I-129 is one of the imporant nuclides that must be determined in the disposal process of radioactive waste in many countries. This radionuclide emits gamma-ray and x-ray photons within the energy range of 29 to 39 keV, consequently, an x-ray detector with high resolution performance is required for the analysis of I-129 activity. An n-type coaxial HPGe detector exhibits higher efficiency characteristics compared to a planar-type HPGe detector, however, its resolution is lower than a planar type. So it is difficult to completely deconvolute and fit the gamma-ray and xray peaks in the spectrum using a general gamma-ray spectrum analysis program such as GammaVision. To address this problem, in a previous study introduced the developed algorithm for the fitting and analysis of I-129 gamma-ray and x-ray spectum by fixing their emission ratios. In this study, we improved the algorithm by considering the variation of the efficiency in the HPGe spectrum, which reflects the actual HPGe crystal condition. And algorithm tests were performed using measured I-129 sample spectra with interfering nuclides acting as background curve are introduced.

Many radionuclides emit two or more gamma rays in a cascade once they decay. At this time, gamma rays are detected at the same time, and the signals are overlapped and measured as one added signal. This is called the summing coincidence effect, and it causes an error of more than 10% depending on the detection efficiency, measurement conditions, and target nuclide. It is known to be greater as the efficiency of the detector increases and as the distance between the source and the detector decreases. It is necessary to consider the summing coincidence effect since the efficiency of the HPGe detector owned by the KHNP CRI is as high as 65%. In this study, We would like to propose an appropriate gamma nuclide analysis method for radioactive waste generated from NPP by evaluating the influence on the summing coincidence effect.

Radioactive waste can be classified according to the concentration level for radionuclides, and the disposal method is different through the level. Gamma analysis is inevitably performed to determine the concentration of radioactive waste, and when a large amount of radioactive waste is generated, such as decommissioning nuclear facilities, it takes a lot of time to analyze samples. The performance of a lot of analysis can cause human errors and workload. In general, gamma analysis is performed using by HPGe detector. Recently, for convenience of analysis, commercial automatic sample changers applicable to the HPGe detectors were developed. The automatic sample changers generate individual analysis reports for each sample. In this study, gamma analysis procedure was improved using the application of the automatic sample changer and the automated data parsing using by Python. The application of automatic sample changers and data parsing technique can solve the problems. The human errors were reduced to 0% compared to the previous method by improving the gamma analysis procedure, and working time were also dramatically reduced. This automation of analysis procedure will contribute to reducing the burden of analysis work and reducing human errors through various improvements.

The management before disposal of spent nuclear fuel is an essential process for safe management. It is important to determine the amount of nuclide inventory in order to ensure the integrity of spent nuclear fuel, as radiation generated from the nuclides is generated along with residual heat in the spent nuclear fuel. Based on the data on the characteristics of spent nuclear fuel generated in Korea, the correlation equation between burnup and enrichment was derived by referring to overseas cases (Sweden). Source term analysis was performed using the SCALE ORIGEN ARP code by securing the burnup history of nuclear fuel. Calculation was performed by inputting the combustion history of the fuel WH14×14 and WH17×17 as a reference for CE16×16 spent fuel. Through this study, the relationship was identified using the burnup, enrichment, and cooling time factors that influence the characteristics of spent nuclear fuel. In addition, the total source and spectrum data from neutrons and gamma sources were used to find out the characteristics of fuel.

Appropriateness of the minimum detectable activity in the analysis of gamma radionuclides is very important. This is reason determine the time factor among the conditions of the analysis when it is rationally determined has the advantage that radioactivity analysis can be performed accurately and quickly. In this study, 100 mL of an unknown sample was diluted in Marinelli Beaker 1L to obtain, review data on gamma radiation analysis results and minimum detectable activity for each measurement time. The measurement was used High Purity Germanium detector, target nuclides are Co-57, Co-58, Y-88 and Cs-137. Since the radioactivity analysis sample will be expected to be the waste subject to selfdisposal or less during the radioactive waste classification, the minimum detectable activity standard was set based on the detection of less than the permissible activity for self-disposal for each nuclide. The measurement methods were measured by classifying it into seven categories: 1000 seconds, 3600 seconds, 10000 seconds, 30000 seconds, 80000 seconds, 100000 seconds, and 150000 seconds. The radioactivity from this measurement are Co-57 2.89 Bq·g−1, Co-58 0.19 Bq·g−1, Y-88 0.20 Bq·g−1, Cs-137 0.15 Bq·g−1, the measurement results under all conditions were similar. On the other hand, the minimum detectable activity showed values above the allowable activity for self-disposal in not but Co-58 at 1000 and 3600 seconds. Only after taking the measurement time of 10000 seconds, the result was derived Co-57 0.0095 Bq·g−1, Co-58 0.0068 Bq·g−1, Y-88 0.0052 Bq·g−1, Cs-137 0.0062 Bq·g−1, which was confirmed to less than the allowable activity for self-disposal by nuclide. Reasonably determining the measurement time in gamma radionuclide analysis is a very important issue in terms of economy of time and accuracy of measurement. Although this study cannot be said to be able to determine a reasonable measurement time for all gamma radionuclide analysis, it is hoped that research on various samples will be made to contribute to the efficient measurement of gamma radioactivity.

