• Journal of Radiation Protection and Research
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• 제43권4호
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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.

• Journal of Radiation Protection and Research
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• 제43권3호
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• pp.85-96
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• 2018

Background: A variety of inorganic scintillators have been developed and improved for use in radiation detection and measurement, and in situ gamma-ray spectrometry in the environment remains an important area in nuclear safety. In order to verify the feasibility of promising scintillators in an actual environment, a performance test is necessary to identify gamma-ray peaks and calculate the radioactivity from their net count rates in peaks. Materials and Methods: Among commercially available scintillators, $LaBr_3(Ce)$ scintillators have so far shown the highest energy resolution when detecting and identifying gamma-rays. However, the intrinsic background of this scintillator type affects efficient application to the environment with a relatively low count rate. An algorithm to subtract the intrinsic background was consequently developed, and the in situ calibration factor at 1 m above ground level was calculated from Monte Carlo simulation in order to determine the radioactivity from the measured net count rate. Results and Discussion: The radioactivity of six natural radionuclides in the environment was evaluated from in situ gamma-ray spectrometry using an $LaBr_3(Ce)$ detector. The results were then compared with those of a portable high purity Ge (HPGe) detector with in situ object counting system (ISOCS) software at the same sites. In addition, the radioactive cesium in the ground of Jeju Island, South Korea, was determined with the same assumption of the source distribution between measurements using two detectors. Conclusion: Good agreement between both detectors was achieved in the in situ gamma-ray spectrometry of natural as well as artificial radionuclides in the ground. This means that an $LaBr_3(Ce)$ detector can produce reliable and stable results of radioactivity in the ground from the measured energy spectrum of incident gamma-rays at 1 m above the ground.

• 분석과학
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• 제13권5호
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• pp.549-557
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• 2000

자외선/가시선 영역에서 뚜렷한 발색단이 없는 화합물 수용액의 반응을 추적하기 위해 FTIR-ATR 분광법이 사용되었다. 구체적인 예로서, 산성 수용액에서 ${\alpha}$-사이클로덱스트린과 ${\gamma}$-사이클로덱스트린의 가수분해 반응을 FRIR-ATR 분광법으로 연구하였다. 각각 1.0M, 0.5M, 0.1M 농도의 HCl 조건에서 두 사이클로덱스트린의 가수분해 반응을 연구한 결과, 강한 산성 용액에서 ${\alpha}$-사이클로덱스트린의 가수분해 생성물은 글루코오스이지만 ${\gamma}$-사이클로덱스트린의 가수분해 생성물은 글루코오스에서 더욱 분해된 물질이었다.

• 대한방사선방어학회:학술대회논문집
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• 대한방사선방어학회 2009년도 춘계 학술발표
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• pp.128-129
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• 2009

• Nuclear Engineering and Technology
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• 제55권1호
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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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• 제2권4호
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• pp.273-278
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• 1970

국내에서 채광한 자연방사능광물의 분석의뢰를 계기로 이에 관심을 갖고 Uranium와Thorium 의 함유량을 원광으로부터 비파괴적이고 간단한 ${\gamma}$-선 분광법으로 측정 분석하여 보았다. Ge(Li) 측정기를 ${\gamma}$-선 분광에 이용한 결과는 재래식방법에 비하여 분석정도를 훨씬 높일 수 있었으며, 자연방사능측정법과 원자로중성자로 조사시킨 activation법으로 Uranium와 Thorium의 함유량을 분석하였던바 하나의 시료(인천에서 채광)에서는 약 0.5%의 Uranium, 또 다른 시료에서는 2%의 Th와 0.1%의 U이 포함되어 있음이 확인되었다. 우리가 처음 시도한 이 분석법은 경제적이고 간편.신속하게 분석할수 있어 핵연료물질조사에 널러 이용할 수 있음을 입증한다.

• Journal of Radiation Protection and Research
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• 제38권2호
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• pp.119-123
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• 2013

HPGe 검출기를 사용하는 감마 분광분석계의 건전성을 점검하기 위한 실험실 선원이 개발되었다. 점검 선원은 0.154 mm 이하의 라듐이 풍부한 토양을 밀봉된 원통형 시료 용기에 담은 것으로, 검출기 교정에 사용할 12 개의 감마선이 방출된다. 점검 선원의 스펙트럼은 1년 동안 1개월 간격으로 측정하였으며, 스펙트럼에 나타난 감마선 피크들의 특성을 조사하였다. 감마 분광분석계가 정상일 때 라듐과 그 붕괴 생성물에서 3% 이상 방출률을 갖는 감마선들의 피크면적과 반치폭은 77 keV 피크를 제외하고는 각각 표준편차 2%와 3% 이내에서 일정하였다. 따라서 점검 선원은 77 keV부터 2202 keV까지 영역에 있는 10개의 피크를 사용하여 분광분석계의 건전성을 점검하는데 충분한 것으로 판단되었다.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2927-2934
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• 2023

