• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.105-112
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• 2004

A waste container, one of the key components of a multi-barrier system in a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic medium and the swelling pressure of the bentonite buffer. Also, it delays potential release of radionuclides for a certain period of time, before it is corroded by intruding impurities. Even though the material of a waste container is carefully chosen and its manufacturing processes are under quality assurance processes, there is a possibility of initial defects in a waste container during manufacturing. Also, during the deposition of a waste container in a repository, there is a chance of an incident affecting the integrity of a waste container. In this study, the appropriate Features, Events, and Processes(FEP's) to describe these incidents and the associated scenario on radionuclide release from a container to the biosphere are developed. Then the total system performance assessment on the Initial waste Container Failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set used in this paper, the annual individual dose for the ICF scenario meets the Korean regulation on the post closure radiological safety of a repository.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.109-116
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• 2003

The amount of radioactive waste is expected to be increased continuously because of the rapid growth of the domestic nuclear industry, full power operation of the HANARO reactor and the increased research activities of the nuclear fuel cycle. Accordingly the efforts are focused to achieve the handling of radioactive waste in safe and reduce the volume of radioactive waste. The PIEF is carrying out the PIE (post irradiation examination) of spent fuel rods related to the identification of cause defect and evaluation of integration safety. This study describes the technologies and experiences of compaction, shredding and cutting of the solid radioactive waste used in the PIE. The quantity of the high level waste was reduced by 1/12 using the 100-ton compressor installed in hot-cell. Also middle and low level waste was reduced by 1/8 using the 60-ton compressor installed in intervention area. Plastic drums were shredded by crusher to be compacted in the ratio of 1/5, used filters in the ratio of 1/6 and the number of drum is also reduced by cutting procedure for the non-volatile materials such as metal.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.544-548
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• 2003

The Irradiation test of bar code label tagged on radioactive waste container was done to determine the effect of radiation. Low and medium radioactive waste is that below total activity of 4,000 Bq/g according to the Korean nuclear law. The irradiation amount to radiate bar code label tagged on radioactive waste container was calculated by MCNP-4b computer code. The nuclide such as Co-60 and Cs-137 was assumed to contribute 50% of total activity. Real irradiation amount for bar code label was finally calculated by the dimensions of the container and the bar code label. The Identification of post and the physical deflection of irradiated bar code label was tested by the bar code reader. The coated bar code label was suitable to use on low and medium radioactive waste container.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.157-167
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• 2020

Removing radioactive concrete is crucial in the decommissioning of nuclear power plants. However, this process generates radioactive aerosols, exposing workers to radiation. Although large amounts of radioactive concrete are generated during decommissioning, studies on the internal exposure of workers to radioactive aerosols generated from the cutting of radioactive concrete are very limited. In this study, therefore, we calculate the internal radiation doses of workers exposed to radioactive aerosols during activities such as drilling and cutting of radioactive concrete, using previous research data. The electrical-mobility-equivalent diameter measured in a previous study was converted to aerodynamic diameter using the Newton-Raphson method. Furthermore, the specific activity of each nuclide in radioactive concrete 10 years after nuclear power plants are shut down was calculated using the ORIGEN code. Eventually, we calculated the committed effective dose for each nuclide using the IMBA software. The maximum effective dose of ¹⁵²Eu constituted 83.09% of the total dose; moreover, the five highest-ranked elements (¹⁵²Eu, ¹⁵⁴Eu, ⁶⁰Co, ²³⁹Pu, ⁵⁵Fe) constituted 99.63%. Therefore, we postulate that these major elements could be measured first for rapid radiation exposure management of workers involved in decommissioning of nuclear power plants, even if all radioactive elements in concrete are not considered.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.54-59
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• 2003

Several radionuclides are considered as an object of the assesment to develop a scaling factor and a periodical verification method which are needed for the evaluation of radionuclide inventory of various radioactive wastes from nuclear power plants in Korea. A selective separation of $^{55}Fe$, $^{90}FSr$ 및 $^{94}Nb$ which should be recovered individually for the radiochemical analysis was described in detail. Sorption and desorption behaviours of ion exchange and extraction chromatographic resins for Fe, Sr, Nb and co-existing metal ions were Investigated using an artificial waste solution simulated of chemical composition of real radioactive wastes. Separation conditions available for the sequential recovery of these metal ions from a single sample were optimized to minimize a discharge of radioactive wastes produced through the analytical process and a radiation exposure to analysts. Their recovery yields were measured with reliability.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.81-84
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• 2004

방사성폐기물 처분 사업을 둘러싼 논란은 지난 50 여년간 국내외에서 계속되었다. 세계 각국은 뛰어난 과학 기술의 도입과 대규모의 홍보에도 불구하고 많은 경우 처분 사업의 실패를 맛보았다. 그러나 핀란드, 스웨덴 등 일부 유럽 국가들은 다른 나라들에 비해 비교적 적은 예산과 기술 인력을 투입함에도 불구하고 부지 확보 및 안전성 확보에서 괄목할만한 성과를 거두고 있다. 본 논문에서는 이러한 현상에 주목해 방사성폐기물 처분 연구 사업에서 쌍 방향 대화의 중요성을 간파하고 이를 적용하는 방안을 연구한 유럽 연합의 RISCOM II 프로젝트 중 프랑스, 핀란드, 스웨덴 사례에 대해 살펴본다.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.14 no.2
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• pp.143-150
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• 2001

