미량 물 누출에 의한 2.25Cr-1Mo 시편 및 Target 에서의 반응특성

정경채(Jeong Kyung Chai),김병호(Kim Byung Ho),황성태(Hwang Sung Tai) 한국공업화학회 2000 공업화학 Vol.11 No.1

N/A

고점성 모사용액 내 Carbon Black 입자의 분산특성

정경채 ( Kyung Chai Jeong ),엄성호 ( Sung Ho Eom ),김연구 ( Yeon Ku Kim ),조문성 ( Moon Sung Cho ) 한국공업화학회 2013 공업화학 Vol.24 No.2

An external gelation method in place of an internal gelation method applied to the fabrication process of an intermediated compound of Uranium Oxy-Carbide (UCO) kernel spheres for Very High Temperature Reactor (VHTR) fuel preparation is under development in Korea. For the preliminary experiments of the UCO kernel sphere preparation using an external gelation method, the carbon black dispersion experiments were carried out using a simulated broth solution. From the selection experiments of various kinds of carbon black through dispersion experiments in a viscous metal salt solution, Cabot G carbon black was selected owing to its dispersion stability, and the homogeneous dispersing state of carbon black particles in our system. For the effective dispersion of nano-size aggregated carbon black particles in a high viscous liquid, the carbon black particles in a metal salt solution were first de-aggregated with ultrasonic force. The mixed solution was then dispersed secondly by the use of the extremely high-speed agitation with a mechanical mixer of 6000 rpm after feeding the Poly Vinyl Alcohol (PVA) in the solution. This results in the broth solution with good stability and homogeneity alongside no further changes in physical properties.

Self - wastage 에 의한 2.25Cr-1Mo Steel 시편의 Reopen 현상

정경채(Kyung chai Jeong),권상운(Sang woon Kwon),최종현(Jong hyeun Choi),박진호(Jin ho Park),황성태(Sung tai Hwang) 한국공업화학회 1999 공업화학 Vol.10 No.4

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나트륨-물 반응에 의한 전열관 표면에서의 부식과 반응특성에 대한 연구

정경채 ( Kyung Chai Jeong ),정지영 ( Ji Yong Jeong ),김병호 ( Byung Ho Kim ),김태준 ( Tae Joon Kim ),최종현 ( Jong Hyeun Choi ),김의식 ( Eui Sik Kim ) 한국공업화학회 2003 공업화학 Vol.14 No.1

본 연구에서는 증기발생기 전열관 재질로 예상되는 페라이트 강을 사용해서, 액체 나트륨 분위기에서 미량 물누출 실험을 수행하여, 시편의 누출 부위와 표적표면에서 발생되는 재질의 부식현상. 나트륨-물 반응에 의한 온도 상승현상 등에 대한 특성을 관찰하였다. 실험결과, 시편의 누출구멍 주변은 나트륨부위에서 나트륨-물 반응에 의해 발생되는 부식 및 침식에 의한 wastage 현상에 의해 손상을 입어 누출 구멍의 직경이 확대되었으며, 누출 부위에서의 손상은 초기 누출구멍 크기인 150㎛에서 200㎛정도로 확대되었고, 표적표면에 나타난 손상부위의 크기는 약 1000㎛ 정도로 분석되었다. 또한, 표적표면에서의 온도는 나트륨-물 반응에 의해 형성되는 반응열에 의해 약 70-80℃ 정도 상승하였으며, 손상부위 표명을 EPMA 및 EDX로 분석한 결과, 나트륨 산화물과 전열관의 모지 금속을 구성하는 Fe이나 Cr 등의 원소와 반응하여 (NaOH+Na_2O)ㆍFe_xO_y나 (NaCrO_2+Na_2Cro_4) 같은 형태의 복합산화물을 형성하는 것으로 판명되었다. In this study we observed and analyzed the corrosion phenomena in a leak site and he effect of temperature at the target surface of heat transfer tube material by sodium-water reaction through water leakage in liquid sodium atmosphere. Surrounding of the leak hole of the specimen was damaged by the wastage phenomena due to corrosion and erosion of material, and the hole size, by damage of leak site, was extended from 150㎛to about 200㎛. Also, the damaged size appeared in the target surface was observed to be about 1000㎛ diameter. The temperature of target surface increased about 70~80℃ by the heat of sodium-water reaction even with a small leakage of 15mL water. From the results of EPMA and EDX analysis, damaged areas are found to be covered with complex compounds such as (NaOH+Na_2O)·Fe_xO_y or (NaCrO_2+Na_2CrO_4) mixed up with the sodium oxides by sodium-water reaction and the Fe or Cr element by matrix of tube material.

