• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.190-200
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• 2023

As the field of application of in-situ gamma spectroscopy is diversified, proficiency is required for consistent and accurate analysis. In this study, a program was developed to virtually create gamma energy spectra of artificial nuclides, which are difficult to obtain through actual measurements, for training. The virtual spectrum was created by synthesizing the spectra of the background radiation obtained through actual measurement and the theoretical spectra of the artificial radionuclides obtained by a Monte Carlo simulation. Since the theoretical spectrum can only be obtained for a given geometrical structure, representative major geometries for in-situ measurement (ground surface, concrete wall, radioactive waste drum) and the detectors (HPGe, NaI(Tl), LaBr₃(Ce)) were predetermined. Generated virtual spectra were verified in terms of validity and harmonization by gamma spectrometry and energy calibration. As a result, it was confirmed that the energy calibration results including the peaks of the measured spectrum and the peaks of the theoretical spectrum showed differences of less than 1 keV from the actual energies, and that the calculated radioactivity showed a difference within 20% from the actual inputted radioactivity. The verified data were assembled into a database and a program that can generate a virtual spectrum of desired condition was developed.

• Nuclear Engineering and Technology
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• v.5 no.4
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• pp.291-309
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• 1973

Exposure rates due to leakage gamma-rays from operating reactors TRIGA Mark II and III were measured in a horizontal plane by means of scintillation spectrometry using a 3"$\times$3" cylindrical Nal(T1) detector associated with a 400 channel pulse height analyzer under varied conditions of reactor operation. In determining exposure rate due to the leakage gamma-rays at each point of measurement, Moriuchi's spectrum-exposure rate conversion theory was applied instead of using conventional responce matrix method which necessitates very complicated procedures to convert a spectrum into exposure rate. The results show that a basic pattern of "typical" spectrum of the reactor leakage gamma-rays is neither affected by thermal output of the reactor, nor influenced by overall attenuation in radiation intensity. It was indicated that he attenuation of the leakage gamma-rays in air in terms of exposure rate as a whole follows an exponential law, and the total exposure rate due to the leakage gamma-rays at a certain point is nearly proportional to thermal output of the reactor. The complexity in spectrum measured for a movable core reactor, TRIGA Mark III, was analyzed through spectrum resolution, and proper judgement of the leakage gamma-rays in a complex spectrum was discussed.ctrum was discussed.

• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.457-464
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• 2000

In this paper, we describe the results of various calculations performed for a design of the thickness gauges that use the gamma-ray backscattering method. The radiation source is assumed to be the $_{24}$1Am(60keV gamma-ray) and the detector is a single crystal scintillator in a cylindrical form. The source is located at the center of the detector with the collimator of a cylindrical shape. First, when gamma-rays are incident on a material with a constant angle, we compute the variations of the spectrum for the photons scattered into different angular intervals. Next, we compute for an optimal size for the collimator cylinder for a fixed detector size and an optimal distance from the detector to the material. Finally, we compute the number of observed photons for different thickness of two different materials, a plastic film and an Al foil.

• Journal of Sensor Science and Technology
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• v.22 no.3
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• pp.223-226
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• 2013

We designed a single polyethylene bonner sphere with several thermo-luminescence dosimeters (TLD), for measurement of neutron energy spectrum. For the separation of the neutron dosage in the neutron-gamma mixed field, we used 21 ea TLD-600s and TLD-700s, respectively. Because, TLD-600 is sensitive to neutron and gamma rays, and, TLD-700 is sensitive only to gamma-rays, we could determine the each dose by neutron and gamma rays. The neutron response function of the bonner sphere with TLDs was calculated by MCNPX (ver. 2.5.0) Monte Carlo simulation in the energy range from $10^{-1}$ to 20 MeV. For the Cf-252 standard neutron source in KRISS, we could estimate the neutron energy spectrum by unfolding method using the response function.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.567-572
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• 2011

A sensitivity analysis of the methods used to evaluate the transport properties of a CdZnTe detector was performed using two different radiations (${\alpha}$ particle and gamma-ray) emitted from an $^{241}Am$ source. The mobility-lifetime products of the electron-hole pair in a planar CZT detector ($5{\times}5{\times}2\;mm^3$) were determined by fitting the peak position as a function of biased voltage data to the Hecht equation. To verify the accuracy of these products derived from ${\alpha}$ particles and low-energy gamma-rays, an energy spectrum considering the transport property of the CZT detector was simulated through a combination of the deposited energy and the charge collection efficiency at a specific position. It was found that the shaping time of the amplifier module significantly affects the determination of the (${\mu}{\tau}$) products; the ${\alpha}$ particle method was stabilized with an increase in the shaping time and was less sensitive to this change compared to when the gamma-ray method was used. In the case of the simulated energy spectrum with transport properties evaluated by the ${\alpha}$ particle method, the peak position and tail were slightly different from the measured result, whereas the energy spectrum derived from the low-energy gamma-ray was in good agreement with the experimental results. From these results, it was confirmed that low-energy gamma-rays are more useful when seeking to obtain the transport properties of carriers than ${\alpha}$ particles because the methods that use gamma-rays are less influenced by the surface condition of the CZT detector. Furthermore, the analysis system employed in this study, which was configured by a combination of Monte Carlo simulation and the Hecht model, is expected to be highly applicable to the study of the characteristics of CZT detectors.

