• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2258-2265
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• 2024

The present investigation aims to study the role of Pr₂O₃ on the structural, physical, and radiation shielding properties of a dense cadmium lead borate glass. The XRD was used to affirm the glassy amorphous structure of fabricated sample materials. Moreover, the FTIR was used to record the change in the FT-IR spectra due to the addition of Pr₂O₃ in the wavenumber interval between 400 and 4000 cm^-1. The features of glass surfaces and the elemental analyses for the synthesized Pr₂O₃-reinforced cadmium lead borate glasses were performed using a SEM, supported by an energy-dispersive spectrometer. The γ-ray protection capacity was evaluated using the Monte Carlo method in a wide energy interval ranging between 0.015 and 15 MeV. The linear attenuation coefficient (LAC) at 1 MeV was reduced by a factor of 10 % from 0.372 cm^-1 to 0.340 cm^-1. The decrease in the LAC values negatively affected the other shielding properties such as half-value thickness and the transmission factor. Although the linear attenuation coefficient is decreased slightly with the partial substitution of CdO by Pr₂O₃ compound, the fabricated glass samples still have a high shielding capacity compared to the traditional commercial glasses as well as previous similar reported glasses.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.643-650
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• 2020

The materials with a high processiblity for radiation shielding, in particular for 3D printable materials, are highly demanding for producing robots working in nuclear plants and designing customized personal protection equipment. In this study, we suspend tungsten particles in a polymeric matrix of either PLA or ABS to compose tungsten-polymer composite filaments; PLA and ABS are widely used for conventional FDM-based 3D printing. The weight fraction of tungsten particles can be increased up to 50% without forming macroscopic aggregates. The composite filaments can be used to print 3D architectures with any shape and geometry. To demonstrate one of potential applications, we print parts for robot actuator and assemble them to protect PCB against gamma ray.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3558-3565
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• 2024

The linear attenuation coefficients (LAC) of four soils (Black cotton (S1), Sandy (S2), Clay (S3), and Sandy (S4)) samples were measured at photon energies released from radioisotopes Co⁵⁷ (122 keV), Ba¹³³ (356 keV), ²²Na (511 and 1275 keV), Cs¹³⁷ (662 keV), Mn⁵⁴ (840 keV), and Co⁶⁰ (1330 keV) using a gamma spectrometer includes a NaI (Tl) scintillation detector. The experimental measurements were confirmed utilizing the Monte Carlo N-particle transport code. The linear attenuation coefficient values enhanced from 0.256 cm^-1 to 0.296 cm^-1 (at E_γ of 122 keV), from 0.126 cm^-1 to 0.142 cm^-1 (at E_γ of 662 keV), and from 0.0938 cm^-1 to 0.105 cm^-1 (at E_γ of 1275 keV), raising the (Fe + Mn) concentration from 0.912 wt% to 11.214 wt%, as well as raising the soil samples density from 1.62 g/cm³ to 1.79 g/cm³. The study also shows an enhancement in the half value thickness, transmission factor, radiation protection efficiency and lead's equivalent thickness due to the enrichment of Fe + Mn concentrations within the studied soils. The results show that the Black cotton soil exhibits better shielding properties for γ-ray than the other soils.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.65-70
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• 1988

A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.

• Journal of the Korean Society of Radiology
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• v.2 no.1
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• pp.31-34
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• 2008

The neutron capture spectrum for the light nuclide was very useful to study the nuclear structure. In the present study, the capture gamma-ray from the 27-keV resonance of $^{19}F(n,g)^{20}F$ reaction were measured with an anti-Compton NaI(Tl) spectrometer and the 3-MV Pelletron accelerator of the Research Laboratory for Nuclear Reactors at the Tokyo institute of technology. A neutron Time-of-Flight method was adopted with a 1.5 ns pulsed neutron source by the $^7Li(p,n)^7Be$ reaction. In the present experiment, a Teflon(($CF_2$)n) sample was used The sample was disk with a diameter of 90mm. The thickness of sample was determined so that reasonable counting rates could be obtained and the correction was not so large for the self-shielding and multiple scattering of neutrons in the sample, and was 5mm. The primary gamma-ray transitions were compared with previous measurement of Kenny.

• Journal of Radiation Protection and Research
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• v.6 no.1
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• pp.8-24
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• 1981

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute(ANSI) N666. These data are used to calculated the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from $2.5{\times}10^{-8}$ to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoetiergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be a useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.818-824
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• 2019

In the present work, we determined the gamma-ray attenuation characteristics of eight different polymers(Polyamide (Nylon 6) (PA-6), polyacrylonitrile (PAN), polyvinylidenechloride (PVDC), polyaniline (PANI), polyethyleneterephthalate (PET), polyphenylenesulfide (PPS), polypyrrole (PPy) and polytetrafluoroethylene (PTFE)) using transmission geometry utilizing the high resolution HPGe detector and different radioactive sources in the energy range 81-1333 keV. The experimental linear attenuation coefficient values are compared with theoretical data (WinXCOM data). The linear attenuation coefficient of all polymers reduced quickly with the increase in energy, at the beginning, while decrease more slowly in the region from 267 keV to 835 keV. The effective atomic number of PVDC and PTFE are comparatively higher than the $Z_{eff}$ of the remaining polymers, while PA-6 possesses the lowest effective atomic number. The half value layer results showed that PTFE ($C_2F_4$, highest density) is more effective to attenuate the gamma photons. Also, the theoretical results of macroscopic effective removal cross section for fast neutrons ($\sum_{R}$) were computed to investigate the neutron attenuation characteristics. It is found that the $\sum_{R}$ values of the eight investigated polymers are close and ranged from $0.07058cm^{-1}$ for PVDC to $0.11510cm^{-1}$ for PA-6.

• 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.

• Journal of Radiation Industry
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• v.17 no.1
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• pp.119-125
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• 2023

Burnup of spent nuclear fuel should be determined accurately for the safety storage of spent nuclear fuel. In this study, a gamma detection system was developed as a part of basic research to measure the burnup of spent nuclear fuel, and its performance was evaluated using a calibration source. The prototype of the gamma detection system was based on a semiconductor sensor using a CZT (Cadmium Zinc Telluride). For quantitative evaluation, tests were conducted using ¹³⁷Cs, ¹³⁴Cs and ²⁵²Cf calibration source. In the performance evaluation, Its field applicability was verified by assessing the energy resolution, the detection linearity and the shielding attenuation according to the nuclide.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.19 no.9
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• pp.124-129
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• 2018

This study was conducted to investigate the penetration dose and shielding rates of tungsten shields used in apron material by changing the type of source used in the nuclear medicine department, thickness of shielding material and distance between the source and detector. For the experiment, the source, shield, and detector were arranged in a straight line and measured with an inspector at a height of 100 cm. The highest shielding effect of tungsten was measured for $^{201}Tl$, while $^{123}I$ showed a higher shielding effect than $^{99m}Tc$. For the sources used in the experiment, the penetration dose decreased with distance and the shielding rate was measured with thicker thickness. However, the shielding rate of $^{13}1I$ and $^{18}F$ sources was found to be lower than when there was no shielding at 0.25 mmPb shield. Therefore, even if the radiation shielding effect of tungsten is high, considering the characteristics according to the type of source and the thickness of the shielding material, it may be helpful to reduce the exposure.