질산과 옥살산 용액에서 세륨옥살레이트의 침전평형

정동용,김응호,신영준,유재형,김종득 ( Dong Yong Chung,Eung Ho Kim,Young Joon Shin,Jae Hyung Yoo,Jong Duk Kim ) 한국공업화학회 1995 공업화학 Vol.6 No.4

질산과 옥살산 용액 중에서 세륨옥살레이트 침전 평형농도를 구하였다. 평형농도는 0∼2.0M의 질산농도 중의 세륨니트레이트 용액에 옥살산을 첨가하여 세륨옥살레이트를 침전시킨 후, 여과액의 세륨농도를 ICP를 이용하여 분석함으로써 구하였다. 수용액상에서 세륨이 자유 옥살레이트 이온과 Ce(C₂O₄)_n^(3-2n)(n=1, 2, 3) 형태의 착물을 형성한다고 하여 이를 기초로 한 용해도 모델을 사용했으며, 용해도 예측을 위한 침전 후 용액의 이온강도, 옥살산농도, 산농도 등을 침전반응의 화학양론적 물질수지를 이용하여 계산하였다. 이때 필요한 각 이온들의 활동도계수는 수정된 Debye-Hu¨ckel식을 이용하여 구하였다. 재시된 용해도 모델 및 계산방법을 이용하여 침전 후 세륨옥살레이트의 평형농도를 계산한 결과 실험치를 잘 예측하고 있음을 볼 수 있었다. The precipitation equilibrium of cerium oxalate in nitric acid and oxalic acid was investigated. The cerium oxalate was precipitated by addition of oxalic acid to cerium nitrate solution in 0∼2.0M nitric acid. The precipitates to be turned cut were filtered by using 0.2㎛ membrane filter. The concentration of cerium in filtrate was determined by ICP(Induced Coupled Plasma Spectroscopy). In aqueous solution, we can suppose that cerium ion can be reacted with free oxalate ion to form Ce(C₂O₄)_n^(3-2n)(n=1, 2, 3) complex, in that case it can be proposed the solubility model of cerium oxalate hydrate. Next, in order to estimate the solubility of cerium oxalate. we should be gotten the value of ionic strength, oxalic acid and hydrogen ion concentration in precipitation reaction. Also we use the modified Debye-Hu¨ckel equation to calculate the activity coefficients of three valence ions. Concentration values estimated from the solubility model agreed with experimental data well.

모의 방사성용액에서 란탄족과 악티늄족원소의 옥살산침전

정동용(Dong Yong Chung),김응호(Eung Ho Kim),이일희(Eil Hee Lee),유재형(Jae Hyung Yoo),박현수(Hyun Soo Park) 한국공업화학회 1999 공업화학 Vol.10 No.7

N/A N/A

질산용액으로부터 이온성 액체를 이용한 Am(III)과 Eu(III)의 추출 거동

김익수,정동용,이근영,이일희,Kim, Ik-Soo,Chung, Dong-Yong,Lee, Keun-Young,Lee, Eil-Hee 한국방사성폐기물학회 2018 방사성폐기물학회지 Vol.16 No.3

The applicability of room temperature ionic liquids (RTILs), 1-alkyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([$C_nmim$] [$Tf_2N$]), was investigated for the extraction of Am(III) and Eu(III) from nitric acid using n-octyl(phenyl)-N,N-diisobutyl carbamoylmethyl phosphine oxide (CMPO) and tri-n-butylphosphate (TBP) as extractants. The distribution ratios of Am(III) and Eu(III) in CMPO-TBP/[$C_nmim$][$Tf_2N$] were measured as a function of various parameters such as the concentrations of nitric acid, CMPO, and TBP. The results were compared with those obtained in CMPO-TBP/n-dodecane (n-DD). With comparable concentrations of the extractants, the distribution ratios obtained with RTILs were much higher than those obtained with n-DD. It was observed that the extraction efficiency was less for Eu(III) than for Am(III). The extraction of Am(III) and Eu(III) decreased with increases in the feed acidity for all three RTILs. The results suggest that the extraction of Am(III) and Eu(III) by CMPO in RTILs from nitric acid proceeds through the cation-exchange mechanism. The distribution ratios of Am(III) and Eu(III) increased with increases in the concentration of CMPO for all three RTILs. A linear regression analysis of the extraction data resulted in a straight line with a slope of about 3, suggesting the involvement of 3 molecules of CMPO during the extraction process.

