• Journal of the Korean Society of Radiology
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• v.7 no.1
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• pp.31-36
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• 2013

The radioactive gas radon ($^{222}Rn$), which is generated from the decay process of uranium ($^{238}U$) originating from the soil of more than 85 percent higher the porosity of the soil, the soil can radiate out the possibility that many isotopes. In order to protect the human body from radon, above all, the development of accurate measurement techniques to formulate appropriate measures should be followed. This study Gamma-ray spectrometry using a high purity germanium (HPGe) detector, if you want to measure radon unstable the nature radiation of the background problems can be reduced, radium and radon daughter nuclides after radioactive equilibrium leads to Radon concentration was measured, the soil samples from the Gamma-ray emitting nuclides, and the energy spectrum is analyzed.

• Journal of Radiation Protection and Research
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• v.34 no.2
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• pp.69-75
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• 2009

To obtain precise information about characteristics of gamma ray detector system responses, we developed new GUI computer program to analize full-energy absorption peak using our developed Delphi computer code for educational purpose. By use of the well known 4 nonlinear peak shaping functions, peaks were fitted with least square fit method in this code. In this paper, we described the methods to search for 12 coefficients in above 4 nonlinear peak shaping functions by use of our developed code in details. The computer code was tested for 1 $\mu$Ci $^{137)Cs$ 661 keV gamma ray peak spectrum detected by 25 % relative efficiency HPGe detector with 5.35 cm (D) $\times$ 5.5 cm (L) size.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1036-1042
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• 2020

Fast-Neutron Multiplicity Counter based on Liquid Scintillator Detector can directly measure the fast neutron multiplicity emitted by UO₂ rod. HPGe gamma spectrometer; which has superior energy resolution; is routinely used for the gamma energy spectrum measurement. Combing Fast-Neutron Multiplicity Counter and HPGe γ-spectrometer, the n/γ joint measurement system is developed. The fast neutron multiplicity and gamma energy spectrum of UO₂ rod assemblies under different conditions are measured by the n/γ joint measurement system. The induced fission rate and the ²³⁵U abundance, thereby the ²³⁵U mass; can be obtained for UO₂ rod assemblies. The ²³⁵U mass deviation of the measured value from the reference value is less than 5%. The results show that the n/γ joint measurement system is effective and applicable in the measurement of the ²³⁵U mass in samples.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.860-866
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• 2019

The lead element or its salts are good radiation shielding materials. However, their toxic effects are high. Due to less toxicity of bismuth salts, the radiation shielding properties of the bismuth salts have been investigated and compared to that of lead salts to establish them as a better alternative to radiation shielding material to the lead element or its salts. The transmission geometry was utilized to measure the mass attenuation coefficient (${\mu}/{\rho}$) of different salts containing lead and bismuth using a high-resolution HPGe detector and different energies (between 81 and 1333 keV) emitted from point sources of $^{133}Ba$, $^{57}Co$, $^{22}Na$, $^{54}Mn$, $^{137}Cs$, and $^{60}Co$. The experimental ${\mu}/{\rho}$ results are compared with the theoretical values obtained through WinXCOM program. The theoretical calculations are in good agreement with their experimental ones. The radiation protection efficiencies, mean free paths, effective atomic numbers and electron densities for the present compounds were determined. The bismuth fluoride ($BiF_3$) is found to have maximum radiation protection efficiency among the selected salts. The results showed that present salts are more effective for reducing the intensity of gamma photons at low energy region.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3824-3832
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• 2022

Prompt capture gammas are an important part of the fission reactor gamma field. Because some of the structural materials after neutron capture can emit photons with high energies forming the dominant component of the gamma spectrum in the high energy region, the following study of the high energy capture gamma was carried out. High energy gamma radiation may play a major role in areas of the radiation sciences as reactor dosimetry. The HPGe measurements and calculations of the high-energy aluminum capture gamma were performed at two moderator levels in the VR-1 pool-type reactor. The result comparison for nominal levels was within two sigma uncertainties for the major 7.724 MeV peak. A larger discrepancy of 60% was found for the 7.693 MeV peak. The spectra were also measured using a stilbene detector, and a good agreement between HPGe and stilbene was observed. This confirms the validity of stilbene measurements of gamma flux. Additionally, agreement of the wide peak measurement in 7-9.2 MeV by stilbene detector shows the possibility of using the organic scintillators as an independent power monitor. This fact is valid in these reactor types because power is proportional to the thermal neutron flux, which is also proportional to the production of capture gammas forming the wide peak.

