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사용후핵연료의 연소도 변화에 따른 산화 및 OREOX 공정에서 핵분열기체 방출 특성
박근일,조광훈,이정원,박장진,양명승,송기찬,Park, Geun-Il,Cho, Kwang-Hun,Lee, Jung-Won,Park, Jang-Jin,Yang, Myung-Seung,Song, Kee-Chan 한국방사성폐기물학회 2007 방사성폐기물학회지 Vol.5 No.1
Quantitative analysis on release behavior of the $^{85}Kr\;and\;^{14}C$ fission gases from the spent fuel material during the voloxidation and OREOX process has been performed. This thermal treatment step in a remote fabrication process to fabricate the dry-processed fuel from spent fuel has been used to obtain a fine powder The fractional release percent of fission gases from spent fuel materials with burn-up ranges from 27,000 MWd/tU to 65,000 MWd/tU have been evaluated by comparing the measured data with these initial inventories calculated by ORIGEN code. The release characteristics of $^{85}Kr\;and\;^{14}C$ fission gases during the voloxidation process at $500^{\circ}C$ seem to be closely linked to the degree of conversion efficiency of $UO_2\;to\;U_3O_8$ powder, and it is thus interpreted that the release from grain-boundary would be dominated during this step. The high release fraction of the fission gas from an oxidized powder during the OREOX process would be due to increase both in the gas diffusion at a temperature of $500^{\circ}C$ in a reduction step and in U atom mobility by the reduction. Therefore, it is believed that the fission gases release inventories in the OREOX step come from the inter-grain and inter-grain on $UO_2$ matrix. It is shown that the release fraction of $^{85}Kr\;and\;^{14}C$ fission gases during the voloxidation step would be increased as fuel burn-up increases, ranging from 6 to 12%, and a residual fission gas would completely be removed during the OREOX step. It seems that more effective treatment conditions for a removal of volatile fission gas are of powder formation by the oxidation in advance than the reduction of spent fuel at the higher temperature.
IRN-150 혼상수지의 이온 흡착특성 및 폐수지로부터 탈착용액을 이용한 $^{14}C$ 핵종의 제거 특성
양호연,원장식,최영구,박근일,김인태,김광욱,송기찬,박환서,Yang, Ho-Yeon,Won, Jang-Sik,Choi, Young-Ku,Park, Geun-Il,Kim, In-Tae,Kim, Kwang-Wook,Song, Kee-Chan,Park, Hwan-Seo 한국방사성폐기물학회 2006 방사성폐기물학회지 Vol.4 No.4
Spent ion-exchanged resin generated from various purification systems in CANDU reactor was contaminated with high activity of $^{14}C$ radionuclide. This paper describes the results of fundamental study to develop the applicable technology for the treatment of this spent resin. Based on the adsorption capacity of inactive $HCO_3$ ion and other anions on IRN-150 mixed resin, the removal characteristics of $HCO_3$ ion adsorbed on to IRN-150 by various stripping solutions were evaluated. Maximum adsorption amount of the $HCO_3$ ion onto IRN-150 raw resin was about 11 mg-C/g-resin which agrees with the theoretical adsorption amount of this resin. Adsorption affinity of various anions such as $CS,\;CO,\;Na\;NH_4$ was analyzed in single and multi-component systems. From the results of removal characteristics of the $HCO_3$ ion adsorbed on IRN-150 by various stripping solutions, $NH_4H_2PO_4$ stripping solution is more effective than $NaNO_3,\;Na_3PO_3$ solutions for the complete removal of $^{14}C$ radionuclide from the IRN-150 spent resin.
고연소도 경수로 사용후핵연료의 열처리에 따른 세슘 방출거동
박근일,조광훈,이정원,박장진,양명승,송기찬,Park, Geun-Il,Cho, Kwang-Hun,Lee, Jung-Won,Park, Jang-Jin,Yang, Myung-Seung,Song, Kee-Chan 한국방사성폐기물학회 2007 방사성폐기물학회지 Vol.5 No.1
The dynamic release behavior of Cs from high burn-up spent PWR fuel was experimentally performed under the conditions of a thermal treatment process such as voloxidation and sintering conditions. In voloxidation process, influence of the oxidation and reduction atmosphere on the Cs release characteristic using fragment type of spent fuel heated up to $1,500^{\circ}C$ was compared. In sintering process, temperature history effect on Cs release behavior was evaluated using green pellet under 4% $H_2/Ar$ environment. Temperature range for complete Cs release from spent fuel fragment under voloxidation condition was about $800^{\circ}C{\sim}1,200^{\circ}C$, but that of green pellet under the reduction atmosphere was $1,100^{\circ}C{\sim}1,400^{\circ}C$. Key parameters on Cs release behavior from spent fuel was powder formation as well as the diffusion rate of Cs compound to grain boundary and fuel surface.
월성 원전발생 폐수지로부터 제거된 $^{14}C$ 핵종의 인산용액을 이용한 $^{14}CO_2$로의 기체화 특성
양효연,원장식,최영구,박근일,김인태,김광욱,송기찬,박환서,Yang, Ho-Yeon,Won, Jang-Sik,Choi, Young-Ku,Park, Geun-Il,Kim, In-Tae,Kim, Kwang-Wook,Song, Kee-Chan,Park, Hwan-Seo 한국방사성폐기물학회 2006 방사성폐기물학회지 Vol.4 No.4
Removal characteristics of $H^{14}CO_3$ ion from IRN-150 mixed resin contaminated with $^{14}C$ radionuclide and a gasification behavior of $^{14}C$ radionuclide to $^{14}CO_2$ were investigated. The stripping solutions used for the removal of $^{14}C$ from spent resin were $NaNO_3,\;Na_3PO_4,\;NH_4H_2PO_4,\;H_3PO_4$. The influence of stripping solution concentration on the desorption characteristics of inactive $HCO_3$ ion into stripping solution from IRN-150 mixed resin and the gasification of this ion to $CO_2$ was analyzed. The gasification behavior to $CO_2$ by using NaOH, $HNO_3$, HCl was also compared to that of phosphate solution. Real spent resin stored in Wolsung nuclear power plant was used to evaluate the gasification characteristics of $^{14}C$ radionuclide to $^{14}CO_2$. Gamma radionuclides such as $^{137}Cs,\;^{60}Co$ in residual striping solutions after desorption experiment were analyzed.
윤완기,송기찬,이재설 한국경영과학회 1989 한국경영과학회 학술대회논문집 Vol.- No.1
The simulation of the nuclear spent fuel interim storage system is performed with statistical discrete event models. Proposed components and process sequence of the system are simulated with experimental input date for various cases of the system operation. This will provide the informations on the facility design and the best operation scheme with reference to the design. Critical parameters to affect the system design and operation are analyzed.
김영민,오원진,송기찬,이근우,최왕규 한국공업화학회 1998 응용화학 Vol.2 No.1
The removal of silver in nitric acid solution was carried out by the electrodeposition in a conventional three electrode H-cell in order to investigate the possibility of the direct electrolytic removal as an alternative to a chemical treatment process. The silver ion could be removed In nitric acid concentrations of less than 3 M with a good removal efficiency. The lower concentration of the nitric acid and the lower temperature, the greater removal efficiency could be obtained. Potential shift occurred as the time when the hydrogen evolution reaction took place along with the depletion of almost all silver in the nitric acid solution and it could be known that the electrolysis should be completed to obtain the maximum removal efficiency. Consequently, the possibility of the silver removal by electrodeposition in nitric acid solution as an alternative to a chemical treatment process was confirmed.