• Applied Chemistry for Engineering
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• v.7 no.6
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• pp.1164-1173
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• 1996

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 $CO_2$ 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 $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ 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 $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.415-420
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• 1998

시판중인 대표적 유·무기이온교환수지를 이용하여 방사성폐액의 주 방사성핵종인 코발트와 세슘에 대하여 방사성패액에 함유되어 있는 대표적 일반이온인 나트륨이온이 이들의 이온교환에 미치는 영향을 분석하였다. 나트륨 존재 하에서 세슘이온에 대한 선택도와 수지단위 부피당 폐액 처리 부피는 무기이온교환수지인 DT 30과 Durasil 230이 가장 높으며 반면에 유기이온교환수지는 두 수치 모두 낮아서 유기이온교환수지는 저농도의 나트륨 이온을 갖는 증발기 응축수의 처리에는 적합하나 나트륨이온의 농도가 비교적 높은 폐액처리에는 적합하지 않으며, 또한 코발트 제거 면에서는 무기이온교환수지인 DT 10 보다 유기이온교환수지인 Amberlite IRN 77이 바람직하다는 결론을 얻었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.375-380
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• 1997

모의 방사성 세탁폐액을 제조하여 오존에 의한 세제 파괴를 확인하고 활성탄 및 이온교환수지를 이용하여 세제 및 Co, Cs 제거율을 조사하였으며 모의 방사성 세탁폐액을 오존으로 부분적으로 산화ㆍ파괴시킨후 활성탄 및 이온교환수지에 의한 흡착 및 이온교환 실험을 수행하여 오존의 세제 파괴가 방사성 물질 제거에 미치는 영향을 조사하였다. 오존에 의해 세제는 75% 정도 제거될 수 있었고 활성탄으로 방사성 모의세탁폐액을 처리할 때 세제농도가 증가하면 방사성 핵종 제거율이 감소하였다. 이온교환수지로 세제를 제거할 때 성취가능 제거율은 Co의 경우 99% 이상이었으며, 세제 존재시 방사성 Co 및 Cs 제거율은 감소하며, 방사성 모의세탁폐액을 오존으로 조사후 활성탄과 이온교환수지로 방사성 핵종을 제거할 때 그 제거율은 거의 변화가 없었다. 이상과 같은 실험 결과로부터 오존으로 부분적으로 산화시켜 활성탄의 세제 제거효율을 최대화하고, 역삼투막에 의한 방사성 핵종을 제거하며 이온교환수지로 잔류 방사성 핵종을 완전히 처리할 수 있는 복합 공정을 도출하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.507-512
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• 1996

저준위 방사성 폐액의 전처리공정으로써 한외여과막의 적용성을 평가하기 위해서 폐액내에 있는 오일과 계면활성제의 상호작용을 규명하고 오일에멀젼 용액의 fouling정도를 조사하였다. 막의 fouling을 감소시키는 한 방법으로 계면활성제에 의해 막의 표면을 개질하므로서 막투과 flux는 크게 증가되는 효과를 얻었다. 친수성막과 소수성막에 대하여 몇가지 계면활성제로 처리한 후 성능을 비교한 결과 소수성인 폴리솔푼막에 대한 SDS의 표면개질이 가장 유리하였다. 표면처리 막의 적용성을 평가한 결과 미량의 계면활성제나 염이 포함된 오일에멀젼 용액에 대해서는 매우 우수한 투과 성능을 얻을 수 있으나 CMC 이상의 계면활성제가 포함된 오일용액에 대해서는 처리가 불가능하였다.

• The Science & Technology
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• v.11 no.2 s.105
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• pp.43-44
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• 1978

한국원자력 연구소방사성오염처리실장 이상훈 박사팀(김관식, 송희열, 박상훈)은 최근 원자력 발전소(PWR)에서 생성되는 방사성폐액을 국내실정에 맞게 고화처리 할수 있는 새로운 공정과 함께 이를 위한 시험공장(Pilot plant)설계도 완료하여 앞으로 본격화된 원자력 발전소 가동에 크게 대비하게 되었다.

• KSBB Journal
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• v.10 no.1
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• pp.1-8
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• 1995

A lot of water soluble proteins and lipids are released from minced mackerel meat and lost into the washing waste during the leaching process of Kamaboko or surimi manufacture. The removed proteins and lipids are not only an edible things but also a big burden for treating the wastewater. In order to recover the proteins from the effluent and to use as food stuff, the "pH-shifting" treatment, a modified isoelectric point precipitation method, was tried. This method is based on a myogen-aggregation phenomenon, which occurs when a solution of sarcoplasmic proteins is acidified or alkalified beyond the critical pH zone of 2∼3 or 12∼13 respectively and then neutralized. The maximum amount of precipitation was obtained by shifting the pH of the wastewater from original pH to isoelectric point (pH 4) or alkali pH 12 and then changing to neutral pH. The precipitates were easily collected by filteration or centrifuging at 10,000rpm. The oils which were only floating in the washing wastewater are easily recovered by seperating with oil separator after pouring. The recovered proteins were slightly denaturated during this pH shifting precipitation process, while the composition of amino acids was good balance as a food.

• Clean Technology
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• v.14 no.4
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• pp.275-280
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• 2008

In this study, we carry out economic and technical feasibility study for treatment process of waste solvent such as SD waste solution and stripper solution from LCD manufacturing process. Extensive screening work has been done for selecting the best extraction solvent. As a result, it is shown that $CHCl_3$ is the most adequate extraction solvent for separation and recovery of both SD waste solvent and stripper waste. Rigorous simulation study has been carried out for extraction process and distillation process as a candidate of waste solvent treatment. The result shows that the solvent extraction process is more beneficial than the simple distillation process.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2014.11a
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• pp.86-88
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• 2014

니켈도금폐액은 재자원화되지 못하고 대부분 외부로 위탁 폐수처리되고 있으며, 고가의 금속인 니켈을 회수되지 못하고 버려지고 있는 실정이다. 이에 니켈도금폐액으로부터 니켈을 회수하고, 회수된 니켈을 이용하여 니켈화합물을 제조하여 니켈도금폐액을 재자원하는 기술을 개발하였다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.665-670
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• 2003

This study investigated the thermodynamic and the thermal decomposition properties of high concentration nitrate salts waste water for the lagoon sludge treatment. The thermodynamic property was carried out by COACH and GEMINI II based on the composition of nitrate Salts waste water. The thermal decomposition property was carried out by TG-DTA and XRD. Ammonium nitrate and sodium nitrate were decomposed at $250^{\circ}C$$730^{\circ}C$$450^{\circ}C$$Na_2O$ into stable $Na_2O$.$Al_2O_3$. The flow sheet for nitrate salts waste water treatment was proposed based on the these properties data. These will be used by the basic data of the process simulation.

• Applied Chemistry for Engineering
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• v.8 no.1
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• pp.99-107
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• 1997

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, $Na_2U_2O_7$ 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 power preparation, is treated with NaOH to takeup fluorine(F) in the waste because fluorine is valuable and toxic material. In the finally treated waste, uranium was not detected.