원자력발전소의 1차 계통에서 오염된 장비들을 취급이 용이하고 안전하게 운반하기 위한 운반용기는 내부의 방사성 물질에 대한 방사능 평가에 의하여 방사성물질 A형 운반용기로 분류된다. 방사성물질 A형 운반용기는 IAEA Safety Standard Series No. ST-1 및 국내 원자력법 등 관련규정의 기술기준을 만족하여야 하는데, 운반용기는 중량에 따라 0.3∼1.2m의 높이에서 소성이 일어나지 않는 단단한 바닥면으로 가장 심각한 손상을 주는 방향으로 낙하시키는 정상운반조건(normal transport conditions)에 대하여 구조적 건전성을 유지하여야 한다. 여기서는 ABAQUS/Explicit 코드를 이용하여 컨테이너형태의 A형 운반용기에 대하여 최대손상이 야기되는 0.9m 경사낙하조건에 대한 3차원 충격해석을 수행하고 구조적 건전성을 평가하였는데, 운반용기는 경사낙하시 코너피팅(corner fitting)의 분쇄(crush)에 의하여 대부분의 충격을 흡수하였으며 운반용기의 격납경계는 구조적 건전성을 유지하였다.

• Nuclear industry
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• v.23 no.1 s.239
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• pp.64-78
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• 2003

한국원자력연구소에서는 과학기술부에서 주관하는 국제 협력 기반 조성 사업 과제의 일환으로 한국-핀란드 양국간 원자력 협력 증진을 위한 프로젝트를 수행하고 있다. 특히 이 연구에서는 방사성 폐기물 관리와 관련된 양국간 이해 증진과 향후 협력을 모색하기 위한 방안을 수립하고자 하였다. 본 연구에서는 이와 같은 관점에서 세계 최초로 사용후 핵연료 영구 처분장 부지를 확보하고 우리나라와 지질 조건이 유사한 결정질 암반에 신규로 심지층 처분 연구 실증 시설인 온칼로(Onkalo) 프로젝트를 계획하고 있는 핀란드의 방사성 폐기물 관리기관인 POSIVA 등과 관련 협력 기관, 정부 기관 등과 함께 향후 구체적인 협력 방안을 모색하고, 핀란드의 사용후 핵연료 직접 처분 연구사업 계획을 벤치 마킹하여 2003년도에 시작하는 국내 고준위 방사성 폐기물 처분 연구 과제 계획 수립에 도움을 주고자 하였으며, 이와 병행하여 핀란드 신규 원전 사업과 관련된 국내 산업체의 참여 가능성을 타진해 보고자 하였다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.485-495
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• 2005

This study was conducted to investigate sorption and desorption behaviors of radionuclides (Cobalt and Strontium) in natural soil. Sorption kinetics and isotherms were analyzed to predict sorption behaviors of radionuclides in natural soil and the experimental data were fitted to several sorption models. Desorption experiments were also performed with or without CMCD at constant pH and ion strength conditions. The results showed that $Sr^{2+}$ was more strongly sorbed than $Co^{2+}$ in natural soil. Both $Co^{2+}$ and $Sr^{2+}$ followed a pseudo-second order kinetics and Sips model. The desorption-resistance of $Co^{2+}$ and $Sr^{2+}$ was estimated using a natural surfactant Carboxymethyl-${\beta}$-cyclodextrin(CMCD) or non-desorbing fraction. Desorption of radionuclides was partially irreversible and $Sr^{2+}$ was more resistant than $Co^{2+}$ Addition of CMCD facilitated desorption of $Co^{2+}$ and $Sr^{2+}$ from soil.

• Tunnel and Underground Space
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• v.7 no.3
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• pp.246-252
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• 1997

원자력 발전 과정에서 부산물로 발생되는 사용후 핵연료와 같은 고준위 방사성폐기물은 수백 만년동안 인간 및 자연 환경에 영향ㅇ르 미치기 때문에 엄격한 관리가 요구된다. 이를 위하여 세계 각국에서는 KBS-3 개념과 같이 고준위 방사성폐기물을 지하 500미터 심도의 암반에 영구 처분하기 위하여 연구를 수행사고 있다. 이러한 연구 활동의 일환으로 고준위 방사성폐기물에서 발생하는 방사성 붕괴열로 인한 처분장 인접 암반에서의 응력 변화 및 이에 따른 주변 암반대에서의 지하수 유동 현상 규명을 위한 연구가 지난 1980년대부터 활발하게 진행되고 있는 바, 그 대표적인 연구 과제가 DECOVALEX 국제 공동 연구이다. 이 글에서는 현재 진행 중인 DECOVALEX 연구 현황과 향후 전망에 관하여 조명하고자 한다.