Currently, Japan is undertaking a nationwide project to measure and map radioactive contamination around Fukushima, as part of the efforts to restore normalcy following the nuclear accident. The Japan Atomic Energy Agency (JAEA) manages the Fukushima Environmental Safety Center, located approximately 20 km north of the Fukushima Daiichi nuclear power plant in Minamisōma City, Fukushima Prefecture. In collaboration with the JAEA, this study involved conducting comparison experiments and analyses with radiation detectors in high radiation environments, a challenging task in Korean environments. Environmental radiation surveys were conducted using three types of detectors: CZT, NaI(Tl), and LaBr3(Ce), across two contaminated areas. Dose rate values were converted using dose rate conversion factors for each detector type, and dose rate maps were subsequently created and compared. The detectors yielded similar results, demonstrating their feasibility and reliability in high radiation environments. The findings of this study are expected to be a crucial reference for enhancing the verification and supplementation of procedures and methods in future radiation measurements and mobile surveys in high-radiation environments, using these three types of radiation instruments.

In this research, the dose rate was measured using a backpack-type scan survey device at 4 sites in sites around Nuclear Power Plants (Kori, Wolsong, Hanbit, Hanul), and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. Kori, Wolsong and Hanul power plants were measured within 2 km of the power plant, and Hanbit power plants were measured about 6.7 km from the power plant. As a result of measuring the dose rate with a backpacktype scan survey device, the average dose rate was the lowest in the measurement site 1 at 0.090 μSv/h, and the highest in the measurement site 4 at 0.145 μSv/h. All measurement points showed the domestic environmental dose rate level. The data obtained by the scan survey was visualized using the classed post and gridding functions of the surfer program. As a result of measurement with the HPGe detector, 137Cs was not detected, and only natural nuclides were detected. Among the detected natural nuclides, the radioactivity ratio was the highest for 40K with an average of 94.56%, and the lowest for 214Pb with an average of 0.26%. The results of this research can be used as basic data for radiation environment surveys around nuclear power plants. Further studies are needed to evaluate the radiation impacts by region and environment through periodic measurements.

n this research, the dose rate was measured using backpack-type scan survey device at 4 sites on Jeju Island, and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. As a result of measuring the dose rate with a backpack-type scan survey device, the average dose rate was the lowest in the measurement site 3 at 0.049 Sv/h, and the highest in the measurement site 1 at 0.066 Sv/h. The average dose rate of the 4 sites on Jeju Island was 0.056 Sv/h, and the dose rate on Jeju Island was lower than that of other regions. The data acquired by scan survey were interpreted using classed post and gridding function of surfer program. The radioactivity ratio of each nuclide in the gamma spectrum measured by the HPGe detector was the highest for K-40 with an average of 87.62%, and the lowest for Pb-214 with an average of 0.63%. In the case of the Jeju Island site, Cs-137 was detected, and the average radioactivity ratio of Cs-137 was 3.27%, which was the background level. The results of this research can be used as basic data on the radioactivity ratio for each nuclide and dose rate at the Jeju Island site. Further studies on the assessment of dose rates and radioactivity ratios in other regions are needed.

RADTRAN is a code that assesses the radiation risk of radioactive material transportation. RADTRAN assumes that the package is a point source or a line source regardless of package type and corrects the external dose rate using a shape factor which depends on the critical dimension of the package. However, the external dose rate calculated using a shape factor may be different from the actual external dose rate. Therefore, it is necessary to analyze the effect of the shape factor on the external dose rate. In this study, the effect of the shape factor on the external dose rate in RADTRAN was analyzed by comparison with MCNP. This study analyzed change in external dose rate depending on the distance from the package and the critical dimension. The distance from the package was in the range of 1–800 m. The shape of the package was assumed to be cylindrical with a radius of 1 m, and the critical dimensions of the package were assumed to be 2, 4, and 8 m. Attenuation and build-up in the air were not considered to consider only the effect on the shape factor. When simulating the exposure situation using MCNP, the package was assumed to be a volume source, and flux by distance from the package was calculated using F5 tally. The dose rate at 1 m from the package was normalized to 2 mSv·hr−1. As a result of the analysis, the external dose rates of the package were higher in RADTRAN than in MCNP. For the critical dimension of 2, 4, and 8 m, when the distance from package is 1–10 m, the RADTRAN was 1.83, 4.08, and 5.27 times higher on average than MCNP, respectively. And when the distance from the package was 10–100 m and 100–800 m, RADTRAN was 1.10, 2.02, 3.01 times and 1.04, 1.92, 2.43 times higher than MCNP, respectively. It was found that the larger the distance from the package is and the smaller the critical dimension of the package is, the less conservatively RADTRAN assessed. It is because the shape of the package gets closer to the point source as the distance from the package increases, and the shape factor decreases as the critical dimension of the package decreases. The result of this study can be used as the basis for radiation risk assessment when transporting radioactive materials.

This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g−1 60Co and 2.63×105 Bq·g−1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr−1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr−1) as per global standards.

Technegas를 사용한 검사는 단순 확산 누적을 통해 폐 영상을 이미지화하기 때문에 검사를 마친 후에 검 사실이 오염될 수 있다. 따라서 방사선 작업 종사자와 검사를 기다리는 환자는 technegas 흡입으로 인한 내 부 피폭의 영향을 받게 된다. 이에 중력환기 전후의 시간경과에 따른 공간선량율 분포를 비교, 분석함에 따 라 방사선사, 의료진, 대기 환자의 피폭선량 저감화 방법을 모색하고자 한다. 중력환기 전후 환자의 호흡기 위치에서 거리별, 각도별로 공간선량율을 10분 동안 측정하고 평균값, 표준 편차 및 감소율을 계산하였다. 실험 결과, 중력 환기 전후 감소율은 최고 95.31%였고 가장 높은 감소율은 1 ∼ 3분 사이에서 나타났다. 중 력환기를 통해서 방사선 작업종사자, 대기환자, 환자 보호자 및 간호사의 피폭선량을 감소시킬 수 있다. 결 론적으로 중력환기를 통한 피폭선량 감소 결과는 방호 최적화를 이루는 역할을 할 것이며 ICRP 103에서 권고한 의료 피폭 저감화에 부합된다.