The influence of the atmospheric radon background on the airborne gamma spectrum can seriously affect researchers' judgement of ground radiation information. However, due to load and endurance, unmanned aerial vehicle (UAV)-borne gamma-ray spectrometry is difficulty installing upward-looking detectors to monitor atmospheric radon background. In this paper, an advanced spectral-ratio method was used to correct the atmospheric radon background for a UAV-borne gamma-ray spectrometry in Inner Mongolia, China. By correcting atmospheric radon background, the ratio of the average count rate of U window in the anomalous radon zone (S5) to that in other survey zone decreased from 1.91 to 1.03, and the average uranium content in S5 decreased from 4.65 mg/kg to 3.37 mg/kg. The results show that the advanced spectral-ratio method efficiently eliminated the influence of the atmospheric radon background on the UAV-borne gamma-ray spectrometry to accurately obtain ground radiation information in uranium exploration. It can also be used for uranium tailings monitoring, and environmental radiation background surveys.

• Journal of Radiation Protection and Research
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• 제46권3호
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• pp.127-133
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• 2021

Background: Phosphogypsum is material produced as a byproduct in fertilizer industry and is generally used for building materials. This material may contain enhanced radium-226 (²²⁶Ra) activity concentration compared to its natural concentration that may lead to indoor radon accumulation. Therefore, an accurate measurement method is proposed in this study to determine ²²⁶Ra activity concentration in phosphogypsum sample, considering the potential radon leakage from the sample container. Materials and Methods: The International Atomic Energy Agency (IAEA) phosphogypsum reference material was used as a sample in this study. High-purity germanium (HPGe) gamma spectrometry was used to measure the activity concentration of the ²²⁶Ra decay products, i.e., ²¹⁴Bi and ²¹⁴Pb. Marinelli beakers sealed with three different sealing methods were used as sample containers. Due to the potential leakage of radon from the Marinelli beaker (MB), correction to the activity concentration resulted in gamma spectrometry is needed. Therefore, the leaked fraction of radon escaped from the sample container was calculated and added to the gamma spectrometry measured values. Results and Discussion: Total activity concentration of ²²⁶Ra was determined by summing up the activity concentration from gamma spectrometry measurement and calculated concentration from radon leakage correction method. The results obtained from ²¹⁴Bi peak were 723.4 ± 4.0 Bq·kg^-1 in MB1 and 719.2 ± 3.5 Bq·kg^-1 in MB2 that showed about 5% discrepancy compared to the certified activity. Besides, results obtained from ²¹⁴Pb peak were 741.9 ± 3.6 Bq·kg^-1 in MB1 and 740.1 ± 3.4 Bq·kg^-1 in MB2 that showed about 2% difference compared to the certified activity measurement of ²²⁶Ra concentration activity. Conclusion: The results show that radon leakage correction was calculated with insignificant discrepancy to the certified values and provided improvement to the gamma spectrometry. Therefore, measuring ²²⁶Ra activity concentration in TENORM (technologically enhanced naturally occurring radioactive material) sample using radon leakage correction can be concluded as a convenient and accurate method that can be easily conducted with simple calculation.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.415-420
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• 2013

A determination method for $^{237}Np$ in spent nuclear fuel samples was developed using an isotope dilution method with $^{239}Np$ as a spike. In this method, inductively coupled plasma mass spectrometry (ICP-MS) was taken for the $^{237}Np$ instead of the previously used alpha spectrometry. $^{237}Np$ and $^{239}Np$ were measured by ICP-MS and gamma spectrometry, respectively. The recovery yield of $^{237}Np$ in synthetic samples was $95.9{\pm}9.7$% (1S, n=4). The $^{237}Np$ contents in the spent fuel samples were 0.15, 0.25, and $1.06{\mu}g/mgU$ and these values were compared with those from ORIGEN-2 code. A fairly good agreement between the measurements (m) and calculations (c) was obtained, giving ratios (m/c) of 0.93, 1.12 and 1.25 for the three PWR spent fuel samples with burnups of 16.7, 19.0, and 55.9 GWd/MtU, respectively.