• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1436-1443
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• 2019

A fast gamma-ray dose rate assessment method for complex geometries based on stylized model reconstruction and point-kernel method is proposed in this paper. The complex three-dimensional (3D) geometries are imported as a 3DS format file from 3dsMax software with material and radiometric attributes. Based on 3D stylized model reconstruction of solid mesh, the 3D-geometrical solids are automatically converted into stylized models. In point-kernel calculation, the stylized source models are divided into point kernels and the mean free paths (mfp) are calculated by the intersections between shield stylized models and tracing ray. Compared with MCNP, the proposed method can implement complex 3D geometries visually, and the dose rate calculation is accurate and fast.

• 방사성폐기물학회지
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• 제6권2호
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• pp.73-100
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• 2008

고준위 방사성폐기물 처분장으로부터 유출된 핵종에 의한 선량률을 계산하기 위한 생태계평가 코드 ACBIO를 일반적인 구획모델링도구인 AMBER를 이용하여 BIOMASS 방법론을 적용하여 개발하였다. ACBIO의 유용성을 보이고, 구획의 변화나 일부 파라미터값의 변화에 따른 구획 내 농도와 방사능, 그리고 구획간의 플럭스의 민감도도 검토하였다. 지하매질-생태계 경계(GBI)를 통해 넘어오는 핵종의 유출플럭스에 따른 선량환산인자를 각 핵종별로 구하여 결정집단내 개인의 최대피폭선량율을 선량환산인자로 얻는 계산을 수행하였다. 또한 생태계 요소의 구획모델링이나 가능한 피폭집단의 설정, 그리고 GBI의 인지 등이 생태계평가에 중요한 요소가 되는 것을 확인하였다.

• Journal of Radiation Protection and Research
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• 제34권3호
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• pp.144-150
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• 2009

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2085-2091
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• 2020

The license for Kori-1, the first commercial reactor in Busan, Korea, was terminated in June 2017; therefore, preparations are being made for its decommissioning. Because the radioactivity of Bio-shield varies greatly throughout the structure, the doses received by the workers depend on the location, order, and duration of dismantling operations. Thus, a model for evaluating the worker external dose during the dismantling of the Kori-1 bio-shield was developed, and work scenarios for dose assessment were designed. The Dose evaluation code VISIPLAN was used for dose assessment. The dose rate around the bio-shield was evaluated and the level of exposure to the operator was evaluated according to the work scenario. The maximum annual external dose was calculated as 746.86 mSv for a diamond wire saw operator under dry cutting conditions, indicating that appropriate protective measures, such as changing dismantling sequence, remote monitoring, shield installation, and adjustment of work team are necessary for the safe dismantling of the bio-shield. Through these protective measures, it was found that the worker's dose could be below the dose limit.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.325-326
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• 2017

• Journal of Radiation Protection and Research
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• 제30권4호
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• pp.161-165
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• 2005

The gamma-ray dose rates in air at 233 locations in Korea have been determined. The contribution to the gamma-ray dose rates in air due to the presence of $^{232}Th-series,\;^{238}U-series\;and\;^{40}K$ is as follows: 47.3% $(36\;nGyh^{-1})\;^{232}Th-series$ 14.5% $(11\;nGyh^{-1})\;^{238}U-series$ and 38.2% $(29\;nGyh^{-1})\;^{40}K$. The mean gamma-ray dose rate theoretically derived from $^{232}Th-series,\;^{238}U-series\;and\;^{40}K\;was\;76{\pm}17\;nGyh^{-1}$. This corresponds to an annual effective dose of $410\;{\mu}Sv$ and an annual collective dose of 18900 person-Sv for all provinces under study. The results have been compared with other global radiation dose.