$(Zr-DEHPA)/n-dodecane-HNO_3$ 금속함유 추출 계에 의한 악티나이드(III)및 RE의 공추출 및 상호 분리

이일희,임재관,정동용,양한범,김광욱,Lee, Eil-Hee,Lim, Jae-Kwan,Chung, Dong-Yong,Yang, Han-Beom,Kim, Kwang-Wook 한국방사성폐기물학회 2007 방사성폐기물학회지 Vol.5 No.2

This study was performed to evaluate the co- and mutual separation for Am, Cm and RE elements from the simulated multi-component solution equivalent to real HLW level by a Zr-DEHPA(di-(2-ethylhexyl) phosphoric acid containing Zirconium)/$NDD(n-dodecane)-HNO_3$ extraction system. Zr-DEHPA was self-synthesized and the optimal condition of (15g/L Zr-1M DEHPA)/NDD-1M $HNO_3$ was selected taking into consideration of prevention of the third phase, and effects of concentration of DEHPA, nitric acid and impregnant amount of Zr on the co-extraction of Am, Cm and RE. In that condition, the extraction yields were 81% (Am), 85% (Cm), more than 80% (RE elements), 98% (Mo), 85% (Fe), 98% (U), 73% (Np), and less than 5% (other elements) so that the system developed for the co-extraction of Am-Cm/RE was proved to be available. For that, however, U, Np, Mo and Fe was elucidated to have to be removed in advance, and Zr inducing the third phase formation was found to be practically excluded. The co-extracted Am-Cm/RE were sequentially separated in an order of Am-Cm (stripping agent : 0.05 M DTPA-1M Lactic acid of pH 3.6)${\rightarrow}RE$ (stripping agent : 5M $HNO_3$), and then their separation factors were evaluated. At above conditions, Am of 65.4%, Cm of 63.9%, RE (except for Y) of more than 85% were stripped.

수용액상에서 과산화수소의 분해

김응호,김영환,정동용,신영준,유재형,최청송 ( Eung Ho Kim,Young Hwan Kim,Dong Yong Chung,Young Joon Shin,Jae Hyung Yoo,Cheong Song Choi ) 한국화학공학회 1996 Korean Chemical Engineering Research(HWAHAK KONGHA Vol.34 No.2

The Kinetics of the decomposition of H₂O₂ in the acidic solution was investigated. The investigation was conducted within the range of HNO₃ concentration of 0-4 M at temperature of 70, 80 and 90℃, respectively. The decomposition reaction is first order with respect to [H₂O₂] and is enhanced by acid-catalytic effect above HNO₃ of 2 M. The rates are as follows; In[H₂O₂]/[H₂O₂]_0= -2.23×10^(10) exp(-18200/RT) ·t ([H^+]<2M) In[H₂O₂]/[H₂O₂]_0=-[2.23×10^(10) exp(-18200/RT)+2.1×10^(12) exp(-21200/RT) ([H^+] -2)] ·t ([H^+]>2M) The effect of UO₂^(+2), Nd^(+3), Pd^(+2), F^(+3), MoO₂^(+2), Sr^(+2), and Cs^+ on the decomposition of H₂O₂ in the solution were examined too, and the rate was compared with that obtained from metal ion-free solution.