감마분광분석 시스템 상에서는 226Ra(186.2 keV)과 235U(185.7 keV)가 방출하는 감마선 에너지의 피크 중첩이 발생한다. 226Ra의 직접분석을 위해서는 중첩된 피크로부터 235U의 기여를 제거해주거나 보정상수를 이용하여 실제 226Ra의 방사능 값 으로 보정 해주어야 한다. 235U가 방출하는 다른 감마선 피크를 참조하여 235U의 기여를 제거할 경우 복잡한 수계산이 필요하며, 참조피크에서 기인하는 큰 불확도로 인해 높은 정량한계를 갖는다. 반면에 보정상수를 이용하여 226Ra을 평가할 경우 간단한 계산으로 평가가 가능하며, 간접측정시 요구되는 222Rn의 용기건전성과 방사평형 복구기간이 필요하지 않아 226Ra의 신속 측정시 유용한 방법이다. 따라서 해당 방법을 통해 원료물질 3종과 공정부산물 3종, 총 93여개 시료에 대해서 보정상수로 산출된 226Ra의 방사능 농도와 방사평형 된 214Bi의 방사능 농도의 비교를 통해 유효성을 확인하였다. 대부분 ± 20% 내에서 유효하였지만 인산석고의 경우 약 50%의 오차를 보였다. 이는 보정상수를 유도하기 위한 가정 중 238U과 226Ra의 방사평형 관계가 달라진 것으로 판단된다. 특이성을 반영한 보정상수를 적용하여 226Ra의 방사능 농도에 대한 유효성을 평가한 결 과 약 ±10%로 좀 더 정밀한 결과를 얻을 수 있었다. 본 연구에서 산출된 보정상수를 통한 226Ra의 방사능 농도 평가 방법은 복잡한 수계산이 필요하지 않고 용기선택으로부터 자유로우며 방사평형 복구를 위한 기간이 필요하지 않아 원료물질 및 공정부산물의 226Ra의 신속한 농도 분포 평가시 유효한 방법이다.

본 연구에서는 감마선으로 유도된 돌연변이체들에서 공통으로 발현되는 방사선 관련 유전자들의 발현을 연구하기 위하여, B. lentimorbus WJ5 의 방사선 유도 돌연변이체에서 발현되는 유전자를 DNA microarray로 동시에 탐색하였다. DNA microarray는 B. lentimorbus WJ5 genome을 무작위로 절단하여 2,000 단편으로 구성하였으며, 감마선 (60/Co)으로 유도된 7 돌연변이체의 발현을 정량적으로 관찰하였

본 연구는 EBT3 필름을 이용하여 감마나이프 퍼펙션 모델의 3차원적인 선량분포 측정하고 기준값과 비 교 분석하여 표준화된 측정방법의 기초로 활용하고자 한다. 2개 종합병원에 설치된 감마나이프 퍼펙션 모 델의 선량 분포를 EBT3 필름을 이용하여 정확도와 정밀도를 평가하였다. 정확도 평가를 위해 4 ㎜ 콜리메터를 사용하여 기계적인 중심축과 선량중심축의 일치도를 측정하였다. A병원 0.098 ㎜, 0.195 ㎜ 이며 B 병원 0.229 ㎜, 0.223 ㎜ 로 허용 오차 0.3 ㎜ 이하로 측정되었다. 정밀도 평가는 4, 8, 16 ㎜ 콜리메터(collimater) 각각의 x, y, z 3차원면 에서의 반치폭(FWHM : Full Width at Half Maximum)을 이미지-제이 프로그램을 이용하여 평가하였다. A 병원은 –0.283∼0.583 ㎜, B 병원은 –0.857∼ 0.810 ㎜로 50%선 ± 1 ㎜ 이하의 기준에 적합하였다. 이미지−제이 프로그램을 이용한 선량 분포 분석의 경우 측정자 간의 오차가 발생 가능함으로 측정점에 대한 명확한 기준을 확립할 필요가 있으며, 감마나이프 방사선 수술이 시행되어지는 고선량 영역에서 사용 가능한 선량영역이 높은 필름을 이용한 치료계획과 실제 치료 조사면의 비교가 필요하다고 생각된다.

감마나이프 방사선수술 전방향 치료계획과 역방향 치료계획을 비교 분석하였다. 10 case의 청신경초종 영상을 이용 하여 동일한 조건으로 전방향 치료계획 1, 2(FP-1,2) 및 역방향 치료계획(IP)을 수립하고, 샷의 수(No of shot), conformity index(CI), Paddic conformity index(PCI), Gradiant index(GI), 치료시간 등을 비교 하였다. IP가 FP에 비 하여 샷의 수가 적었으며, 표적용적이 증가할수록 샷의 수는 증가하였다. CI는 FP-1:0.85, FP-2 :0.86, IP:0.94, PCI 는 FP-1:0.79, FP-2 :0.81, IP:0.78로 IP가 높거나 비슷한 결과를 보였다. GI는 FP-1:2.94, FP-2:2.94, IP:3.01로 비슷한 값을 나타내었다. FP를 기준으로 상대적 조사시간은 전체적으로 IP가 짧은 것으로 나타났다. IP는 FP와 비슷 하거나 우수한 평가값을 나타내고 치료계획에 소요되는 시간이 짧고 치료시간이 짧아 임상적으로 유용한 것으로 판단 된다.