• Nuclear Engineering and Technology
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• v.3 no.4
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• pp.185-197
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• 1971

Korea's TRIGA Mark-Ⅱ reactor was primarily designed in 1950's and was constructed in 1962 for 100 kw thermal output, but it was upgraded to 250 kw in July 1969. Nevertheless, the shield remains unchanged, although the radiation level has increased. The result of computation On this paper shows that, with the existing shield, it is safe for the fast neutrons even after the power upgrading by 2.5 times. It is, however, somewhat dangerous for the gamma rays which are comprised of primary and secondary. For the analysis of the reactor shielding design, an attempt is made for the computation toward the horizontal direction. From theoretical point of view, it can be concluded that some layer of additional shield must be reinforced to the existing concrete in order to be radiologically safe in the reactor hall.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.477-484
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• 2024

Experimental investigations on gamma - rays attenuation parameters and dielectric spectroscopic properties were done on a polymeric mixture with chemical composition (100-x) polyethylene + x basalt, where x = 0, 1, 3, 5, 10, and 20 wt%. Using the melting blending technique,six nanocomposite polymeric samples were prepared. The linear attenuation coefficient µ of each prepared set of samples was measured using a gamma-ray spectrometer including High Purity Germanium detector (HPGe) at energies 662.5, 1173.24, and 1332.51 keV. Based on the measured values of (µ) and sample density, the other effective shielding parameters were calculated. The values of µ showed an increase with increasing the dopant ratios from 0.0 up to 20.0 wt%. In addition, the µ values decreased with the photon's energy. The µ values were found 0.0847 up to 0.1175 cm^-1, 0.0571 up to 0.0855 cm^-1, and 0.0543 up to 0.075 cm^-1 at 662.5, 1173.24, and 1332.51 keV. for B0 up to B20, respectively. The ATR spectroscopy was done on the prepared samples, and a good evidence of adding the filler to the pure polyethylene (HDPE) was obtained. Besides, an enhancement in dielectric constant by insertion of basalt NPs also recorded and can be attributed to the large dielectric constant of basalt compared to pure HDPE.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2919-2926
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• 2023

In this paper, rare earth element (REE)/polyester composites were designed with lanthanum oxide, gadolinium oxide, and lutetium oxide as ray shielding agents, and polyethylene terephthalate (PET) as the base. Monte Carlo simulation was carried out using FLUKA software. We found that the radiation protection performance of the composite is affected by the type and amount of REE; a higher amount of REE equated to a better radiation protection performance of the composite. When the thickness of the composite and total thickness of the REE is constant, the number of superimposed layers inside the composite does not affect its shielding performance. Compared with a single-type REE/PET composite, a mixed-type REE/PET composite has a wider range of γ-ray absorption and better radiation protection performance. When the mass ratio of PET to REE is 2:8 and different types of REE are mixed with equal mass, several 0.2 cm-thick mixed-type REE/PET composites can shield >70% of 60 and 80 KeV γ-rays.

• Journal of the Korean Society of Radiology
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• v.15 no.7
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• pp.957-964
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• 2021

Fluorine-18 used in PET/CT scans is a radioactive isotope that emits positrons, and high energy annihilation gamma rays and beta rays cause exposure to radiation workers. In this study, as part of a plan to reduce the exposure dose of radiation workers working in the Department of Nuclear Medicine, the cause of the low shielding efficiency of Apron for F-18 was identified, and the effectiveness of the Apron double-shielded with acrylic was evaluated. L-Block, Apron+acrylic, Apron, Acrylic+Apron, and Acrylic five shields are used to measure the dose, and the tendencies were compared by performing a Monte Carlo simulation. As a result, it was found that the shielding rate of Apron double shielded with acrylic was about 4 to 8% higher than that of Apron single shielded. To the extent that it does not significantly affect the user's activity, double-shielded personal protective clothing with an appropriate acrylic thickness could help reduce radiation workers' exposure.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.545-552
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• 2002

We developed an ultra low background gamma ray spectrometer particularly suitable for experiment which require lower detection limit. The background of a germanium spectrometer is suppressed by applying active and passive shielding technique at the same time. The active shielding devices consist of plastic scintillating plates of 50 mm thick and anti-coincidence electronic system. The shielding is made of 150 mm thick walls of very low activity lead,20 mm with activity of <10 Bq/kg and 130 mm with activity of <50 Bq/kg. The observed background count rates are 1.2 $s^{-1}$ and 0.36 $s^{-1}$ without and with the active shielding, respectively, overall the energy regions from 30 keV to 3 MeV The cosmic ray induced background is suppressed by a rate of 0.8 $s^{-1}$ at the present work. The detection efficiency curve necessary to obtain the radioactivity of environmental samples has been precisely determined on the energy regions from 80 to 2000 keV with a 10$^3$ ml marinelli beaker sample, consisting of the calibrated radionuclides $^{109}$ Cd, $^{57}$ CO, $^{139}$ Ce, $^{203}$ Hg, $^{113}$ Sn, $^{85}$Sr, $^{137}$ Cs, $^{60}$ Co and $^{88}$ Y. Virtues Of the method are demonstrated by measuring the activity of $^{137}$ Cs contained in the powdered milk.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.470-477
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• 2018

