• 대한방사선방어학회:학술대회논문집
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• 대한방사선방어학회 2009년도 춘계 학술발표
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• pp.272-273
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• 2009

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.

• 한국의학물리학회지:의학물리
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• 제27권3호
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• pp.146-155
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• 2016

As the probability of exposure to radiation increases due to an increase in the use of radioisotopes and radiation generators, the importance of a radiation safety management field is being highlighted. We intend to help radiation workers with exposure management by identifying the degree of radiation exposure and contamination to determine an efficient method of radiation safety management. The personal exposure doses of the radiation workers at the Korea Institute of Radiological & Medical Sciences measured every quarter during a five-year period from Jan. 1, 2011 till Dec. 31, 2015 were analyzed using a TLD (thermoluminescence dosimeter). The spatial dose rates of radiation-controlled areas were measured using a portable radioscope, and the level of surface contamination was measured at weekly intervals using a piece of smear paper and a low background alpha/beta counter. Though the averages of the depth doses and the surface doses in 2012 increased from those in 2011 by about 14%, the averages were shown to have decreased every year after that. The exposure dose of 27 mSv in 2012 increased from that in 2011 in radiopharmaceutical laboratories and, in the case of the spatial dose rate, the rate of decrease in 2012 was shown to be similar to the annual trend of the whole institute. In the case of the surface contamination level, as the remaining radiation-controlled area with the exception of the I-131 treatment ward showed a low value less than $1.0kBq/m^2$, the annual trend of the I-131 treatment ward was shown to be similar to that of the entire institute. In conclusion, continuous attention should be paid to dose monitoring of the radiation-controlled areas where unsealed sources are handled and the workers therein.

• Journal of Radiation Protection and Research
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• 제37권4호
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• pp.231-236
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• 2012

This study derived measures to reduce exposure doses by identifying factors which affect the external radiation dose rate of patients treated with radiopharmaceuticals for PET-CT tests. The external radiation dose rates were measured on three parts of head, thorax and abdomen at a distance of 50cm from the surface of 60 PET-CT patients. It showed there are changes in factors affecting the external radiation dose rate over time after the administration of F-18 FDG. The external radiation dose rate was lower in the patients with more water intake than those with less water intake before the injection of radiopharmaceuticals at all three points: right after the injection of radiopharmaceuticals (average 4.17 mins), after the pre-PEET-CT urination step (average 77.47 mins), and right after the PET-CT test (average 114.15 mins). The study also found there is a need to increase the amount of water intake before the injection of radiopharmaceuticals in order to maintain a low external radiation dose rate in patients. This strategy is only possible under the assumption that the quality of the video has not changed after conducting this study on the relations between the image and quality. This study also found a need to use radiopharmaceuticals with the minimum amount needed for each patient because F-FDG doses affects the external radiation dose rate at the point right after the injection of radiopharmaceuticals. Urination frequency was the most significant factor to affect the external radiation dose rates at the point right after the PET-CT test and the point after the pre-PET-CT urination step. There is a need to realize the strategy to increase the urination frequency of patients to maintain the external radiation dose rate low (average 77.47 mins) before and after the injection of radiopharmaceuticals. In addition, at this point, there is a need to take advantage of personal strategies because the external radiation dose rate is lower if the fasting time is shorter, the contrast medium is used, and the amount of water intake is increased after the administration of radiopharmaceuticals. Finally this study found the need to be able to generalize these findings through an in-depth research on the factors affecting the external radiation dose rate, which includes radiopharmaceutical dose, urination frequency, the amount of water intake, fasting time and the use of contrast medium.

• 방사성폐기물학회지
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• 제20권4호
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• pp.511-521
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• 2022

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y^-1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

• 대한방사선기술학회지:방사선기술과학
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• 제43권5호
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• pp.353-357
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• 2020

IVR procedures are on the rise, and patient doses are on the rise. It is necessary to evaluate fluoroscopy dose in IVR procedure. Evaluate ESD on IVR equipment as a reference to DRL settings, I would like to present the direction of improvement in the ESD rate test criteria for fluoroscopy dose. The experimental method is measured with 6cc ionization chamber under the 20cm PMMA Phantom. Radiation is subject to abdominal procedure. The average dose rate of the incident surface was 21.6 ± 11.4 mGy/min. The highest dose equipment was 58.5 mGy/min, and there was no equipment exceeding the domestic standard of 100 mGy/min. However, there were five units above 50 mGy/min. To reduce fluoroscopy dose, it is recommended to reduce pulse rate, The dose increases as the image receptor ages. It is recommended to modify the domestic inspection criteria to 50 mGy/min.

