• Journal of Radiation Protection and Research
• /
• 제39권2호
• /
• pp.96-102
• /
• 2014

소변시료 바이오어세이는 분석설비가 비교적 간단하고 시료 채취가 용이하여 내부피폭선량 평가를 위하여 널리 사용되고 있는 대표적인 간접측정법이다. 본 연구에서는 소변시료 바이오어세이 결과에 대해 보다 객관적인 성능검사를 수행하기 위하여 미국 NIST에서 주관한 NRIP (NIST raiochemistry inercomparison pogram)에 참여한 결과를 소개하였다. 60일간의 분석기간 동안 인공합성소변으로 제작된 검사시료의 방사능분석결과를 보고하는 cstomary exercise에서는 12가지 방사성핵종에 대한 측정 결과 ANSI N13.30에서 제시하는 성능검사 기준을 모두 만족하는 것으로 확인되었다. 비상상황에 대비하여 8시간 이내에 방사능분석결과를 신속하게 보고하는 eergency preparedness exercise에서는 9가지 방사성핵종에 대하여 -35 ~ 45%의 차이를 나타내어 확인된 오차범위 내에서 비상시 신속한 내부피폭 분류에 적용하기에 적합한 것으로 확인되었다.

• Restorative Dentistry and Endodontics
• /
• 제13권1호
• /
• pp.69-78
• /
• 1988

The study was designed to establish a more nearly quantitative method for assessing the marginal leakage of dental restorations. 27 Class V cavities with $45^{\circ}$ bevel joint were prepared and classified into 2 groups. One group was filled with Scotchbond and silux. The other group was filled with glass ionomer cement, Scotchbond and silux. After finishing, all specimens were subjected manually to 100 thermal cycles at $0^{\circ}C$ and $100^{\circ}C$ water-bath. They were soaked in a samarium nitrate solution for 3 hours, irradiated with flux of $6{\times}10^{12}$ neutrons/$cm^2$/sec for 11 hours, cooled for 200 hours, counted with the HPGE detector and the tracer uptake was determined by comparison with a standard of samarium (10 ${\mu}g$). The following results were obtained. 1. Both of the two groups showed a considerable amounts of marginal leakage. 2. The group filled without glass ionomer cement base showed more marginal leakage than the group filled with glass ionomer cement base. 3. Neutron Activation Analysis produced a good quantitative method to measure the marginal leakage and samarium was appropriate to measure the marginal leakage of resin restorations using neutron activation analysis.

• Journal of Radiation Protection and Research
• /
• 제24권3호
• /
• pp.171-184
• /
• 1999

본 논문에서는 저준위 환경방사능 측정시 이용되는 검출한계에 대한 기본개념 및 수식을 통계학적 이론을 기초로 서술하였다. 방사능을 정확하게 검출할 신뢰도를 95%로 설정하여 알파 베타 및 감마선 측정기에 대한 검출한계치값을 계산하였다. 또한 환경 방사능 계측에 많이 사용되는 방사능 핵종에 대해 최소 검출 방사능 농도값을 검출한계치 개념을 근거로 계산하여 저준위 환경 방사능 분석시 환경방사능 측정결과에 대한 신뢰도 평가에 활용 가능 하도록 하였다.

• 방사성폐기물학회지
• /
• 제14권1호
• /
• pp.57-61
• /
• 2016

The Tomographic Gamma Scan (TGS) technique partitions radioactive waste drums into $10{\times}10{\times}16$ voxels and assays both the density and concentration of radioactivity for each voxel thus providing for improved accuracy, when compared to the traditional Non-Destructive Assay(NDA) techniques. It could decrease the degree of precision measurement since there is a trade-off between spatial resolution and precision. This latter drawback is compensated by expanding the Region of Interest (ROI) that differentiates the full energy peaks, which, in turn, results in an optimized degree of precision. The enlarged ROI, however, increases the probability of interference among those nuclides that emit energies in the adjacent spectrum. This study has identified the cause of such interference for the reference nuclide of the TGS technique, $^{137}Cs$ (661.66 keV, half-life 30.5 years), to be $^{110m}Ag$ (657.75 keV, half-life 249.76 days). A new calibration method of determining the optimized ROI was developed, and its effectiveness in accurately characterizing $^{137}Cs$ and eliminating the interference was further ascertained.

• Nuclear Engineering and Technology
• /
• 제40권1호
• /
• pp.93-98
• /
• 2008

Several adhesive tapes have been studied in terms of their suitability for securing gold wires into positions for neutron flux measurements in the reactor core and irradiation facilities surrounding the core of the Open Pool Australian Light water (OPAL) reactor. Gamma ray spectrometry has been performed on each irradiated tape in order to identify and quantify activated components. Numerous metallic impurities have been identified in all tapes. Calculations relating to both the effective neutron shielding properties of the tapes and the error in measurement of the $^{198}Au$ activity caused by superfluous activity due to residual tape have been made. The most important identified effects were the prolonged cooling times required before safe enough levels of radioactivity to allow handling were reached, and extra activity caused by residual tape when measured with an ionisation chamber. Knowledge of the most suitable tape can allow a minimal contribution due to these effects, and the use of gamma spectrometry in preference to ionisation chamber measurements of the flux wires is shown to make all systematic errors due to the tape completely negligible.