핵연료 분말제조 공정에서 발생하는 폐액의 처리에 관한 연구

정경채,김태준,최종현,박진호,황성태 ( Kyung Chai Jeong,Tae Joon Kim,Jong Hyun Choi,Jin Ho Park,Seong Tae Hwang ) 한국공업화학회 1996 공업화학 Vol.7 No.6

현재 국내에서 가동중인 원자력발전소 공급용 핵연료 분말제조 공정에서 발생되는 폐액의 물성과 처리방법에 대한 연구가 수행되었다. 중수로형과 경수로형 발생 폐액에 함유된 우라늄을 회수/처리하기 위하여, 공히 폐액 속의 탄산이온의 제거가 필수적이다. 중수로형은 ADU 형태로 경수로형의 경우 UO₄화합물 형태로 처리하는 것이, 최종 폐액의 우라늄 농도를 최소화할 수 있었다. 처리후 폐액의 우라늄 농도는 중수로형 폐액의 경우, 폐액을 가열하여 ADU를 제조한 후 여액에 lime을 처리하는 방법으로 1ppm까지, 경수로형 폐액의 경우 UO₄·2NH₄F형태로 우라늄을 침전시킬 경우 0.8ppm까지 여액중의 우라늄 농도를 낮출 수 있었다. 최적 처리조건은 중수로형 폐액의 경우 101℃까지 단순 가열방법이, 경수로형 폐액의 경우 가열한 후 60℃에서 암모니아로 pH를 9.5로 조절한 후 과산화수소 용액을 첨가하여 1시간 반응시키는 경우로 나타났다. 폐액으로부터 회수된 우라늄 화합물은, 중수로형 폐액인 경우 pH가 낮을수록 회수된 ADU 입자의 크기가 증가하였으며, 경수로형 폐액인 경우 회수된 uranium peroxide 화합물을 공기분위기에서 열분해시킨 결과 기존의 AUC 분말이 열분해되어 나타내는 특성과 동일한 특성을 보임에 따라 핵연료분말 제조공정으로 recycle이 가능한 것으로 판단되었다. Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for CO₂ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as UO₄·2NH₄F compounds. Optimum condition was found at 101℃ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at 60℃ after heating the waste in order to expelling CO₂. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWR waste. Also, in case of uranium peroxide compound recovered from PWR waste, the property of U₃O_8 powder obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.

겔침전과 화학증착법에 의한 구형 UO₂ 입자와 TRISO 피복입자 제조

정경채,김연구,오승철,조문성,Jeong, Kyung-Chai,Kim, Yeon-Ku,Oh, Seung-Chul,Cho, Moon-Sung 한국세라믹학회 2010 한국세라믹학회지 Vol.47 No.6

HTGR using a TRISO coated particles as nuclear raw fuel material can be used to produce clean hydrogen gas and process heat for a next-generation energy source. For these purposes, a TRISO coated particle was prepared with 3 pyro-carbon (buffer, IPyC, and OPyC) layers and 1 silicone carbide (SiC) layer using a CVD technique on a spherical $UO_2$ kernel surface as a fissile material. In this study, a spherical $UO_2$ particle was prepared using a modified sol-gel method with a vibrating nozzle system, and TRISO coating fabrication was carried out using a fluidized bed reactor with coating gases, such as acetylene, propylene, and methyltrichlorosilane (MTS). As the results of this study, a spherical $UO_2$ kernel with a sphericity of 1+0.06 was obtained, and the main process parameters in the $UO_2$ kernel preparation were the well-formed nature of the spherical ADU liquid droplets and the suitable temperature control in the thermal treatment of intermediate compounds in the ADU, $UO_3$, and $UO_2$ conversions. Also, the important parameters for the TRISO coating procedure were the coating temperature and feed rate of the feeding gas in the PyC layer coating, the coating temperature, and the volume fraction of the reactant and inert gases in the SiC deposition.