Na₂CO₃-H₂O₂ 탄산염 용액의 안정성 평가

이일희,임재관,정동용,양한범,김광욱,Lee, Eil-Hee,Lim, Jae-Gwan,Chung, Dong-Yong,Yang, Han-Beum,Kim, Kwang-Wook 한국방사성폐기물학회 2011 방사성폐기물학회지 Vol.9 No.3

연구는 $Na_2CO_3-H_2O_2$ 탄산염 용액의 숙성시간에 따른 안정성을 U의 산화 용해액, Cs/Re의 선택적 침전 여과액 및 U의 산성화 침전 여과액으로 구분하여 검토하였다. 숙성시간에 따른 산화 용해액 내 조성 변화는 거의 없었으며, Cs/Re의 선택적 침전에도 아무 영향이 없이 산화 용해액으로부터 순차적으로 Re과 Cs의 침전제거가 가능하였다. 그러나 U의 산성화 침전에서는 산화 용해액이나 Cs/Re의 선택적 침전 여과액을 장시간 동안 숙성시킬 경우 U이 uranyl peroxocarbonato complex에서 uranyltricarbonate로 일부 전환되어 U의 침전회수를 감소시켰다. 그러므로 99% 이상의 U을 회수하기 위해서는 산화 용해액 및 Cs/Re의 선택적 침전 여과액의 숙성시간을 각각 7일 이내에서 처리하는 것이 효과적이다. 그리고 SF 의 산화/용해${\rightarrow}$Cs과 Re(/Tc)의 선택적 침전${\rightarrow}$ U의 산성화 침전 등을 순차적으로 수행하여, 산화/용해에서는 대부분의 U과 FP 중 일부가 함께 용해 되었으며, 함께 용해된 FP 중 Re과 Cs은 각각 TPPCl 및 NaTPB로 99% 이상을 침전제거할 수 있었다. 그리고 산성화 (pH 4) 침전에서는 U을 거의 100% 침전회수 하여 $Na_2CO_3-H_2O_2$ 탄산염 용액에서 침전법으로 SF로부터 U 만의 회수 타당성을 확인하였다. This study was carried out to examine the stability of $Na_2CO_3-H_2O_2$ carbonate solution with aging time in the dissolving solution after oxidative dissolution of U by a carbonate solution, the Cs/Re filtrate after selective precipitation of Cs and Re (as a surrogate for Tc), and the acidification filtrate after precipitation of U by acidification, respectively. The compositions of dissolving solution were not changed with ageing time, and the selective precipitation of Re and Cs was also not affected without regard to the aging time of dissolving solution. The successive removal of Cs and Re from a carbonate dissolving solution was possible. However, the recovery yield of U by acidification was decreased with increasing the aging time of the dissolving solution and the Cs/Re-filtrate, respectively, because U was converted from the uranyl peroxocarbonato complex to the uranyltricarbonate in the solution aged for a long time. Accordingly, it is effective that a dissolving solution and a Cs/Re filtrate are treated within the aging of 7 days, respectively, in order to recover U more than 99%. On the other hand, the recovery of U was carried out in order of the oxidative dissolution of U selective precipitation of Re and Cs precipitation of U by acidification. Almost all of U and a part of FP were co-dissolved in oxidative dissolution step. Over 99% of Re and Cs from the FP co-dissolved with U could be removed by a TPPCl (tetraphenylphosphonium chloride) and a NaTPB (sodium tetraphenylborate), respectively. U was precipitated nearly 100% by acidification to pH 4. Therefore, it was confirmed that the validity of recovery of U by precipitation methods from a SF (spent fuel) in the $Na_2CO_3-H_2O_2$ solution.

질산매질에서 UV 조사에 의한 옥살산 분해

김응호,김영환,정동용,유재형 ( Eung Ho Kim,Young Hwan Kim,Dong Yong Chung,Jae Hyung Yoo ) 한국공업화학회 1997 공업화학 Vol.8 No.1