• Analytical Science and Technology
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• v.35 no.2
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• pp.69-81
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• 2022

The method of measuring and classifying the energy category of neutrons directly using raw data acquired through a CZT detector is not satisfactory, in terms of accuracy and efficiency, because of its poor energy resolution and low measurement efficiency. Moreover, this method of measuring and analyzing the characteristics of low-energy or low-activity gamma-ray sources might be not accurate and efficient in the case of neutrons because of various factors, such as the noise of the CZT detector itself and the influence of environmental radiation. We have therefore developed an efficient method of analyzing radiation characteristics using a neutron and gamma-ray analysis algorithm for the rapid and clear identification of the type, energy, and radioactivity of gamma-ray sources as well as the detection and classification of the energy category (fast or thermal neutrons) of neutron sources, employing raw data acquired through a CZT detector. The neutron analysis algorithm is based on the fact that in the energy-spectrum channel of 558.6 keV emitted in the nuclear reaction ¹¹³Cd + ¹n → ¹¹⁴Cd + in the CZT detector, there is a notable difference in detection information between a CZT detector without a PE modulator and a CZT detector with a PE modulator, but there is no significant difference between the two detectors in other energy-spectrum channels. In addition, the gamma-ray analysis algorithm uses the difference in the detection information of the CZT detector between the unique characteristic energy-spectrum channel of a gamma-ray source and other channels. This efficient method of analyzing radiation characteristics is expected to be useful for the rapid radiation detection and accurate information collection on radiation sources, which are required to minimize radiation damage and manage accidents in national disaster situations, such as large-scale radioactivity leak accidents at nuclear power plants or nuclear material handling facilities.

• Journal of Radiation Protection and Research
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• v.35 no.2
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• pp.77-80
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• 2010

The spectroscopy toolkit has been developed and tested. The toolkit consists of a CsI(Tl)/PIN diode detector, integrated electronics, and a multi.channel.analyzer and its size was 40 cm(width) by 20 cm(length) by 6 cm(high). It is compact, very portable and simpler and cheaper compared to the conventional spectroscopy system. The gamma energy resolutions of the toolkit were 7.9% for the 660 keV of $^{137}Cs$ and 4.9% for 1,332 keV of $^{60}Co$ respectively. The linearity for gamma energies was good. When the energy spectrum of a ceramic sample containing $^{232}Th$ was measured with the spectroscopy toolkit for 20 minutes, there were significant peaks of the heavy metal. These results show that the resolution of the spectroscopy toolkit is sufficient to accumulate a quality spectrum in a few minutes by using weak, encapsulated commercial sources. Furthermore a toolkit experiment that how to measure energy spectra using the toolkit, and how to identify specific isotopes in a pottery piece, could be widely adopted for education and even for more sophisticated and higher level experiments.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1771-1776
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• 2020

Adverse effects in the measured gamma spectrum caused by radioactive statistical fluctuations, gamma ray scattering, and electronic noise can be reduced by energy spectrum denoising. Wavelet threshold denoising can be used to perform multi-scale and multi-resolution analysis on noisy signals with small root mean square errors and high signal-to-noise ratios. However, in traditional wavelet threshold denoising methods, there are signal oscillations in hard threshold denoising and constant deviations in soft threshold denoising. An improved wavelet threshold calculation method and threshold processing function are proposed in this paper. The improved threshold calculation method takes into account the influence of the number of wavelet decomposition layers and reduces the deviation caused by the inaccuracy of the threshold. The improved threshold processing function can be continuously guided, which solves the discontinuity of the traditional hard threshold function, avoids the constant deviation caused by the traditional soft threshold method. The examples show that the proposed method can accurately denoise and preserves the characteristic signals well in the gamma energy spectrum.

• Journal of Radiation Protection and Research
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• v.43 no.4
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• pp.137-142
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• 2018

Background: Gamma-ray detectors having a thin window of a material with low atomic number can increase the true coincidence summing effects for radionuclides emitting X-rays or gamma-rays. This effect can make efficiency calibration or spectrum analysis more complicated. In this study, a Cu shield was tested as an X-ray filter to neglect the true coincidence summing effect by X-rays and gamma-rays in gamma-ray spectrometry, in order to simplify gamma-ray energy spectrum analysis. Materials and Methods: A Cu shield was designed and applied to an n-type high-purity germanium detector having an $X-{\gamma}$ summing effect during efficiency calibration. This was tested using a commercial, certified mixed gamma-ray source. The feasibility of a Cu shield was evaluated by comparing efficiency calibration results with and without the shield. Results and Discussion: In this study, the thickness of a Cu shield needed to avoid true coincidence summing effects due to $X-{\gamma}$ was tested and determined to be 1 mm, considering the detection efficiency desired for higher energy. As a result, the accuracy of the detection efficiency calibration was improved by more than 13% by reducing $X-{\gamma}$ summing. Conclusion: The $X-{\gamma}$ summing effect should be considered, along with ${\gamma}-{\gamma}$ summing, when a detection efficiency calibration is implemented and appropriate shielding material can be useful for simplifying analysis of the gamma-ray energy spectra.

• Analytical Science and Technology
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• v.16 no.1
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• pp.12-24
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• 2003

For the design and set-up of neutron induced prompt ${\gamma}$-ray spectroscopy system using $^{252}Cf$ neutron source, the effects of shielding and moderator materials have been examined. The $^{252}Cf$ source being used for TLD badge calibration in Korea Atomic Energy Research Institute was utilized for this preliminary experiment. The ${\gamma}$-ray background and prompt ${\gamma}$-ray spectrum of the sample containing Cl were measured using HPGe (GMX 60% relative efficiency) located at the inside of the system connected to notebook PC at the outside of the system (about 20 meter distance). The background activities of neutron and ${\gamma}$-rays were measured with neutron survey meter as well as ${\gamma}$-ray survey meters, respectively and the system was designed to minimize the activities. Prompt ${\gamma}$-ray spectrum was measured using ${\gamma}$-${\gamma}$ coincident system for reduce the background and the continuum spectrum. The optimum system was designed and set up using the experimental data obtained.