핵연료분말 제조공정에서 발생된 여액으로부터 우라늄 회수 및 회수된 우라늄 화합물의 열분해 특성

정경채,정지영,김병호,김태준,최종현,Jeong, Kyung-Chai,Jeong, Ji-Young,Kim, Byung-Ho,Kim, Tae-Joon,Choi, Jong-Hyeun 한국세라믹학회 2002 한국세라믹학회지 Vol.39 No.2

본 연구에서는 AUC 공정에서 발생되는 액체폐기물에 미량 함유되어 있는 우라늄을 회수/재사용하기 위해 액상에서 침전법을 이용하여 용해도가 작은 우라늄화합물을 얻었으며, 이 화합물에 대한 chemical analysis, thermal analysis, x-ray diffraction analysis 및 FT-IR 분석을 통해 물성 특성을 해석하였다. 연구결과, 화학분석 및 FT-IR 분석으로부터 우라늄화합물은 $UO_4{\cdot}2NH_4F$ 형태를 가지고 있음을 알 수 있었으며, 평균 2∼3${\mu}m$ 입자 크기를 갖는 hexagonal 형태를 나타내었다. 열 분해시 분해 온도에 따라 중간물질로 $UO_4F,\;UO_4,\;UO_3,\;U_3O_8$ 등으로 변환되었으며, 상온에서 800$^{\circ}C$까지의 공기분위기에서 일정한 가열속도로 열분해시킬 경우, $UO_4{\cdot}2NH_4F{\rightarrow}UO_4F{\rightarrow}UO_4{\rightarrow}UO_3{\rightarrow}U_3O_8$의 반응 메커니즘을 나타내었다. In this study, $UO_4{\cdot}2NH_4F$, the precipitates which has low solubility, was obtained by chemical precipitation method to recover and reuse the trace uranium from the liquid waste producing in AUC process and for this compound it was characterized by means of chemical analysis, TG-DTA, XRD and FT-IR analyses. This compound was analyzed as $UO_4{\cdot}2NH_4F$ and shape of this precipitate was hexagonal type, having the size of 2∼3 ${\mu}m$. Also, the intermediates were obtained as $UO_4F,\;UO_4,\;UO_3,\;and\;U_3O_8$ by the thermal decomposition over the temperature of 220, 310, 515 and 640$^{\circ}C$, respectively. It is concluded that under the condition of a constant heating rate of 5$^{\circ}C$/min in air atmosphere range of between room temperature and 800$^{\circ}C$, thermal decomposition reaction mechanism of $UO_4{\cdot}2NH_4F$ is as follow; $UO_4{\cdot}2NH_4F{\rightarrow}UO_4F{\rightarrow}UO_4{\rightarrow}UO_3{\rightarrow}U_3O_8$.

변형 Sol-Gel 방법을 이용한 고온가스로 핵연료 미세구입자 제조

정경채,김연구,오승철,조문성,Jeong, Kyung-Chai,Kim, Yeon-Ku,Oh, Seung-Chul,Cho, Moon-Sung 한국세라믹학회 2009 한국세라믹학회지 Vol.46 No.6

$UO_2$ microsphere particles, core material of HTGR(High Temperature Gas Reactor) nuclear fuel, were prepared using by the GSP(Gel Supported Precipitation) method which is a modified-method of the wet sol-gel process. The spherical shape of initial liquid ADU droplets from the vibration nozzle system was continuously kept till the conversion to the final $UO_2$ microsphere. But the size of a final $UO_2$ microsphere was shrunken to about 25% of an initial ADU droplet size. Also, we found that the composition of dried-ADU gel particles was constituted of the very complicated phases, coexisted the U=O, C-H, N-H, N-O, and O-H functional groups by FT-IR. The important factors for obtain the no-crack $UO_2$ microsphere during the thermal treatment processes must perfectly wash out the remained-$NH_4NO_3$ within the ADU gel particle in washing process and the selections of an appropriate heating rate at a suitable gas atmosphere, during the calcining of ADU gel particles, the reducing of $UO_3$ particles, and the sintering of $UO_2$ particles, respectively.

고온가스로용 핵연료 UO₂ Kernel 입자제조

정경채,김연구,오승철,조문성,Jeong, Kyung-Chai,Kim, Yeon-Ku,Oh, Seung-Chul,Cho, Moon-Sung 한국세라믹학회 2007 한국세라믹학회지 Vol.44 No.8

The broth solution was prepared by the mixing of an uranyl nitrate, THFA, PVA, and water. The uranium concentration of the broth solution was $0.5{\sim}0.8$ mole-U/L and the viscosity of it was $30{\sim}80cSt$. The droplets of this broth solution were farmed in air and ammonia by the vibrating nozzle with the frequency of 100 Hz at the amplitude of $100{\sim}130V$. The diameter of the droplet was about $1900{\mu}m$ from using the nozzle diameter of 1 mm. The diameter of the aged gel was about $1400{\mu}m$ after aging in ammonia solution at $60{\sim}80^{\circ}C$, and the dried gel with the diameter of about $900{\mu}m$ was obtained after drying at room temperature or partially vacuum condition. The diameter of the calcined $UO_3$ microsphere after calcination at $600^{\circ}C$ appeared about $800{\mu}m$ in air atmosphere. Although the droplets of the same sizes were formed, the calcined microspheres of different sizes were manufactured in the case of the broth solutions of the different uranium concentration. The droplets of the desired diameters were obtained by the change of the nozzle diameters and the broth flow rates.

