• Journal of Radiation Protection and Research
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• 제9권1호
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• pp.11-18
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• 1984

Well type 전리함과 미세전류 측정희로를 사용하여 표준 감마선 전리함 장치를 개발하였다. 미세전류는 automatic Townsend balance with stepwise compensation방법을 사용하여 측정하였다. $^{226}Ra$을 기준 선원으로 택하여 감마 방출핵종인 $^{241}Am,\;^{133}Ba,\;^{60}Co,\;^{134}Cs,\;^{137}Cs^{22}Na$에 대한 비교 교정인자를 산출하였으며, 감마에너지의 함수로서 검출 효율 곡선을 구하였다.

• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3925-3932
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• 2024

During outages at nuclear power plants, much more care for radiation workers against internal exposure should be ensured given that more hot particles exist relative to the amount during normal operation. If fuel-type hot particles (FTHP) are inhaled, they can cause more severe health risks compared to activation-type hot particles (ATHP), which contain ⁶⁰Co, due to the alpha-emitting nuclides within FTHPs. The activities of difficult-to-measure nuclides within FTHPs inhaled by workers are inferred by the age-dating technique using a¹⁴¹Ce/¹⁴⁴Ce ratio as measured by whole-body counters. However, this method may be limited to outages that last for only a few months due to the short half-life (32.5 days) of ¹⁴¹Ce. We studied the feasibility of utilizing ²⁴¹Am, a nuclide with a long half-life of 432.6 years, as an alternative to ¹⁴¹Ce. Additionally, we improved the performance of a stand-type whole-body counter for low-energy gamma spectroscopy to meet the criterion (RMSE ≤0.25) specified in ANSI/HPS N13.30-2011 by employing an artificial neural network (ANN). This study can contribute to more rapid and accurate internal dose assessments for workers who have inhaled FTHPs during long-term outages at nuclear power plants.

• Nuclear Engineering and Technology
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• 제38권5호
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• pp.423-426
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• 2006

A developed neutron activation analysis(NAA) and gamma-spectrometry were applied to improve the analytical sensitivity and precision of impurities in electronic-circuit raw materials. It is well known that soft errors in high precision electronic circuits can be induced by alpha particles emitted from naturally occurring radioactive impurities such as U and Th. As electronic circuits have recently become smaller in dimension and higher in density, these alpha-particle emitting radioactive impurities must be strictly controlled. Therefore, new NAA methods have been established using a HTS(Hydraulic Transfer System) irradiation facility and a background reduction method. For eliminating or stabilizing fluctuated background caused by Rn-222 and its progeny nuclides in air, a nitrogen purging system is used. Using the developed NAA and gamma-spectrometry, ultra trace amounts of U(0.1ng/g) and Th(0.01ng/g) in an alumina ball and high purity silica used for an epoxy molding compound (EMC) could be determined.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.971-976
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• 2012

Silica-coated Au with Ag, Co, Cu, and Ir bimetallic radioisotope nanoparticles were synthesized by neutron irradiation, after coating $SiO_2$ onto the bimetallic particles by the sol-gel St$\ddot{o}$ber process. Bimetallic nanoparticles were synthesized by irradiating aqueous bimetallic ions at room temperature. Their shell and core diameters were recorded by TEM to be 100 - 112 nm and 20 - 50 nm, respectively. The bimetallic radioisotope nanoparticles' gamma spectra showed that they each contained two gamma-emitting nuclides. The nanoparticles could be used as radiotracers in petrochemical and refinery processes that involve temperatures that would decompose conventional organic radioactive labels.

• 한국방사선학회논문지
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• 제18권6호
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• pp.595-603
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• 2024

후쿠시마 원전 사고 이후 식품을 통한 방사능 섭취에 대한 관심이 매우 증가하였다. 방사능은 무색, 무미, 무취의 성질로 오염 여부를 판단하기 매우 어렵고, 섭취, 흡수 시에는 장기간 피폭으로 이어질 수 있어 식품의 방사능 안전을 확인하는 것은 중요하다. 특히 어린아이의 경우 성인에 비해 활발한 대사 활동을 하기 때문에 섭취 후 흡수에 대한 위험도가 크다고 판단되어 성인보다 엄격한 허용 기준을 적용하고 있다. 그럼에도 불구하고 방사능 오염에 대한 불안을 적지 않다. 이에 본 연구에서는 2021년 국민영양통계를 기준으로 전 연령에서 영유아기에 가장 섭취량이 많은 우유의 감마핵종을 분석하여 방사능 안전을 확인하고자 한다. 시료는 K-MILK 인증을 통해 국산 원료 100%로 제작된 국내 우유 10 종을 선정하였다. 시료 분석은 '식품의 기준 및 규격'의 방사능 시험 검사 방법에 따라 진행되었다. 분석 결과 10 종의 우유 모두 ¹³¹I, ¹³⁴Cs, ¹³⁷Cs 핵종이 MDA(최초검출가능농도) 미만의 값으로, 불검출로 판단하였다. 이에 국산 원료로 제조된 우유에서 ¹³¹I, ¹³⁴Cs, ¹³⁷Cs 핵종의 오염은 없는 것으로 판단되며 측정된 MDA 값을 이용하여 보수적으로 우유의 연간 섭취선량을 평가하였을 때 연간 유효선량한도의 0.001% 수준으로 국내 우유의 방사능 안전을 확인하였다.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.274-281
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• 2016

