The process of carbonization followed by a high-temperature halogenation removal of radionuclides is a promising approach to convert low-radioactivity spent ion-exchange (IE) resins into freereleasable non-radioactive waste. The first step of this process is to convert spent ion-exchange resins into the carbon granules that are stable under high-temperature and corrosive-gas flowing conditions. This study investigated the kinetics of carbonization of cation exchange resin (CER) and the changes in structures during the course of carbonization to 1,273 K. Both of model-free and modelfitted kinetic analysis of mixed reactions occurring during the course of carbonization were first conducted based on the non-isothermal TGAs and TGA-FTIR analysis of CER to 1,272 K. The structural changes during the course of carbonization were investigated using the high-resolution FTIR and C-13 NMR of CER samples pyrolyzed to the peak temperature of each reaction steps established by the kinetic analysis. Four individual reaction steps were identified during the course of carbonization to 1,273 K. The first and the third steps were identified as the dehydration and the dissociation of the functional group of —SO3-H+ into SO2 and H2O, respectively. The second and the fourth steps were identified as the cleavage of styrene divinyl benzene copolymer and carbonization of pyrolysis product after the cleavage, respectively. The temperature and time positions of the peaks in the DTG plot are nearly identical to those of the peaks of the Gram Schmidt intensity of FTIR. The structural changes in carbonization identified by high-resolution FTIR and DTG are in agreement with those by C-13 NMR. The results of a detailed examination of the structural changes according to NMR and FTIR were in agreement with the pyrolysis gas evolution characteristics as examined by TGA-FTIR.

Nuclear power plants in Korea stores approximately 3,800 drums of paraffin solidification products. Due to the lack of homogeneity, these solidification products are not allowed to be disposed of. There is therefore a need for the separation of paraffin from the solidification products. This work developed an equipment for a selective separation of paraffin from the solidification product using the vacuum evaporation and condensational recovery method in a closed system. The equipment mainly consists of a vacuum evaporator and a condensational deposition recovery chamber. Nonisothermal vacuum TGAs, kinetic analyses and kinetic predictions were conducted to set appropriate operation conditions. Its basic operability under the established conditions was first confirmed using pure paraffin solid. Simulated paraffin solidification product fixing dried boric acid waste including nonradioactive Co and Cs were then fabricated and tested for the capability of selective separation of paraffin from the simulated waste. Paraffin was selectively separated without entertainment of Co and Cs. It was confirmed that the developed equipment could separate and recover paraffin in the form of nonradioactive waste.

Dry active wastes (DAWs) are a type of combustible radioactive solid waste, which includes decontamination paper, protective clothing, filters, plastic bags, etc. generated from operating nuclear facilities and decommissioning projects. The volume of DAWs could be increased over time, disadvantage to higher disposal costs and space utilitization of disposal site. Additionally, incineration methods cannot be applied to DAWs, unlike general environmental waste, due to concerns about air pollution and the release of harmful chemicals with radioactive nuclides into the atmosphere. Recently, KAERI developed an alternative thermochemical process for reducing the volume of DAW, which involves a step-wise approach, including carbonization, chlorination, and solidification. The purpose of this process is to selectively separate the radioactive nuclides from carbonized DAWs that are less than clearance criteria, which can be disposed of as non-radioactive waste. In this research, we investigated the thermal decomposition characteristics of DAWs using nonisothermal thermogravimetric analysis, which was performed with different categorized wastes and heating conditions. As a result, the cellulose DAWs such as decontamination paper and cotton were thermally decomposed in three or four-step depending on the heating conditions. On the other hand, the hydrocarbon and rubber DAWs such as plastic bags and latex were thermally decomposed in one or two-step. Therefore, it could be suggested the thermochemical treatment conditions that minimize the decomposition of DAWs by controlling the reaction steps, and we will try to apply these results for cellulose type DAWs such as decontamination paper and cotton, which is generated majorly from the nuclear facilities in the future.