우라늄 정광의 용해 / 정제 및 핵연료 분말 가공공정에서 발생된 폐액의 처리에 관한 연구

정경채,황성태 ( Kyung Chai Jeong,Seong Tae Hwang ) 한국공업화학회 1997 공업화학 Vol.8 No.1

핵연료분말 변환공정 중 우라늄 정광의 용해/정제 및 가공공정에서 발생하는 폐액의 처리에대한 연구가 수행되었다. 우라늄 정광의 용해/정제공정에서 발생된 폐액은 pH 1 이하의 강산성으로 AUC 분말 제조공정에서 발생된 폐액 중의 우라늄을 ADU 형태로 회수한 후 발생된 2차 여액 속의 미세 ADU 입자 용해를 위해 사용된다. 2차 여액 속의 미세 ADU 입자들의 용해를 위해 용해/정제 공정의 폐액을 사용해서 pH 4로 전처리한 후, lime을 이용하여 pH 9.2로 30분 정도 반응시킬 경우 여액 중의 우라늄 농도를 3ppm 이하로 처리할 수 있었다. 가공 폐액은 미세 oil droplet들이 emulsion 형태로 발생하며, 약 300ppm의 우라늄 농도를 나타내었다. 먼저, emulsion을 파괴시키는 방법은 질산을 가하여 급속가열시키는 것이 효과적이었다. Emulsion 파괴 후 1mole NaOH를 가하여 Na₂U₂O_7 형태로 우라늄을 회수하였으며, pH 11.5에서 최적 처리조건을 나타내었으나 최종 여액 중의 우라늄 농도는 5ppm을 나타냈다. 여액 중의 우라늄 농도를 최소화하기 위해 lime으로 처리하는 방법이 연구되었으며, 가공폐액을 직접 lime 처리하기 위해 4N 질산으로 emulsion을 파괴 시킨 후, pH 1.6에서 lime을 1.5g/100ml로 반응시킬 경우 여액 중의 우라늄 농도를 1ppm까지 낮출 수 있었다. 한편, 경수로형 분말 제조공정 중 우라늄 회수공정에서 발생된 폐액 중의 미량 우라늄은 NaOH를 가하여 우라늄을 침전시킨 결과, Na·U·F·NH₄ 등이 혼합된 침전물이 얻어졌으며, 여과후 상등액에서는 우라늄은 감지할 수 없었다. This study provides the treatment methods of liquid wastes from the dissolution/purification process of nuclear fuel raw material and the fabrication process of nuclear fuel powder. One of the treatment methods is to process liquid waste from uranium raw material dissolution/purification process. This waste, of the strong acid, can be reused to dissolve the fine ADU particles in filtrate which is ADU waste of pH 8.0 converted from AUC waste after recovery of uranium. To dissolve the fine ADU particles, ADU filtrate was pretreated to pH 4.0 with the dissolution/purification waste, and then mixed with the lime to pH 9.2 and aged for 30 minutes. From this processing, uranium content of the filtrate was decreased to below 3ppm. The waste from fuel powder fabrication is emulsified solution dispersed with fine oil droplets. This emulsion was destroyed effectively by adding and mixing the nitric acid with rapid heating at the same time. After this processing, Na₂U₂O_7 compound is produced by addition of NaOH. Optimum condition of this processing was shown at pH 11.5, and uranium content of the filtrate was analyzed to 5ppm. To remove the trace of uranium in the filtrate, lime should be added. Otherwise, 4N nitric acid was used to destroy the emulsion directly, and then lime was added to this waste. Uranium content of the treated filtrate was below 1 ppm. In addition to these wastes, the trace of uranium in filtrate after recovery of uranium from the AUC waste which is produced during PWR powder preparation, is treated with NaOH to takeup fluorine(F) in the waste because fluorine is valuable and environmental toxic material. In the finally treated waste, uranium was not detected.