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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.
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.
유로퓸 고용(固溶) 우라늄산화물(酸化物)의 황화반응(黃化反應) 특성(特性)
이재원,박근일,이정원,Lee, Jae Won,Park, Geun Il,Lee, Jung Won 한국자원리싸이클링학회 2013 資源 리싸이클링 Vol.22 No.3
희토류산화물인 $Eu_2O_3$, 우라늄산화물인 $UO_2$ 및 $U_3O_8$, $Eu_2O_3$와 우라늄산화물의 혼합물에 대한 선택적 황화반응을 조사한 후에, $(U,Eu)O_2$ 및 $(U,Eu)_3O_8$와 같은 Eu 고용 우라늄산화물, Eu 고용 우라늄산화물의 고온 산화열처리 상분리 생성물인 Eu 농도가 높은 $(U,Eu)_4O_9$와 $U_3O_8$의 혼합상에 대한 황화반응 특성을 $400-800^{\circ}C$에서 조사하였다. $Eu_2O_3$ 및 우라늄산화물의 혼합물의 경우에는 $450^{\circ}C$에서 Eu와 우라늄 산화물간의 반응이 없이 $Eu_2O_3$만 $Eu_3S_4$로 전환되었다. $(U,Eu)_3O_8$ 및 $(U,Eu)O_2$에서는 반응온도 $600^{\circ}C$까지는 우라늄산화물과 동일한 황화반응 거동을 보였으며, $800^{\circ}C$에서는 Eu 농도가 높은 $(U,Eu)S_x$과 ${\alpha}-US_2$ 상이 생성되었다. 고온 산화열처리 상분리 생성물은 $600^{\circ}C$에서 $(U,Eu)S_x$과 UOS 상이 생성되었다. 상분리 생성물을 환원하여 얻은 Eu 농도가 높은 $(U,Eu)O_2$와 $UO_2$의 혼합상은 $450^{\circ}C$에서 $(U,Eu)O_2$은 산황화물인 (U,Eu)OS로 전환되고 $UO_2$는 반응하지 않았다. Sulfurization reaction characteristics of $Eu_2O_3$, uranium oxides($UO_2$, $U_3O_8$), mixture of $Eu_2O_3$ and uranium oxides, Eu-doped uranium oxides($(U,Eu)O_2$, $(U,Eu)_3O_8$), and phase-separated products prepared by HOX (High temperature OXidation) of $(U,Eu)O_2$ were investigated in the temperature range from 400 to $800^{\circ}C$. Only $Eu_2O_3$ in the mixture of $Eu_2O_3$ and uranium oxides was converted into $Eu_3S_4$ by sulfurization reaction at $450^{\circ}C$ without reaction between them. Sulfurization reaction behavior of $(U,Eu)_3O_8$ and $(U,Eu)O_2$ up to $600^{\circ}C$ was similar to $U_3O_8$ and $UO_2$, respectively, while they were sulfurized into Eu-rich $(U,Eu)S_x$ and ${\alpha}-US_2$ at $800^{\circ}C$. In the sulfurization of RE-rich $(U,Eu)_4O_9$ and $U_3O_8$ prepared by high temperature oxidation, it was confirmed that RE-rich $(U,Eu)S_x$ and UOS phases were formed at $600^{\circ}C$. For Eu-rich $(U,Eu)O_2$ and $UO_2$ prepared by reduction of HOX products, it was identified that Eu-rich (U,Eu)OS was formed at $450^{\circ}C$ by sulfurization of Eu-rich $(U,Eu)O_2$, while $UO_2$ remained unreacted.
단일 전해액 배출만을 가지는 pH조절용 연속식 이온 교환막 전해 시스템의 개발과 그 특성
김광욱,김인태,박근일,이일희,Kim Kwang-Wook,Kim In-Tae,Park Geun-Il,Lee Eil-Hee 한국방사성폐기물학회 2005 방사성폐기물학회지 Vol.3 No.4
In order to produce only a pH-controlled solution without discharging any unwanted solution, this work has developed a continuous electrolytic system with a pH-adjustment reservoir being placed before an ion exchange membrane-equipped electrolyzer, where as a target solution was fed into the pH-adjustment reservoir, some portion of the solution in the pH-adjustment reservoir was circulated through the cathodic or anodic chamber of the electrolyzer depending on the type of the ion exchange membrane used, and some other portion of the solution in the pH-adjustment reservoir was discharged from the electrolytic system through the other counter chamber with its pH being controlled. The internal circulation of the pH-adjustment reservoir solution through the anodic chamber in the case of using a cation exchange membrane and that through the cathodic chamber in the case of using an anion exchange membrane could make the solution discharged from the other counter chamber effectively acidic and basic, respectively. The phenomena of the pH being controlled in the system could be explained by the electro-migration of the ion species in the solution through the ion exchange membrane under a cell potential difference between anode and cathode and its consequently-occurring non-charge equilibriums and electrolytic water- split reactions in the anodic and cathodic chambers.
우라늄 및 희토류(稀土流) 산화물(酸化物)의 황화반응(黃化反應)에 대한 열역학적(熱力學的) 고찰(考察)
이정원,이재원,강권호,박근일,Lee, Jung-Won,Lee, Jae-Won,Kang, Kweon-Ho,Park, Geun-Il 한국자원리싸이클링학회 2012 資源 리싸이클링 Vol.21 No.1
우라늄 및 희토류(RE: rare-earth) 산화물의 황화반응에 대한 $M-O_2-S_2$ 상태도 및 Gibbs 자유에너지 변화(${\Delta}G^{\circ}$)와 같은 열역학적 특성 자료를 비교, 분석하여 우라늄 및 회토류 산화물의 혼합상에서 황화반응에 의해 희토류산화물만 희토류황화물로의 선택적 반응이 가능한지를 조사하였다. 황화제로는 $CS_2$가 적합하였는데, $CS_2$는 다른 황화제보다 강한 황화재이며 반응온도를 낮출 수 있다. $CS_2$를 황화제로 이용하여 $U_2-O_2-S_2$ 및 $RE-O_2-S_2$의 상태도를 비교한 결과, $UO_2$은 반응하여 UOS로 전환되며 희토류산화물은 반응하여 희토류황화물이 되었다. 희토류산화물의 황화반응에 의한 ${\Delta}G^{\circ}$는 우라늄산화물의 ${\Delta}G^{\circ}$보다 낮았다. 따라서 희토류와 우라늄 산화물 혼합상은 $300{\sim}800^{\circ}C$에서의 황화반응 시에 평형상태에서 우라늄산황화물과 희토류황화물이 우선적으로 생성된다. In order to evaluate the feasibility of selective sulfidization of uranium and rare-earth(RE) oxides, an analysis on thermodynamic data, such as $M-O_2-S_2$ phase stability diagram and changes of Gibbs free energy for sulfidization of uranium and rare-earth oxides were carried out. Comparing $RE-O_2-S_2$ with $U-O_2-S_2$ phase stability diagram at wide range of sulfur potential, $UO_2$ remains unreacted, while RE oxides are sulfidized. The Gibbs free energy change(${\Delta}G^{\circ}$) of sulfidization of RE oxides is lower than that of uranium oxides. Thus, the selective formation of RE sulfides is possible during sulfidization of RE and uranium oxides at lower temperature. $CS_2$ was selected as a sulfidizing agent, because it is a stronger sulfidizing agent than other agents and reacts at lower 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.