• Nuclear Engineering and Technology
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• 제54권1호
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• pp.234-243
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• 2022

This study aims to characterize the irradiated RBMK-1500 nuclear graphite in terms of both structural and radiological properties. The experimental results of morphological and structural analysis of the irradiated graphite samples by using SEM, Raman spectroscopy as well as the theoretical evaluation of primary displacement damage are presented. Moreover, the experimental and theoretical evaluation of the neutron flux is provided and the presence of several γ emitters in the analyzed graphite samples is assessed. Furthermore, the improved version of rapid analysis method for ¹⁴C activity determination is applied and the experimentally obtained results are compared with calculated ones. Results indicate that structural changes are uniform enough in all the analyzed samples. However, the distribution of radionuclides is non-homogeneous in the irradiated RBMK-1500 reactor graphite matrix. The comprehensive understanding of both structural and radiological characteristics of nuclear graphite is very important when dealing with decision about irradiated graphite waste management strategy or treatment options prior to its final disposal.

• Journal of Radiation Protection and Research
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• 제45권2호
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• pp.81-87
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• 2020

Background: Kori unit #1 is permanently shut down after a 40-year lifetime. The Nuclear Safety and Security Commission recommends establishing initial decommissioning plans for all nuclear and radwaste treatment facilities. Therefore, the Korea Atomic Energy Research Institute (KAERI) must establish an initial and final decommissioning plan for radwaste-treatment facilities. Radiation safety assessment, which constitutes one chapter of the decommissioning plan, is important for establishing a decommissioning schedule, a strategy, and cost. It is also a critical issue for the government and public to understand. Materials and Methods: This study provides a method for assessing external radiation dose to workers during decommissioning. An external dose is calculated following each exposure scenario, decommissioning strategy, and working schedule. In this study, exposure dose is evaluated using the deterministic method. Physical characterization of the facility is obtained by both direct measurement and analysis of the drawings, and radiological characterization is analyzed using the annual report of KAERI, which measures the ambient dose every month. Results and Discussion: External doses are calculated at each stage of a decommissioning strategy and found to increase with each successive stage. The maximum external dose was evaluated to be 397.06 man-mSv when working in liquid-waste storage. To satisfy the regulations, working period and manpower must be managed. In this study, average and cumulative exposure doses were calculated for three cases, and the average exposure dose was found to be about 17 mSv/yr in all the cases. Conclusion: For the three cases presented, the average exposure dose is well below the annual maximum effective dose restriction imposed by the international and domestic regulations. Working period and manpower greatly affect the cost and entire decommissioning plan; hence, the chosen option must take account of these factors with due consideration of worker safety.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.278-284
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• 2023

The synthetic B₂O₃-As₂O₃ glass ceramic are prepared to investigate the physical properties and the radiation shielding capabilities with the variation of concentration of the As₂O₃ with 10, 20, 30, and 40%, respectively. XRD analyses are performed on the fabricated glass-ceramic and depicted the improvement of crystallinity by adding As₂O₃. The radiation shielding properties are studied for the B₂O₃-As₂O₃ glass ceramic. The values of linear attenuation coefficient (LAC) are varied with the variation of incident photon gamma energy (23.1-103 keV). The LAC values enhanced from 12.19 cm^-1-37.75 cm^-1 by raising the As₂O₃ concentration from 10 to 40 mol% at low gamma energy (23.1 keV) for BAs10 and BAs40, respectively. Among the shielding parameters, the half-value layer, transmission factor, and radiation protection efficiency are estimated. Furthermore, the fabricated samples of glass ceramic have low manufacturing costs and good shielding features compared to the previous work. It can be concluded the B₂O₃-As₂O₃ glass ceramic is appropriate to apply in X-ray or low-energy gamma-ray shielding applications.

• 방사성폐기물학회지
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• 제4권2호
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• pp.187-195
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• 2006

최근에 개정된 중 저준위 방사성폐기물 인도규정에서는 폐기물시설운영자는 처분 전 방사성폐기물이 인도규정에 적합함을 보장하기 위한 계획을 수립하여 운영할 것을 요구하고 있다. 앞으로 건설될 방사성폐기물 처분시설에서 방사성폐기물 인수 시에 충족되어야할 방사성폐기물인수기준이 수립될 것이며, 이 인수기준에는 물리화학적 및 방사선학적 특성규명, 물리화학적인 제한사항, 금지품목, 포장, 인식표, 라벨, 문서요건 등을 규정할 것으로 예상된다. 이들 규정을 준수하기 위하여 방사성폐기물 발생자는 방사성폐기물이 폐기물 인수기준에 만족하는지를 확인하는 폐기물 인증프로그램을 수행하여 방사성폐기물이 그 기준에 만족하는지 확인하여야 하며, 방사성폐기물에 대한 모든 인증서류를 폐기물 처분시설에 제출할 책임이 있다. 본 방사성폐기물 인증프로그램계획은 한국원자력연구소에서 발생되는 방사성폐기물의 인증프로그램에 대한 예비프로그램으로서 개발된 것이며 앞으로 처분장의 인수요건이 구체적으로 확립될 때까지 수정 보완되어야 한다.

• 한국방사선학회논문지
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• 제13권5호
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• pp.791-799
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• 2019

선량 증강 현상에서 발생하는 물리적 특성과 증강 물질과의 상호 작용으로부터 발생하는 이차입자 생성을 평가하였다. Geant 4, MIRD 두부 팬텀을 이용한 몬테카를로 전산 모사를 진행하였으며, 선형가속기에서 발생되는 4, 6, 10, 15, 18, 25 MV X선을 선원으로 적용하였다. 10, 20, 30 mg/g의 금(aurum), 가돌리늄(gadolinium) 증강 물질을 팬텀 내부 종양에 모사하였으며, 물리적 상호작용의 변화와 이차입자 발생에 따른 입자플루언스와 초기 에너지로부터 방사선가중인자를 고려하여 등가선량을 평가하였다. 방사선 선량 증강 물질에 의한 상호작용은 고 원자번호에서 기인하여 광전효과에 의한 에너지 흡수를 높이는 것으로 나타났으며, 10 MV 이상의 에너지에서는 광핵반응의 증가를 나타내었다. 이로 인해, 팬텀 내부에서 양성자, 중성자와 같은 이차입자 발생의 증가를 보였으며, 중성자에 의한 등가선량이 최대 424.2배 증가하는 것으로 나타났다. 본 연구는 선량 증강 현상에서의 에너지 전달, 흡수의 물리적 과정을 모사하여, 증강 현상에서 발생하는 물리적 특성을 분석하고자 하였다. 이러한 결과는 향후 in-vivo, in-vitro 선량 증강 실험을 위한 기초 자료로 활용될 수 있을 것으로 사료된다.