• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.339-339
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• 2004

원자로를 이용하여 장수명핵종(long lived nucleus)을 소멸처리하는 과정에서 초우라늄(TRU, transuranium)과 희토류(RE, rare earth) 금속에 포함되어 있는 소량의 핵분열성(fissile) 물질인 우라늄을 제거할 필요가 있다. 본 실험은 LiCl-KCl 용융염계에서 전해제련법(Electrowinning)을 이용하여 용융염욕에 존재하는 우라늄을 제거하기 위하여 필요한 Cd-Li 양전극 물질을 제조하였고, 제조된 금속을 이용하여 우라늄 및 란탄족(Dy, Ce, Y, Nd, Gd) 금속의 환원 특성을 파악하였다.(중략)

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.293.1-293.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.466.1-466.1
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.466.2-466.2
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.337-337
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• 2004

As one of researches for the P ＆ T purposes, a basic experiment on the recovery of actinide elements from the mixture with rare earth elements by means of electrorefining using a liquid cadmium cathode in the LiCl-KC1 eutectic melt was carried out. In order to examine the behaviors of electrodeposition of metal ions on a liquid electrode, recovery experiments of rare earth metals resulting from forming electrodeposits were performed by a galvanostatic electrolysis method at various current densities. A cyclic voltammetric technique was applied to determine reduction-oxidation potential of each metal element in the melt and to detect the changes of the multi component melt composition for on-line monitoring. Also, a collaboration study with RIAR was completed to test the preliminary feasibility on a recovery of actinide elements from the mixture with rare earth elements using a liquid cadmium cathode and actinide metals. Experimental results showed that the ratio of actinides to rare earths, 9: 0.5∼1 led to the rare earth content of about 5∼10 wt% in the deposit.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.14 no.6
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• pp.35-40
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• 2015

An IP-2 type container is used for the transport of low-level radioactive materials. An IP-2 type metal container was modeled and analyzed for the dropping conditions in normal transport conditions required by NSSC Notice 2014-50, "Regulations on the packaging and transport of radioactive materials." Structural analyses were performed by finite element method, and the acceptability was reviewed by comparing the results with technical standards and maximum allowable stresses of each material. Structural members of the container were modeled as 3D solid elements, and the possible dropping directions were considered in these analyses. Results show that maximum stresses, maximum deformed gaps in the cover, and maximum estimated thickness reduction in structural members were well below the allowable limits; thus, the structural integrity of the container was confirmed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.39-50
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• 2005

Electrolytic reduction of uranium oxide to uranium metal was studied in a $LiCl-Li_{2}O$ molten salt system. The reduction mechanism of the uranium oxide to a uranium metal has been studied by means of a cyclic voltammetry. Effects of the layer thickness of the uranium oxide and the thickness of the MgO on the overpotential of the cathode and the anode were investigated by means of a chronopotentiometry. From the cyclic voltamograms, the decomposition potentials of the metal oxides are the determining factors for the mechanism of the reduction of the uranium oxide in a $LiCl-3\;wt{\%} Li_{2}O$ molten salt and the two mechanisms of the electrolytic reduction were considered with regards to the applied cathode potential. In the chronopotentiograms, the exchange current and the transfer coefficient based on the Tafel behavior were obtained with regard to the layer thickness of the uranium oxide which is loaded into the porous MgO membrane and the thickness of the porous MgO membrane. The maximum allowable currents for the changes of the layer thickness of the uranium oxide and the thickness of the MgO membrane were also obtained from the limiting potential which is the decomposition potential of LiCl.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.32-37
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• 2005

The characteristics of the metal melting and radionuclide distribution of the radioactive has been investigated in a lab-scale arc furnace. The slag former based on the constituents of silica, calcium oxide, aluminum oxide, borate and calcium fluoride additions was used for melting of the stainless and carbon steel. In the melting of the stainless steel, the amount of slag formation increased with an increase of the concentration of the slag former. But the effects of the slag basicity on the amount of stag formation showed a local maximum value of the slag formation with an increase of the basicity index in the melting of the stainless steel as well as in the melting of the carbon steel. With an increase of the amount of slag former addition, the trends of the cobalt distribution into the ingot and the stag depended on the kind of slag former used in the melting of the stainless steel while the effect of the slag basicity on the distribution of the cobalt was not clarified in the melting of carbon steel. Tn the melting of the carbon steel, the strontium was captured at up to $50\%$ into the slag phase. Cesium was completely eliminated from the melt of the stainless steel as well as the carbon steel and distributed to the dust phase.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.435-443
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• 2016

This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.45-57
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• 2014

Academic interests in ionic liquid (IL) technologies have been extended to the nuclear industry and the applicability of ionic liquids for processing radioactive materials have been investigated by many researchers. A number of studies have reported interesting results with respect to the spectroscopic and electrochemical behaviors of metal elements included in spent nuclear fuels. The measured and observed properties of metal ions in TBP(tri-butyl phosphate) dissolved ILs have led the development of alternative technologies to traditional aqueous processes. On the other hand, the electrochemical deposition of metal ions in ILs have been investigated for the application of the solvents to aqueous as well as to non-aqueous processes. In this work, a review on the application of ILs in nuclear fuel cycle is presented for the purpose of categorizing and summarizing the notable researches on ILs.