• Radiation Oncology Journal
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• 제12권2호
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• pp.243-252
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• 1994

원격조종 아프터로딩에 의한 고선량율 관내삽입조사는 체내 발생된 종양에 방사선원을 근접시켜 치료하는 방사선요법으로서 신속한 선량계산과 선량의 정확성 및 다양한 모양의 최적선량분포가 요구된다. 저자들은 크기가 작고 선량율이 높은 고선량율의 방사성동위원소에 대한 정확한 조사선량과 최적선량분포를 얻기 위하여 수학적인 콤퓨터 계산프로그램과 실측으로서 비교하였다. 고선량율 선원에 의한 방사선 조사선량과 조직내 흡수선량분포는 각각 Sievert적분식과 Meisberger의 다항식을 이용하여 작성하였다. 종양크기와 모양에 가장 알맞는 선량분포의 최적화를 실현하기 위하여 저자들은 치료기준점의 선량을 일정한 값으로 고정시키고 선원의 조사시간을 조정하는 선형반복 계산방정식을 이용하였다. 모형선원이 장착된 아프터로딩관을 삽입하고 조준엑스선으로 촬영하여 종양부위를 결정한 후 콤퓨터의 도움으로 아프터로딩관의 축과 평행한 등량곡선 또는 과일모양의 선량분포 및 기관지 모양의 등선량분포가 성취되도록 선량최적화를 시행하였고 선량계에 의한 실측치와 오차가 $3\%$이하로 잘 일치하였다.

• 대한방사선기술학회지:방사선기술과학
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• 제39권2호
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• pp.237-246
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• 2016

방사성동위원소 사용시설 내/외 화장실 표면 방사선량률과 공간 방사선량률을 측정하여 화장실을 이용하는 환자 이외 방사선작업종사자 및 환자보호자 등의 안전성을 확보하고 방사선 방어 연구에 대한 기초 자료로 제시 하고자 한다. 2014년 5월 1일부터 7월 31일까지 인천광역시 소재 종합병원 방사성동위원소 사용시설 내/외 화장실 4곳의 공간 방사선량률과 작업 전/후 표면 방사선량률을 각각 측정하였다. 의료기관별 방사성동위원소 사용시설 내 화장실 이용 실태조사 결과 환자뿐만 아니라 환자 보호자, 일부 방사선 작업종사자까지 다양하게 이용하고 있었다. 화장실 내 공간 방사선량률 측정 결과 핵의학적 검사 중 감마촬영실을 이용하는 화장실의 누적 공간선량률은 8.86 mSv/hr으로 가장 높게 측정되었고, 방사성옥소 치료실 화장실은 7.31 mSv/hr, PET촬영실 화장실 2.29 mSv/hr, 외래 진료과 화장실 0.26 mSv/hr으로 각각 측정되었다. 방사성동위원소 작업 전/후 화장실 내 표면 방사선량률을 측정한 결과 대부분 환자 배설물이 직접 닫는 변기 앞에서 표면 방사선량률이 가장 높게 측정되었고, 화장실 내 중앙, 입구 순으로 측정되었다. 개봉선원은 물리적 반감기가 짧고 에너지가 낮아 비교적 안전하여 방사선 관리구역에서 안전하게 사용되고 있다. 그러나 저에너지 이며 짧은 반감기의 방사선원이라 하더라도 환자에게 투여되면 그 이후 환자는 움직이는 방사선원이 되며 환자가 이용하는 장소는 배설물에 의한 방사선 오염 장소가 된다. 따라서 효과적으로 유효선량을 최소화하고 불필요한 피폭선량을 줄이기 위해 방사성동위원소 투여 후 충분한 수분 섭취를 독려하여 생물학적 반감기를 낮추고, 물리적 반감기가 허용 선량이하로 될 때까지 주변인은 환자로부터 가급적 멀리 떨어져 생활하도록 권고되어야 한다.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3849-3854
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• 2022

The present work aims to optimize the radiation protection efficiency for ion-selective containers used in the liquid treatment for the nuclear power plant (NPP) cooling cycle. Some naturally occurring rocks were examined as filler materials to reduce absorbed dose and equivalent dos received from the radioactive waste container. Thus, the absorbed dose and equivalent dose were simulated at a distance of 1 m from the surface of the radioactive waste container using the Monte Carlo simulation. Both absorbed dose and equivalent dose rate are reduced by raising the filler thickness. The total absorbed dose is reduced from 7.66E-20 to 1.03E-20 Gy, and the equivalent dose is rate reduced from 183.81 to 24.63 µSv/h, raising the filler thickness between 0 and 17 cm, respectively. Also, the filler type significantly affects the equivalent dose rate, where the redorded equivalent dose rates are 24.63, 24.08, 27.63, 33.80, and 36.08 µSv/h for natural rocks basalt-1, basalt-2, basalt-sill, limestone, and rhyolite, respectively. The mentioned results show that the natural rocks, especially a thicker thickness (i.e., 17 cm thickness) of natural rocks basalt-1 and basalt-2, significantly reduce the gamma emissions from the radioactive wastes inside the modified container. Moreover, using an outer cementation concrete wall of 15 cm causes an additional decrease in the equivalent dose rate received from the container where the equivalent dose rate dropped to 6.63 µSv/h.

• Journal of Radiation Protection and Research
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• 제45권4호
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• pp.178-186
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• 2020

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program-neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes-ILW, LLW, and VLLW.