With a highly functional methyl vinyl silicone rubber (VMQ) matrix and filler materials of $B_4C$, PbO, and benzophenone (BP) and through powder surface modification, silicone rubber mixing, and vulcanized molding, a flexible radiation shielding and resistant composite was prepared in the study. The dispersion property of the powder in the matrix filler was improved by powder surface modification. BP was added into the matrix to enhance the radiation resistance performance of the composites. After irradiation, the tensile strength, elongation, and tear strength of the composites decreased, while the Shore hardness of the composites and the crosslinking density of the VMQ matrix increased. Moreover, the composites with BP showed better mechanical properties and smaller crosslinking density than those without BP after irradiation. The initial degradation temperatures of the composites containing BP before and after irradiation were $323.6^{\circ}C$ and $335.3^{\circ}C$, respectively. The transmission of neutrons for a 2-mm thick sample was only 0.12 for an Am-Be neutron source. The transmission of ${\gamma}$-rays with energies of 0.662, 1.173, and 1.332 MeV for 2-cm thick samples were 0.7, 0.782, and 0.795, respectively.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2999-3005
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• 2022

The new experimental setup has been built at the 11b channel of the IBR-2 research reactor at FLNP, JINR, to study the elemental composition of samples by registration of prompt gamma emission during thermal neutron capture. The setup consists of a curved mirror neutron guide and a radiation-resistant HPGe high-purity germanium detector. The detector is surrounded by lead shielding to suppress the natural background gamma level. The sample is placed in a vacuum channel and surrounded by a LiF shield to suppress the gamma background generated by scattered neutrons. This work presents characteristics of the experimental setup. An example of hydrogen concentration determining in a diamond powder made by detonation synthesis is given and on its basis, the sensitivity of the setup is calculated being ~4 ㎍.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3317-3323
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• 2022

Half Value Layer calculations theoretically need prior specification of linear attenuation calculations, since the HVL value is derived by dividing ln(2) by the linear attenuation coefficient. The purpose of this study was to establish a direct computational model for determining HVL, a vital parameter in nuclear radiation safety studies and shielding material design. Accordingly, a typical gamma-ray transmission setup has been modeled using MCNPX (version 2.4.0) general-purpose Monte Carlo code. The MCNPX code's INPUT file was designed with two detection locations for primary and secondary gamma-rays, as well as attenuator material between those detectors. Next, Half Value Layer values of some well-known gamma-ray shielding materials such as lead and ordinary concrete have been calculated throughout a broad gamma-ray energy range. The outcomes were then compared to data from the National Institute of Standards and Technology. The Half Value Layer values obtained from MCNPX were reported to be highly compatible with the HVL values obtained from the NIST standard database. Our results indicate that the developed INPUT file may be utilized for direct computations of Half Value Layer values for nuclear safety assessments as well as medical radiation applications. In conclusion, advanced simulation methods such as the Monte Carlo code are very powerful and useful instruments that should be considered for daily radiation safety measures. The modeled MCNPX input file will be provided to the scientific community upon reasonable request.

• Journal of Radiation Protection and Research
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• v.22 no.3
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• pp.219-226
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• 1997

Ultralow background gamma-ray measurement system was installed to measure and analyze gamma-rays emitted from environmental and swipe samples. The background reduction techniques applied on this system are the passive shielding to surround the HPGe detector, an active external anticosmic shield to shield cosmic-rays and the nitrogen gas supply to minimize the introduction of ubiquitous radon decay nuclei. The performance test result showed that the system background at energies between 50 keV and 2 MeV is reduced about $10^{-2}$ order and the MDA is so low as to be suitable for the environmental sample analysis. But it is appeared that the neutron produced by cosmic-ray increases the background at low energy region.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.171-179
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• 1980

This paper presentss flux-to-dose conversion factors for neutrons and gamma-rays based on the concept of the maximum absorbed dose. Neutron flux-to-does-rate conversion factors for energies from 2.5$\times$10$^{-8}$ to 20 MeV are presented while the conversion factors for gamma-rays are given in the energy range of 0.01 to 15MeV. Flux-to-does-rate conversion factors, which were calculated under the assumption that the radiation energy distribution has nonlinearity in phantom, are different from those values obtained by monoenergetic radiation. Especially, these values obtained here were determined for the cross section libray such as DLC-23, DLC-27, and DLC-31. The flux-to-dose-rate conversion factors obtained in this work are in a good agreement with the values presented by American National Standard Institute (ANSI) N666. These results are used to calculate the dose rate distribution of neutron and gamma-ray in any radiation fields, and will be useful for the radiation shielding analysis, radiation protection and radiation dosimetry concerned with problems of continuous energy distribution.