• 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.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.43-51
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• 2004

In order to decommission the shielding concrete of KRR(Korea Research Reactor) -1＆2, it must be exactly determined activated level and range by neutron irradiation during operation. To determine the activated level and range, it must be sampled and analyzed the core sample. But, there are difficulties in sample preparation and determination of the measurement efficiency by self-absorption. In the study, the full energy efficiency of the HPGe detector was compared with the measured value using standard source and the calculated one using Monte Carlo simulation. Also. self-absorption effects due to the density and component change of the concrete were calculated using the Monte Carlo method. Its results will be used radioactivity analysis of the real concrete core sample in the future.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.526-532
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• 2019

In this study, an in-house $^{152}Eu$ calibration source was produced from a custom epoxy matrix with a density of ${\rho}=1.14g\;cm^{-3}$, which is chemically stable and durable form after its solidification. The homogeneity of $^{152}Eu$ in matrix was obtained better than 98%. For a Marinelli beaker geometry, an efficiency calibration procedure was applied to a n-type, coaxial, 78.5% relative efficient HPGe detector in the energy range of 121.7-1408.0 keV by using in-house $^{152}Eu$ calibration source. Then the measured efficiencies for Marinelli geometry were compared with the results calculated by MEFFTRAN and ANGLE softwares for the validation. Although MEFFTRAN and ANGLE have two different efficiency transfer algorithms to calculate the efficiencies, they usually need to use a reliable and accurate reference efficiency values as input data. Hence, reference efficiency values were obtained experimentally from a multinuclide standard source for the same detector-Marinelli geometry. In the present source characterization, the corrections required for self-absorption and true coincidence summing effects for $^{152}Eu$ gamma-rays were also obtained for a such close counting geometry condition. The experimental results confirmed the validity of efficiency calculations obtained by MEFFTRAN and ANGLE softwares that are calculation tools.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1754-1759
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• 2022

The activity of gamma-ray emitting nuclides is calculated assuming that each gamma-ray is detected individually; thus, the magnitude of the coincidence summing signal must be considered during activity calculations. Here, the correction factor for the coincidence summing effect was calculated, and the detection efficiencies of two HPGe detectors were compared. The CANBERRA Inc. GC4018 high-purity Ge detector provided an estimate for the peak-to-total ratio using a point source to determine the coincidence summing correction factor. The ORTEC Inc. GEM60 high-purity Ge detector uses EFFTRAN in LVis to obtain the parameters of the detector and source model and the gamma-gamma and gamma-X match estimates, in order to determine the coincidence summing correction factor. Nuclide analyses, radioactivity comparisons, and analyses of reference material samples were performed utilizing certified reference materials to accurately determine the detection efficiencies. For both Co-60 and Y-88, the detection efficiency for a point source increased by an average of at least 12-13%, whereas the detection efficiency determined using LVis increased by an average of at least 13-15%. The calculated radioactivity values of the certified reference material and reference material samples were accurate to within 3% and 6% of the measured values, respectively.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1271-1276
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• 2020

The geometrical efficiency of a source-to-detector configuration is considered to be necessary in the calculation of the full energy peak efficiency, especially for NaI(Tl) and HPGe gamma-ray spectroscopy detectors. The geometrical efficiency depends on the solid angle subtended by the radioactive sources and the detector surfaces. The present work is basically concerned to establish a new mathematical approach for calculating the solid angle and geometrical efficiency, based on conversion of the geometrical solid angle of a non-axial radioactive point source with respect to a circular surface of the detector to a new equivalent geometry. The equivalent geometry consists of an axial radioactive point source with respect to an arbitrary elliptical surface that lies between the radioactive point source and the circular surface of the detector. This expression was extended to include coaxial radioactive circular disk source. The results were compared with a number of published data to explain how significant this work is in the efficiency calibration procedure for the γ-ray detection systems, especially in case of using isotropic radiating γ-ray sources in the form of point and disk shapes.