• Journal of Radiation Protection and Research
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• 제40권4호
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• pp.252-260
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• 2015

본 논문은 ICRP103 환경방호를 대비하여 국내에서 개발된 야생동식물 선량평가 코드 K-BIOTA의 기술적 배경 및 적용 사례를 기술한다. K-BIOTA는 스크리닝(screening) 선량평가(Level 1&2)와 부지 특성적 상세평가(Level 3)의 3단계의 단계적 평가방법을 적용한다. 스크리닝 단계평가는 상세평가의 필요성 여부를 판단하기 위한 예비적 평가 단계로 개별적인 생물종보다는 동식물을 그룹별로 구분하여 평가한다. Level 1 평가는 스크리닝 목적의 참조준위로부터 유도된 최대환경매체농도 값과 실제 환경매체농도 값의 비교로부터 위험도 지표(risk quotient)를 계산한다. Level 1 평가 결과 위험도 지표가 1보다 작으면 생태계의 건전성이 유지된다는 결론과 함께 평가를 종료하고, 1 보다 크면 동식물 그룹별로 평균적인 전이계수나 평형분배계수 값을 적용하여 조금 더 실제적인 Level 2 평가를 수행한다. 따라서 Level 2 평가가 Level 1 평가보다 덜 보수적이다. Level 2평가에서 위험도지표가 1 보다 작으면 생태계의 건전성이 유지된다는 결론을 내리고 평가를 종료하고, 1보다 크면 Level 3 평가를 수행한다, Level 3 평가는 부지 특성적 데이터를 고려하는 상세평가단계로, 동식물 그룹별 평가 대신 부지 대표적 동식물에 대한 개별적 선량평가를 수행하며, 대표적 동식물의 종류 및 크기, 거주인자, 전이계수, 평형분배계수에 대한 부지 특성적인 값을 사용한다. 또한 Level 3 평가 단계에서는 전이계수, 평형분배계수, 환경매체농도 (토양농도 또는 물의 농도)에 대한 개별 동식물의 피폭 선량률에 대한 불확실도 분석을 선택적으로 수행할 수 있다. 적용 가능한 확률밀도함수는 정규분포, 로그정규분포, 균일분포, 지수분포의 4가지이다. 국제원자력기구의 EMRAS II (Environmental Modeling for Radiation Safety) 모델 시나리오 비교 공동연구에 참가하여 K-BIOTA의 적용성을 검증하였다. 그 결과로 K-BIOTA는 다양한 오염 환경에서 거주하는 야생동식물의 방사선 영향을 평가하는데 유용함이 입증되었다.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.273-281
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• 2021

The radiological safety of the spent resin treatment facility with a¹⁴C treatment capacity of 1 ton/day was evaluated in terms of the external and internal exposure of worker according to operation scenario. In terms of external dose, the annual dose for close work for 1 h/day at a distance of more than 1 m (19.8 mSv) satisfied the annual dose limit. For 8 h of close work per day, the annual dose exceeded the dose limit. For remote work of 2000 h/year, the annual dose was 14.4 mSv. Lead shielding was considered to reduce exposure dose, and the highest annual dose during close work for 1 h/day corresponded to 6.75 mSv. For close work of 2000 h/year and lead thickness exceeding 1.5 cm, the highest value of annual dose was derived as 13.2 mSv. In terms of internal exposure, the initial year dose was estimated to be 1.14E+03 mSv when conservatively 100% of the nuclides were assumed to leak. The allowable outflow rate was derived as 7.77E-02% and 2.00E-01% for the average limit of 20 mSv and the maximum limit of 50 mSv, respectively, where the annual replacement of the worker was required for 50 mSv.

• Journal of Radiation Protection and Research
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• 제31권1호
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• pp.1-7
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• 2006

The measured mean activities of $^{226}Ra,\;^{232}Th,\;^{40}K\;and\;^{137}Cs$ in the soil of Bahawalpur, Bahawalnagar and Rahimyar Khan Bistricts were 32.9, 53.6, 647.4 and 1.8 Bq $kg^{-1}$. The average absorbed dose rate calculated from these activities was 74.3 nGy $h^{-1}$ and the mean annual effective dose rate was found to be 0.46 mSv $y^{-1}$. Absorbed doses to different body organs were derived from annual effective doses using tissue weighting factors. Radiation induced fatal cancer risks were assessed by using ICRP 60 Model. Estimations incurred 184deaths per year due to cancer.

• 방사성폐기물학회지
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• 제19권1호
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• pp.133-140
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• 2021

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.