본 연구에서는 질산매질에서 UV 광조사에 의한 옥살산 분해연구가 수행되었다. UV 광원은 2537 의 파장을 방출하는 수은램프가 사용되었다. UV 광조사에도 불구하고 옥살산 자체는 분해되지 않았다. 그러나 질산매질하에서 UV 광조사에 의해 옥살산은 쉽게 분해되었다. UV 광조사에 의해 NO₃-으로부터 발생되는 산소라디칼이 옥살산을 분해시키는 것으로 조사되었다. 옥살산 분해율은 질산 0.5M 부근에서 최대를 이루다가 질산농도 증가에 따라 점차 감소하였다. 이것 역시 산소라디칼과 NO₃- 사이에서 반응으로 쉽게 설명될 수 있다. Decomposition of oxalic acid was studied in nitric acid media by using UV radiations. The UV source is Hg-lamp, emitting 2537 , wavelength. Oxalic acid was not decomposed by itself in spite of UV radiation, but in the presence of nitric acid decomposed easily under UV radiation. It is believed that oxygen radical generated from nitrate ion by UV radiation results in the decomposition of oxalic acid. Decomposition rate of oxalic acid reached a maximum in around 0.5M HNO₃ and then gradually decreased with nitric acid concentration. Thd decrease can be also explained to be due to the reaction between oxygen radical and NO₃-.

nPr-BTP/nitrobezene 추출 계에 의한 악티나이드(III)의 선택적 분리

이일희,임재관,정동용,양한범,김광욱,Lee, Eil-Hee,Lim, Jae-Kwan,Chung, Dong-Yong,Yang, Han-Beom,Kim, Kwang-Wook 한국방사성폐기물학회 2008 방사성폐기물학회지 Vol.6 No.1

A selective separation of Actirlide(III) by a nPr-BTP/nitrobezene extraction system was studied. The nPr-BTP (2.6-Bis-(5.6-n-propyl-1.2.4-triazin-3-yl)-pyridine) of a environmentally -friendly CHN type was self-synthesized and its compatability with diluent and stability with nitric acid were investigated. At the 0.1M nPr-BTP/nitrobenzene-1M $HNO_3$ and O/A=2, extraction yields of Am used as a representative of Actinide(III) and Eu were about 85% and 8%, respectively, and the other RE elements such as Nd, Ce and Y were extracted less than 3% (separation factor of Am and Eu was about 60). Thus, there was no problems in the selective extraction of Actinide(III) from RE. The stripping yield of Am with 0.05M $HNO_3$ at O/A= 1, however, was about 43% and the maximum stripping yield was 65% at O/A=0.3. It is necessary to develop the stripping system including the stripping agent instead of nitric acid solution.

알카리화 및 산성화에 의한 우라늄 함유 슬러지의 열분해 고체 폐기물로부터 우라늄 제거

이일희,양한범,이근영,김광욱,정동용,문제권,Lee, Eil-Hee,Yang, Han-Beom,Lee, Keun-Young,Kim, Kwang-Wook,Chung, Dong-Yong,Moon, Jei-Kwon 한국방사성폐기물학회 2013 방사성폐기물학회지 Vol.11 No.2

This study has been carried out to elucidate the characteristics of the dissolution for Thermal Decomposed Solid Waste of uranium-bearing sludge (TDSW), the removal of impurities by an alkalization in a nitric acid dissolving solution of TDSW, and the selective removal (/recovery) of uranium by an acidification in an carbonate alkali solution, respectively. TDSW generated by thermal decomposition of U-bearing sludge which was produced in the uranium conversion plant operation, was stored in KAERI as a solid-powder type. It is found that the dissolution of TDSW is more effective in nitric acid dissolution than oxidative-dissolution with carbonate. At 1 M nitric acid solution, TDSW was undissolved about 30wt% as a solid residue, and uranium contained in TDSW was dissolved more than 99%. In order to the alkalization for the nitric acid dissolving solution of TDSW, carbonate alkalization is more effective with respect to remove the impurities. At the carbonate alkali solution controlled to about 9 of pH, Al, Ca, Fe and Zn co-dissolved with U in dissolution step was removed about $98{\pm}1%$. On the other hand, U could be recovered more than 99% by an acidification at pH about 3 in a carbonate alkali solution, which was nearly removed the impurities, adding 0.5M $H_2O_2$. It was found that uranium could be selectively recovered (/removed) from TDSW. 본 연구는 우라늄 변환시설 운전 중에 발생된 우라늄 함유 슬러지를 가열 처리하여 분말 형태로 저장 중인 우라늄 함유 슬러지의 열분해 고체폐기물 (Thermal Decomposed Solid Waste of uranium-bearing sludge : TDSW)을 대상으로 TDSW의 용해, TDSW 질산 용해액의 알카리화에 의한 불순물 제거 및 탄산염 알카리화 용액의 산성화에 의한 U 선택적 제거/회수 특성 등을 규명하였다. TDSW의 용해는 질산용해가 탄산염 산화용해 보다 효과적이었다. 1M 질산에서 TDSW의 약 30wt%가 고체 잔류물로 불용해되었고, TDSW 내 함유 U은 99% 이상이 용해되었다. TDSW의 질산 용해액의 알카리화는 탄산염에 의한 알카리화가 불순물 제거 측면에서 보다 효과적이며, 탄산염 알카리화 (pH 약 9)에서 U과 공용해된 Ca, Al, Zn 및 Fe 등의 $98{\pm}1%$가 제거되었다. 그리고 불순물이 거의 제거된 알카리화 용액 (0.5 M $H_2O_2$ 첨가)의 산성화 (pH 약 3) 에서 U의 99% 이상을 회수할 수 있어 TDSW로부터 U을 선택적으로 제거/회수할 수 있었다.