• 한국표면공학회지
• /
• 제56권1호
• /
• pp.84-93
• /
• 2023

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

• Nuclear Engineering and Technology
• /
• 제55권6호
• /
• pp.1959-1965
• /
• 2023

The objective of this study was the establishment of clearance method that can ensure radiological safety and reasonably minimize radioactive waste when demolishing contaminated buildings at KRR-1&2. By reviewing Korean and international laws related to decommissioning, the method for clearance of contaminated buildings presented in this study is to first decontaminate the building and then conduct a radiological safety assessment, such as measuring residual radioactivity, to determine whether the radiation dose criteria for clearance are satisfied. The measurement results meet the radiation dose criteria, the contaminated buildings are regarded as clearance and can be converted into the general buildings. The demolition of the cleared buildings is carried out using conventional demolition methods. The waste generated during the demolition is classified as general construction waste and is disposed of according to relevant laws. The proposed method significantly optimized the number of samples analyzed and reduced the time and cost associated with the decommissioning. The established method will be applied to the ongoing decommissioning of contaminated buildings at KRR-1&2, and its application will be verified by regulatory bodies. The study suggests that this method could be used for the decommissioning of contaminated buildings at other Korean nuclear facilities in the future.

• Clinical and Experimental Pediatrics
• /
• 제46권9호
• /
• pp.921-925
• /
• 2003

저자들은 얼굴 부종, 복부 팽만 및 저알부민혈증을 주소로 내원한 8세 환아에서 원발성 장림프관 확장증으로 진단된 1례를 경험하였기에 문헌 고찰과 함께 보고하는 바이다.

• 방사선산업학회지
• /
• 제18권1호
• /
• pp.89-93
• /
• 2024

Accidents at nuclear facilities and nuclear power plants led to leaks of large amounts of radioactive substances. Of the various radioactive nuclides released, ¹³⁷Cs are radioactive substances generated during the fission of uranium. Therefore, due to the high fission yield (6.09%), strong gamma rays, and a relatively long half-life (30 years), a rapid and efficient removal method and a study of adsorbents are needed. Accordingly, an adsorbent was prepared using Prussian blue (PB), a material that selectively adsorbs radioactive cesium. As a result of evaluating the adsorption performance with the prepared adsorbent, it was confirmed that 82% of the removal efficiency was obtained, and most of the cesium was rapidly adsorbed within 10 to 15 minutes. The purpose of this study was to adsorb cesium using the Prussian blue alginate bead and to compare the change in detection efficiency according to the amount of adsorbent added for quantitative evaluation. However, in this case, it is difficult to determine the detection efficiency using a standard source with the same conditions as the measurement sample, so the efficiency change of the HPGe detector according to the different heights of Prussian blue was calculated through MCNP simulation using certified standard materials (1 L, Marinelli beaker) for radioactivity measurement. It is expected to derive a relational equation that can calculate detection efficiency through an efficiency curve according to the volume of Prussian blue, quantitatively evaluate the activity at the same time as the adsorption of radioactive nuclides in actual contaminated water and use it in the field of nuclear facility operation and dismantling in the future.

• 방사선산업학회지
• /
• 제17권1호
• /
• pp.1-9
• /
• 2023

Since a large amount of radioactive waste is expected to be generated due to permanent shutdown of many nuclear power plants, it is necessary to prepare efficient management methods for radioactive waste. Therefore, there is a need for a based study to apply the In-Situ gamma spectrometry, which can simplify the measurement procedure. The purpose of this study is to analyze research cases of In-Situ gamma spectrometry and to analyze the sensitivity of measurement according to influencing factors on In-Situ gamma spectrometry. Research cases of five institutions, including the CERN and the Imperial College Reactor Centre (ICRC), were selected as the institutions to be investigated. Research on the In-Situ gamma spectrometry was conducted on the satisfaction of the acceptance criteria for radioactive waste and the analysis of residual radioactivity in the site. In-Situ Objective Counting System (ISOCS) was used as a major measuring device. Sampling and computer code were used to verify the analysis results. For evaluation of measuring sensitivity according to influencing factors on In-Situ gamma spectrometry, the thickness of the measurement target, the distance between the detector and the target, the angle of the collimator, and the contamination location were performed using ISOCS's Geometry Composer. In every case, based on 122 keV, the efficiency decreased as the energy increased in the high energy region, and the efficiency decreased as the energy decreased in the low energy region. As the target thickness increased, the efficiency decreased, and as the distance between target and detector increased, the efficiency decreased. As the distance between contamination and detector increased, the efficiency decreased, and as the angle of the collimator increased, the measurement efficiency increased. However, when simulating the measurement situation using Geometry Composer, the background is not considered, and the probability of incident in the background increases as the angle increases, so further research needs to be conducted in consideration of these. This study can be utilized when applying the In-Situ gamma spectrometry of radioactive waste clearance in the future.