient (${\mu}$) and its related parameters for photons shielding, it is necessary to account for its transmitted and reflected photons energy spectra and dose contribution. Monte Carlo simulation was used to study the efficiency of clay ($1.99g\;cm^{-3}$) as a shielding material below 150 keV photon. Am-241 gamma source and an X-ray of 150 kVp were calculated. The calculated value of ${\mu}$ for Am-241 is higher within 5.61% compared to theoretical value for a single-energy photon. The calculated half-value layer (HVL) is 0.9335 cm, which is lower than that of ordinary concrete for X-ray of 150 kVp. A thickness of 2 cm clay was adequate to attenuate 90% and 85% of the incident photons from Am-241 and X-ray of 150 kVp, respectively. The same thickness of 2 cm could shield the gamma source dose rate of Am-241 (1 MBq) down to $0.0528{\mu}Sv/hr$. For X-ray of 150 kVp, photons below 60 keV were significantly decreased with 2 cm clay and a dose rate reduction by ~80%. The contribution of reflected photons and dose from the clay is negligible for both sources.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2740-2747
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• 2024

The sintering process acquired ferromagnetic copper ferrite ceramic material with a small concentration of Ni ion at 1100 ℃ for 1 h. Previously, copper ferrite with Ni proportions powder was acquired by the wet chemical process according to the relation CuFe_2-xNi_xO₄ where x takes values 0.0, 0.015, 0.03, 0.04, and 0.05. The role of Ni ion in the copper ferrite structure was investigated by X-ray analysis, Scanning electron microscope, EDX analysis, and density measurements. The gamma-ray shielding properties for the fabricated CuFeNiO ceramics samples were evaluated using the Monte Carlo simulation method. The obtained results show an enhancement in the linear attenuation coefficient for the fabricated ceramics with increasing the insertions of Ni ions within the fabricated samples, where increasing the Ni ions concentration between 0 and 1.19 wt% increases the linear attenuation by between 1.581 and 1.771 cm^-1 (at 0.103 MeV), 0.304-0.338 cm^-1 (at 0.662 MeV), and 0.160-0.178 cm^-1 (at 2.506 MeV), respectively. Simultaneously, the radiation protection efficiency for a 1 cm thickness of the fabricated samples increased between 14.8 and 16.3% with increasing the Ni ions between 0 and 1.19 wt%. Although the Ni doping concentration does not exceed 1.5 wt% of the total composition of the fabricated ceramics, the shielding capacity of the fabricated ceramics was enhanced by more than 11%, along the studied energy interval. Therefore, the fabricated samples can be used in gamma-ray shielding applications.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.275-284
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• 2006

In the measurement of gamma rays emitted from the nuclide in the radioactive waste drum, to analyze the nuclide concentration accurately, it is necessary to use the proper calibration standards and to correct for the attenuation of the gamma rays. Two drums having a different density were used to analyze the nuclide concentration inside the drum in this study. After carrying out the system calibration, we measured the gamma rays emitted from the standard source inside the model drum with changing the distance between the drum and the detector. The measured values were corrected with the three kinds of gamma attenuation correction methode, as a results, the error was less than 10 % in the low density drum and less than 25 % in the high density drum. The measured activity in the short distance was more accruable than in the long distance. The transmission correction for the mass attenuation showed good results(very Low error) compared to the mean density and the differential peak correction method.

• Proceedings of the KIEE Conference
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• 1983.07a
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• pp.118-119
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• 1983

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.2025-2037
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• 2021

The influence of lead cations on natural quartz (QZ) from Egypt as a glass shielding material for the composition with nominal formula (10Na₂O - (90 - x) QZ - xPbO (where x = 30, 35, 40, 45 and 50 mol %)) was examined. The studied samples are synthesized via the melt quenching method at 1050 ℃. The X-ray diffraction XRD patterns were confirmed the glass nature for studied samples. Moreover, the optical properties, and the transparency for all compositions were examined by UV-Vis spectroscopy. Also, the major elemental composition of the natural quartz were estimated via the X-ray fluorescence (XRF) technique. Further, the density and molar volume were determined. Furthermore, the nuclear shielding parameters such as, mass attenuation coefficient, effective atomic number, electronic density, the total atomic, and electronic cross sections as well as the mean free path, and the half value layer with different gamma ray energies (81 keV-1407 keV) were calculated. Besides, the results showed that the shielding behavior towards the gamma ray radiation for all glass samples was increased as the increment in PbO concentration in the glass system.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2253-2261
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• 2022