Cs-흡착 CHA-Cs 및 CHA-PCFC-Cs 제올라이트계와 Sr-흡착 4A-Sr 및 BaA-Sr 제올라이트계의 고온 열분해

이일희,김지민,김형주,김익수,정동용,김광욱,이근영,서범경,Lee, Eil-Hee,Kim, Ji-Min,Kim, Hyung-Ju,Kim, Ik-Soo,Chung, Dong-Yong,Kim, Kwang-Wook,Lee, Keun-Young,Seo, Bum-Kyoung 한국방사성폐기물학회 2018 방사성폐기물학회지 Vol.16 No.1

For the immobilization of high-radioactive nuclides such as Cs and Sr by high-temperature thermal decomposition, this study was carried out to investigate the phase transformation with calcined temperature by using TGA (thermogravimetric analysis) and XRD (X-ray diffraction) in the Cs-adsorbed CHA (chabazite zeolite of K type)-Cs and CHA-PCFC (potassium cobalt ferrocyanide)-Cs zeolite system, and Sr-adsorbed 4A-Sr and BaA-Sr zeolite system, respectively. In the case of CHA-Cs zeolite system, the structure of CHA-Cs remained at up to $900^{\circ}C$ and recrystallized to pollucite ($CsAlSi_2O_6$) at $1,100^{\circ}C$ after undergoing amorphous phase at $1,000^{\circ}C$. However, the CHA-CFC-Cs zeolite system retained the CHA-PCFC-Cs structure up to $700^{\circ}C$, but its structure collapsed in $900{\sim}1,000^{\circ}C$, and then transformed to amorphous phase, and recrystallized to pollucite at $1,100^{\circ}C$. In the case of 4A-Sr zeolite system, on the other hand, the structure of 4A-Sr maintained up to $700^{\circ}C$ and its phase transformed to amorphous at $800^{\circ}C$, and recrystallized to Sr-feldspar ($SrAl_2Si_2O_8$, hexagonal) at $900^{\circ}C$ and to $SrAl_2Si_2O_8$ (triclinic) at $1,100^{\circ}C$. However, the BaA-Sr zeolite system structure began to break down at below $500^{\circ}C$, and then transformed to amorphous phase in $500{\sim}900^{\circ}C$ and recrystallized to Ba/Sr-feldspar (coexistence of $Ba_{0.9}Sr_{0.1}Al_2Si_2O_8$ and $Ba_{0.5}Sr_{0.5}Al_2Si_2O_8$) at $1,100^{\circ}C$. All of the above zeolite systems recrystallized to mineral phase through the dehydration/(decomposition) ${\rightarrow}$ amorphous ${\rightarrow}$ recrystallization with increasing temperature. Although further study of the volatility and leachability of Cs and Sr in the high-temperature thermal decomposition process is required, Cs and Sr adsorbed in each zeolite system are mineralized as pollucite, Sr-feldspar and Ba/Sr-feldspar. Therefore, Cs and Sr seen to be able to completely immobilize in the calcining wasteform/(solidified wasteform).