본 연구는 선형가속기를 이용하여 10 MV 광자선을 조사하는 과정에서 발생하는 광중성자의 선량률 변 화를 측정하고자 하였다. 또한 방사선 조사가 종료된 후 광중성자의 수명을 분석하고자 하였다. 광중성자 측정은 BF3 비례계수관을 사용하였으며, 광중성자의 선량률 측정결과를 2초 간격으로 3부분으로 나누어 분석 하였다. 측정결과 조사야 내에 금속판이 없는 경우와 납판이 존재할 때 광중성자의 발생이 가장 빠르 게 나타났으며, 최종적으로 백그라운드 수준의 선량률을 나타내는 시간은 물질의 종류와 무관하게 약 1분 40초 정도의 수명시간을 나타내었다. 따라서 광중성자가 수명을 다할 때까지의 시간에 따른 선량률은 물질 의 종류와 임계에너지에 따라 다르게 나타내었다. 그러나 최종 수명시간은 물질의 종류에 관계없이 비슷한 결과를 나타내었으므로 물질의 종류가 광중성자의 수명시간에는 크게 관여하지 않는다고 판단되어진다.

이 연구는 금속쐐기필터를 대신하여 동적쐐기필터를 사용할 수 있을지에 대한 적정성 평가에 관한 것이다. 선형가속기에서 발생되는 엑스선 에너지는 6 MV, 10 MV로 상용화되어있다. 금속쐐기필터의 15°, 30°, 45°, 60°에 각각 100, 200, 300, 400, 500, 600 선량율(MU/min)로 48번 조사하였고, 동적쐐기필터와의 비교를 위해 같은 조건으로 총 96번 조사하였다. 측정조건은 선원필름간 거리 100cm, 조사면 10×10cm 로 측정하였다. 현상된 필름을 스캔하여 선량분석프로그램으로 교정 후 분석하였고, 표준편차를 구해 선량율을 비교하였다. 동적쐐기필터는 선량, 산란선 및 치료시간을 감소시키며, 환자에 조사되는 선량이 적어 매우 유용하다. 동일한 조건에서 선량율에 따른 오차는 연관성이 없으므로 환자의 상황에 따라 고선량율의 치료를 사용하는 것도 가능할 것으로 생각된다.

고용량옥소치료시수평면, 수직면, 방위각에따라3차원적으로방출되는공간선량률을측정하였다. 정확한측정을위해기하학적구조의알루미늄틀을제작하여100 cm 거리에서, 높이(5 측정점), 방위각(8 측정점), 시간적간격(6 측정점)을각각나누어서분석하였다. 수직면상공간선량률분포는131I을경구투여24 시간후환자로부터거리100 cm , 높이100 cm 지점에서환자군평균71.85 µSv/h로가장높았다. 수분섭취를통한공간선량률감쇠정도를두그룹으로나누어실험하였다. 131I을경구투여24시간뒤거리100 cm, 높이100 cm에서공간선량률분포가A 실험군은44.9 µSv/h이고B 실험군은100.28 µSv/h이다. A 실험군이B 실험군과비교하여고용량옥소치료시수분섭취정도에따라약53 %의피폭경감효과를확인하였다.

간담도암의 일차적 치료법은 수술적 제거이나, 초기 병기로 수술을 하여도 재발이 흔하며, 재발 후에는 근치적 목적이 아닌 주로 증상 완화 목적의 치료를 하고 있다. 이에 본 연구에서는 재발한 환자에서 자체 도입한 특수 관을 이용한근치적 목적의 high-dose-rate intraluminal brachytherapy의 치료법과 치료 set-up 재현성 유지에 대하여 고찰하고자 한다. 본 연구는 간담도암으로 수술 후, 6개월 간격으로 복부 자기 공명 영상을 이용한 추적 검사를 시행하여, 수술문합 부위의 재발이 확인된 임상 병기 rcT1N0M0인 환자를 대상으로 하였다. 근접 치료를 위하여 경피적 담도 배액관을 치료 전용관인 Arrow Sheath로 교체 후, intraluminal catheter를 삽입하여, 일회 치료당 3 Gy 씩 총 30 Gy 근접치료를 시행하였다. 협착 된 담도의 distal and proximal 그리고 central axis point로부터 1 cm 떨어진 지점을 선량표준화였으며, 5회에 걸쳐 치료 전 dummy seed 삽입 후, 치료 표적 용적 근위부와 말단부의 기준점에서 흉추 사이의 거리를 측정하여 set-up 재현성을 평가하였다. 매주 치료 전 근위부와 말단부위 치료 표적 용적 기준점에서 흉추 사이의거리를 5회 측정한 값과 simulation image에서 측정된 거리를 비교한 결과 평균값과 표준편차가 오차율 5% 이내로 유의한 수준으로 측정되어 매회 치료시 재현성이 유지 되었다, 본 연구 환자의 추적 검사 상 1년 이상 무병 생존하고 있으며, 심각한 부작용 또한 관찰되지 않았다. 따라서, 수술이 불가능한 재발한 간담도암 환자에서 본원에서 자체 도입한관으로 시행한 근접치료는 주변 인접 장기의 심각한 부작용 발생 없이, 매 회 치료 시 재현성이 유의한 수준으로 유지되는 것으로 평가되었다.