In this study, nano-scaled shielding materials were assembled and fabricated by doping different weight percentages of Nano-mercuric oxide (N-HgO) into Nano-Bentonite (N-Bent) based on using (100-x% N-Bent + x% N-HgO, x = 10, 20, 30, and 40 wt %). The fabricated N-HgO/N-Bent nanocomposites were characterized by FT-IR, XRD, and SEM and evaluated to evaluate their shielding properties toward gamma radiation by using four different γ-ray energies form three point sources; 356 keV from ¹³³Ba, 662 keV from ¹³⁷Cs as well as 1173, and 1332 keV from ⁶⁰Co. The γ-rays mass attenuation coefficients were plotted as a function of the doped N-HgO concentrations into N-HgO/N-Bent nanocomposites. The computed values of mass attenuation coefficients (µ_m), effective atomic number (Z_eff) and electron density (N_el) by the as-prepared samples were found to increase, while the half value layer (HVL) and mean free path (MFP) were identified to decrease upon increasing the N-HgO contents. It was concluded also that the increase in N-HgO concentration led to a direct increase in the mass attenuation coefficient from 0.10 to 0.17 cm²/g at 356 keV and from 0.08 to 0.09 cm²/g at 662 keV. However, a slight increase was observed in the identified mass attenuation coefficients at (1172 and 1332 keV).

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2002.07a
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• pp.25-37
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• 2002

The most important industrial application of gamma radiation in characterizing green compacts is the determination of the density. Examples are given where this method is applied in manufacturing technical components in powder metallurgy. The requirements imposed by modern quality management systems and operation by the workforce in industrial production are described. The accuracy of measurement achieved with this method is demonstrated and a comparison is given with other test methods to measure the density. The advantages and limitations of gamma ray densitometry are outlined. The gamma ray densitometer measures the attenuation of gamma radiation penetrating the test parts (Fig. 1). As the capability of compacts to absorb this type of radiation depends on their density, the attenuation of gamma radiation can serve as a measure of the density. The volume of the part being tested is defined by the size of the aperture screeniing out the radiation. It is a channel with the cross section of the aperture whose length is the height of the test part. The intensity of the radiation identified by the detector is the quantity used to determine the material density. Gamma ray densitometry can equally be performed on green compacts as well as on sintered components. Neither special preparation of test parts nor skilled personnel is required to perform the measurement; neither liquids nor other harmful substances are involved. When parts are exhibiting local density variations, which is normally the case in powder compaction, sectional densities can be determined in different parts of the sample without cutting it into pieces. The test is non-destructive, i.e. the parts can still be used after the measurement and do not have to be scrapped. The measurement is controlled by a special PC based software. All results are available for further processing by in-house quality documentation and supervision of measurements. Tool setting for multi-level components can be much improved by using this test method. When a densitometer is installed on the press shop floor, it can be operated by the tool setter himself. Then he can return to the press and immediately implement the corrections. Transfer of sample parts to the lab for density testing can be eliminated and results for the correction of tool settings are more readily available. This helps to reduce the time required for tool setting and clearly improves the productivity of powder presses. The range of materials where this method can be successfully applied covers almost the entire periodic system of the elements. It reaches from the light elements such as graphite via light metals (AI, Mg, Li, Ti) and their alloys, ceramics ($AI_20_3$, SiC, Si_3N_4, $Zr0_2$, ...), magnetic materials (hard and soft ferrites, AlNiCo, Nd-Fe-B, ...), metals including iron and alloy steels, Cu, Ni and Co based alloys to refractory and heavy metals (W, Mo, ...) as well as hardmetals. The gamma radiation required for the measurement is generated by radioactive sources which are produced by nuclear technology. These nuclear materials are safely encapsulated in stainless steel capsules so that no radioactive material can escape from the protective shielding container. The gamma ray densitometer is subject to the strict regulations for the use of radioactive materials. The radiation shield is so effective that there is no elevation of the natural radiation level outside the instrument. Personal dosimetry by the operating personnel is not required. Even in case of malfunction, loss of power and incorrect operation, the escape of gamma radiation from the instrument is positively prevented.