• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.152-153
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• 2009

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.322.1-322
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• 2002

• Proceedings of the Korean Information Science Society Conference
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• 2011.06c
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• pp.9-12
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• 2011

한국원자력연구원은 연구로 1, 2호기와 우라늄변환시설의 해체활동에서 얻어진 해체 경험과 그간에 쌓인 해체활동에 관한 정보 그리고 해체 대상 시설의 특성자료를 통하여 해체 작업 시 발생하는 유닛 생산성(Unit Productivity)을 자동으로 산출하는 시스템(DEWOCS)을 개발하였다. DEWOCS를 통한 해체작업시 발생하는 유닛의 해체작업 생산성 평가 인자 값은 해체 엔지니어링 시스템의 개발 및 향후 이루어질 원자력발전소를 비롯한 원자력시설의 해체 사업에서 계획 및 해체 설계에 기초자료로써 활용될 것이다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.11-23
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• 2003

A study on the electrolytic dissolution of SUS-304 and Inconel-600 specimen was carried out in neutral salt electrolyte to evaluate the applicability of electrochemical decontamination process for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant in Korea. Although the best electrolytic dissolution performance for the specimens was observed in a Na2s04 electrolyte, a NaNO$_3$ neutral salt electrolyte, in which about 30% for SUS-304 and the same for Inconel-600 in the weight loss was shown in comparison with that in a Na$_2$SO$_4$ solution, was selected as an electrolyte for the electrochemical decontamination of metallic wastes with the consideration on the surface of system components contacted with nitric acid and the compatibility with lagoon wastes generated during the facility operation. The effects of current density, electrolytic dissolution time, and concentration of NaNO$_3$ on the electrolytic dissolution of the specimens were investigated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as UO$_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO$_3$ solution with the current density or In mA/$\textrm{cm}^2$. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO$_3$ neutral salt electrolyte by reducing $\alpha$ and $\beta$ radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.232-232
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• 2004

우라늄 변환시설 내의 라군 슬러지의 처리를 위해 슬러지의 물 첨가 용해를 실시하고, 여과후 발생한 질산염의 안정적 처리를 위한 열분해를 실시하였다. 라군 슬러지의 질산염 및 우라늄 제거공정은 후속처리공정에서의 부담을 최소화 할 수 있도록 1.5배의 물을 첨가 용해하였으며, 두 개의 라군에 저장된 슬러지 처리방법의 효율성 평가를 위하여 각 라군의 개별적, 혹은 혼합하여 실험을 실시하였다.(중략)

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.47-48
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• 2010

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.05a
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• pp.33-33
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• 2007

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.65-66
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• 2009

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.72-80
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• 2004

Electrolytic dissolution study was carried out to evaluate the applicability of electrochemical decontamination process using a neutral salt electrolyte as a decontamination technology for the recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant using SUS-304 and Inconel-600 specimen as the main materials of internal system components of the plant. The effects of type of neutral salt as an electrolyte, current density, and concentration of electrolyte on the dissolution of the materials were evaluated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as $UO_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion plant were peformed in $Na_2SO_4$ and $NaNO_3$ solution. It was verified that the electrochemical decontamination of the dismantled metallic wastes was quite successful in $Na_2SO_4$ and $NaNO_3$ neutral salt electrolyte by reducing $\beta$ radioactivities below the level of self disposal with authorization within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.79-86
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• 2009

Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. Decommissioning and dismantling of these facilities produce considerable quantities such as concrete structure, rubble. In this paper, the characteristics distribution of the radionuclide have been investigated for the effects of the heating and grinding test for aggregate size such as gravel, sand and paste from decommissioning of the TRIGA MARK II research reactor and uranium conversion plant. The experimental results showed that most of the radionuclide could be removed from the gravel, sand aggregate and concentrated into a paste. Especially, we found that the heating temperature played an important role in separating the radionuclide from the concrete waste. Contamination of concrete is mainly concentrated in the porous paste and not in the dense aggregate such as the gravel and sand. The volume reduction rate could be achieved about 80% of activated concrete waste and about 75% of dismantled concrete waste generated from UCP.