• Proceedings of the Korean Nuclear Society Conference
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• 1996.11b
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• pp.583-589
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• 1996

원자력발전소에서 발생되는 방사성폐기물들은 폐기물형태 및 방사능 농도가 다양하며 영구처분장으로 이송전까지는 발전소내의 임시 저장고에 안전하게 보관, 관리하고 있다. 생성된 폐기물드럼내에는 감마방출핵종을 비롯하여 알파 및 베타방출 핵종들이 균질 또는 비균질하게 존재하고 있으며 방사능의 세기나 폐기물의 특성에 따라 안정화시키거나 압축처리하여 드럼에 담겨져 있기 때문에 일반적인 파괴분석에 의한 화학분석법으로는 작업자의 피폭, 시료의 대표성 선정 및 장시간의 화학처리 시간소요 등으로 핵종분석이 곤란하다. 따라서 본 논문은 일반적으로 감마핵종분석시 흔히 사용하고 있는 고순도게르마늄(HPGe) 검출기를 이용하여 드럼의 감마핵종농도를 분석하는 방법과 장치의 개발에 대해 언급하였으며 알파나 베타핵종과 같이 직접 분석이 곤란한 핵종들은 각 폐기물드럼내에 존재하는 Co-60이나 Cs-137과의 상관관계를 미리 예측한 척도인자 (scaling factor)를 이용하여 간접적으로 구하는 방법을 사용하고 있으나 본 논문에서는 드럼으로부터 감마핵종만을 분석하는 방법에 대해서만 언급하였다. 또한 핵종분석시스템의 최적 운전조건을 도출하기 위해 드럼회전테이블의 속도결정 및 모의드럼을 이용한 방사능측정 등을 통해 핵종 농도 분석시의 오차를 30% 이내로 유지할 수 있었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.545-550
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• 1998

PWR 사용후핵연료 내에 존재하는 $^{134}$ Cs/$^{137}$Cs 및 $^{154}$ Eu/$^{137}$Cs의 감마선 핵종비를 써서 각각 연소도를 결정하고, 그들의 차이가 최소가 되는 시간을 찾는 방법으로 사용후핵연료의 냉각시간을 결정하였다. $^{134}$ Cs/$^{137}$ Cs 및 $^{154}$ Eu/$^{137}$Cs의 핵종비로부터 연소도를 구하는 방법은 이들 핵종비에 대한 ORIGEN-5 코드 계산과 감마스캐닝 실험 결과를 비교하는 것이었다$^{[1]}$ . 사용후핵연료의 냉각시간을 임의의 시간으로 가정하고 핵종비 $^{134}$ Cs/$^{137}$ Cs을 써서 구한 연소도와 $^{154}$ Eu/$^{137}$Cs를 써서 구한 연소도의 차이를 계산했으며, 이 차이는 실제 측정대상 핵연료의 냉각시간에서 최소가 될 것을 기대하였다. 감마선 방출 핵분열생성물인 $^{134}$ Cs와 $^{154}$ Eu는 비교적 긴 반감기를 갖고 있으면서도 또 이들의 반감기 차이가 약 6.4년이나 되므로 기존의 방법$^{[2]}$ 에 비해 넓은 범위의 냉각시간을 정확하게 측정할 수 있었다.

• Journal of Radiation Protection and Research
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• v.34 no.4
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• pp.190-194
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• 2009

Gamma spectral results for macroalgae samples taken from the environment of Ulchin nuclear power plants in Korea (east coast), showed 766 keV peaks, which were identified as $^{95}Nb$ by several research institutes. After the enhancement of liquid radioactive waste disposal facility at Ulchin NPP site, the $^{95}Nb$ amount in the liquid radioactive waste outflow has drastically reduced, but the expected reduction in $^{95}Nb$ specific activity from environmental samples did not actually show up on gamma spectroscopy. Detailed re-investigation revealed that along with 766 keV peak, other peaks (63, 92 and 1001 keV) from $^{234}Th-^{234}mPa$ decay series were also detected on spectroscopy, and that the measured half lives of the four peaks were very close to known half life of $^{234}Th-^{234}mPa$ decay series, which is 24.1 day. The measured gamma yield ratios of 766 keV peak to 1001 peak were very close to known ratio 0.35 for $^{234}mPa$. It is concluded that 766 keV peaks on gamma spectroscopy of Ulchin NPP environmental samples were mainly from $^{234}mPa$, which is one of naturally occurring radionuclides.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.321-328
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• 2019