Background: Whole-body counters are widely used to evaluate internal contamination of the internal presence of gamma-emitting radionuclides. In internal dosimetry, it is a basic requirement that quality control procedures be applied to verify the reliability of the measured results. The implementation of intercomparison programs plays an important role in quality control, and the accuracy of the calibration and the reliability of the results should be verified through intercomparison. In this study, we evaluated the reliability of 2 whole-body counting systems using 2 calibration methods. Materials and Methods: In this study, 2 whole-body counters were calibrated using a reference male bottle manikin absorption (BOMAB) phantom and a Radiation Management Corporation (RMC-II) phantom. The reliability of the whole-body counting systems was evaluated by performing an intercomparison with International Atomic Energy Agencyto assess counting efficiency according to the type of the phantom. Results and Discussion: In the analysis of counting efficiency using the BOMAB phantom, the performance criteria of the counters were satisfied. The relative bias of activity for all radionuclides was -0.16 to 0.01 in the Fastscan and -0.01 to 0.03 in the Accuscan. However, when counting efficiency was analyzed using the RMC- II phantom, the relative bias of $^{241}Am$ activity was -0.49 in the Fastscan and 0.55 in the Accuscan, indicating that its performance criteria was not satisfactory. Conclusion: The intercomparison process demonstrated the reliability of whole-body counting systems calibrated with a BOMAB phantom. However, when the RMC-II phantom was used, the accuracy of measurements decreased for low-energy nuclides. Therefore, it appears that the RMC-II phantom should only be used for efficiency calibration for high-energy nuclides. Moreover, a novel phantom capable of matching the efficiency of the BOMAB phantom in low-energy nuclides should be developed.

• 한국방사선학회논문지
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• 제7권1호
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• pp.31-36
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• 2013

우라늄($^{238}U$)의 붕괴과정에서 생성되는 방사성기체인 라돈($^{222}Rn$)은 발생원 중 토양에서 85 % 이상으로 토양의 공극률이 클수록 토양 밖으로 방출할 수 있는 가능성이 많은 동위원소이다. 라돈으로부터 인체를 보호하기 위해서 적절한 대책을 세우는데 무엇보다도 정확한 측정기술의 개발이 선행되어야 한다. 이에 본 연구는 고순도게르마늄(HPGe) 검출기를 이용한 감마선 분광분석법으로 라돈을 측정할 경우에는 불안정한 자연방사능의 백그라운드 문제를 줄일 수 있고, 라듐과 라돈의 딸 핵종들을 방사평형에 이르게 한 후 라돈 농도를 측정하였으며, 토양시료에서의 감마선 방출핵종 및 에너지 스펙트럼을 분석하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.879-884
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• 1995

The results of performance assessment analyses have shown that the long-lived radionuclides such as I-129 control the potential individual dose impact to the public. I-129 is difficult-to-measure(DTM) in low-level waste because it is non-gamma emitting radionuclides and exists at extremely low concentrations in radioactive waste generated by nuclear reactors. In this study, computer modeling technique to predict release rate of I-129 is developed to provide another tools far performance assessment of land disposal facilities and characteristics of radwaste. Model suggested in this study will give conservative values of I-129 release rate far determination of radwaste characteristics. More detailed approach is implemented to account for release conditions of fuel source-nuclides. 1-131 concentration measured from reactor coolant and released fraction from tramp fuel have dominant roles in calculating release rate of I-129 with fuel defect conditions.

• Nuclear Engineering and Technology
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• 제54권5호
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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.

• Journal of Radiation Protection and Research
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• 제45권2호
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• pp.88-94
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• 2020

Background: The International Commission on Radiological Protection (ICRP) has recently published report series on the occupational intakes of radionuclides (OIR) for internal dosimetry of radiation workers. In this study, the optimized monitoring program including the monitoring interval and the minimum detectable activity (MDA) of major radionuclides was suggested to perform the routine individual monitoring of internal exposure based on the ICRP OIR. Materials and Methods: The derived recording levels and the critical monitoring quantities were reviewed from international standards or guidelines by the International Atomic Energy Agency (IAEA), the International Organization for Standardization (ISO), and the European Radiation Dosimetry Group (EURADOS). The OIR data viewer provided by ICRP was used to evaluate the monitoring intervals and the MDA, which are derived from the reference bioassay functions and the dose coefficients. Results and Discussion: The optimal monitoring intervals were determined taking account of two requirement conditions on the potential intake underestimation and the MDA values. The MDA requirement values of the selected radionuclides were calculated based on the committed effective dose from 0.1 mSv to 5 mSv. The optimized routine individual monitoring program was suggested including the optimal monitoring intervals and the MDA requirements. The optimal MDA values were evaluated based on the committed effective dose of 0.1 mSv. However, the MDA can be adjusted considering the practical operation of the routine individual monitoring program in the nuclear facilities. Conclusion: The monitoring intervals and the MDA as crucial factors for the routine monitoring were described to suggest the optimized routine individual monitoring program of the occupational intakes. Further study on the alpha/beta-emitting radionuclides as well as short lived gamma-emitting nuclides will be necessary in the future.