The detection performances of the NaI(Tl), $LaBr_3$(Ce) and $CeBr_3$ scintillation detectors, which can be used to rapidly evaluate the major artificial radionuclides deposited on the soil surface in a nuclear accident or radiological emergency, were compared. Detection performance was assessed by calculating the minimum detectable activity (MDA). The detection efficiency of each detector for artificial radionuclides was semi-empirically determined using mathematical modelling and point-like sources having certified radioactivity. The background gamma-ray energy spectrum for MDA evaluation was obtained from relatively wide and flat grassland, and the MDA values of each detector for the major artificial radionuclides that could be released in nuclear accidents were calculated. As a result, the relative MDA values of each detector regarding surface deposition distribution at normal environmental radiation level were evaluated as high in the order of the NaI(Tl), $LaBr_3$(Ce), and $CeBr_3$ detectors. These results were compared based on each detector's intrinsic and measurement environment background, detection efficiency, and energy resolution for the gamma-ray energy region of the radionuclide of interest.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.52-57
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• 2004

In order to analyze key parameters of important radionuclides in dose evaluation model of decommissioning site, a sensitivity analysis was performed. This analysis assumed a resident farmer for an exposure scenario and 0.037Bq/g for the concentration of radionuclides. As a result of sensitivity analysis, the key parameters of radionuclides considered were the area of contaminated zone, external gamma shielding factor and indoor time fraction for Cs-137 and Co-60. The key parameters for C-14 were the environmental parameters and hydrological parameters of unsaturated zone. Also, the key parameter for Sr-90 was the density of contaminated zone.

• Journal of the Korean Society of Radiology
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• v.7 no.3
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• pp.213-219
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• 2013

Currently, about 35 cyclotrons have been operating in South Korea. Most of them are mainly used for the synthesis of radiopharmaceuticals such as $^{18}FDG$, which is a cancer tracer for nuclear medicine. Highly enriched $H_2{^{18}}O$ containing up to 98% of $^{18}O/O$ isotope ratio is used as the target for $^{18}F$ production. The price of the highly enriched $H_2{^{18}}O$ ranges 60~70 USD/g, and all of them have been imported from foreign country in spite of the very expensive price. The target (enriched $H_2{^{18}}O$) is non-radioactive before the proton beam irradiation. But, the post-irradiation target (used $H_2{^{18}}O$) must be managed following the National Radiation Safety Regulations, because it turns into radioactive by the radioactivation of the impurities within the target. Recently, nevertheless of the fast increasing amount of used $H_2{^{18}}O$ in accordance with the increasing number of nuclear medicine cases, any activation analysis on the used $H_2{^{18}}O$ have been conducted yet in Korea. In this research, activation analysis have been conducted to confirm the specific radioactivity(Bq/g) of each radioisotopes within the used $H_2{^{18}}O$. The analysis have been done on the 3 of 20g samples collected from the used $H_2{^{18}}O$ storages at different cyclotron centers. Based on the results, it was confirmed that the "used $H_2{^{18}}O$" contains gamma emitters such as $^{56}Co$, $^{57}Co$, $^{58}Co$, and $^{54}Mn$ as well as the considerable amount of beta emitter $^3H$. It was also confirmed that the only one sample contained over exemption level of gamma emitters while the specific activity of tritium was lower than the exemption level in all samples. The specific activity of radioisotopes were measured different levels in the samples depending on the elapsed time after irradiation. Further study on the activation of the "used $H_2{^{18}}O$" is definitely necessary, nevertheless the as-is results of this research must be useful in establishing a rational "used $H_2{^{18}}O$" management protocol.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.312-312
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• 1998

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.275-284
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• 2006

In the measurement of gamma rays emitted from the nuclide in the radioactive waste drum, to analyze the nuclide concentration accurately, it is necessary to use the proper calibration standards and to correct for the attenuation of the gamma rays. Two drums having a different density were used to analyze the nuclide concentration inside the drum in this study. After carrying out the system calibration, we measured the gamma rays emitted from the standard source inside the model drum with changing the distance between the drum and the detector. The measured values were corrected with the three kinds of gamma attenuation correction methode, as a results, the error was less than 10 % in the low density drum and less than 25 % in the high density drum. The measured activity in the short distance was more accruable than in the long distance. The transmission correction for the mass attenuation showed good results(very Low error) compared to the mean density and the differential peak correction method.

• Journal of Radiation Protection and Research
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• v.9 no.1
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• pp.11-18
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• 1984

A Standard gamma-ray ionization chamber system was developed with a well type ionization chamber and micro current measuring circuit. Micro current was measured by the automatic Townsend balance with stepwise compensation method. For gamma emitting nuclides such as $^{241}Am,\;^{133}Ba,\;^{60}Co,\;^{134}Cs,\;^{137}Cs,\;and\;^{22}Na$ relative calibration factors to $^{226}Ra$ reference source were calculated and detection .efficiency curve was determined as a fudnction of gamma energy.

• 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.