Radioactive mixed waste (RMW) is containing radioactive materials and hazardous materials. Radioactive wastes containing asbestos are include in RMW. These wastes thus must be treated considering both radioactive and hazardous aspects. In this study, a high temperature melt oxidation system consisting of an electric arc furnace and a molten salt oxidation furnace has been developed for the treatment of of radioactive waste containing asbestos. A surrogate waste of the radioactive waste containing asbestos (content of asbestos: 13wt%) was treated in this system. It was melted and fabricated into a glass waste form in the system. Asbestos was not detected in this glass waste form. This means that the asbestos was converted to a glass component in the glass waste form. The waste form was homogeneous glass, and it had a high value of compressive strength (475.13 MPa). It was also confirmed through a leaching test (ANS 16.1) that the waste form had a high chemical durability (Leaching Index > 6). Based on these results, it is considered that the high temperature melt oxidation system will be utilized for the treatment of a significant amount of radioactive waste containing asbestos generated from decommissioning a nuclear power plant.

The Korea government decided to shut down Kori-1 and Wolsung-1 nuclear power plants (NPPs) in 2017 and 2019, respectively, and their decommissioning plans are underway. Decommissioning of a NPP generates various types of radioactive wastes such as concrete, metal, liquid, plastic, paper, and clothe. Among the various radioactive wastes, we focused on radioactive-combustible waste due to its large amount (10,000–40,000 drums/NPP) and environmental issues. Incineration has been the traditional way to minimize volume of combustible waste, however, it is no longer available for this amount of waste. Accordingly, an alternative technique is required which can accomplish both high volume reduction and low emission of carbon dioxide. Recently, KAERI proposed a new decontamination process for volume reduction of radioactivecombustible waste generated during operation and decommissioning of NPPs. This thermochemical process operates via serial steps of carbonization-chlorination-solidification. The key function of the thermochemical decontamination process is to selectively recover and solidify radioactive metals so that radioactivity of the decontaminated carbon meets the release criteria. In this work, a preliminary version of mass flow diagram of the thermochemical decontamination process was established for representative wastes. Mass balance of each step was calculated based on physical and chemical properties of each constituent atoms. The mass flow diagram provides a platform to organize experimental results leading to key information of the process such as the final decontamination factor and radioactivity of each product.

Membrane-based gas separations are becoming relevant. Ordered mesoporous silicas have uniform pore channels with a mesopore range. The mesopores are advantageous for rapid diffusion of target molecules, can be modified in a variety of ways, allowing for separation applications. We first describe the technologically scalable fabrication based upon mesoporous silica membranes on polymeric hollow fibers. Furthermore, we modify the mesopores using amine-containing polysilsesquioxane molecules. The resulting modified membranes exhibit promising permeability and selectivity in separation of CO2 from N2 or CH4. Also, the CO2 and N2 transport properties of aziridine-functionalized mesoporous silica membranes are investigated. The hyperbranched aminosilica membrane shows bi-functional gas transport property regarding humidity of gas streams.

Radioactive liquid waste has given soil & water pollution and further congenital defect to the humanity. With this regard, there are lots of interests in reducing concentration of radioactive material even if it is in low level (ppm). There are filtration driven by pressure such as ultra-filtration, nano-filtration, and reverse osmosis to separate radioactive liquid waste. Among them, Nano-Filtration membrane shows promising performance in both permeability and selectivity. Commercial nanofiltration membrane were applied to separate model solutions. Model seawater was desalinated to 6,100 ppm with flux of 90 L/ ㎡h. On the other hands, Model nuclides (Co, Cs, and Sr) solutions are separated from 10 ppm to 2 ppm, 0.23 ppm, and 0.13 ppm. Commercial nanofiltration membranes show the trade-off property between selectivity and flux.

사용후핵연료 건식처리공정(pyrochemical process)에서 LiCl-KCl 공융염의 회수는 방사성폐기물 부피 감량과 원료물질 회수를 위해 반드시 필요하다. 본 논문은 진공증류공정을 이용하여 희토류 침전물(희토 류 산염화물 또는 산화물)내 잔류하는 LiCl-KCl 공융염 회수에 관한 것이다. 진공증류시험장치에서 희토 류 침전물내 공융염은 효과적으로 휘발 및 분리되었다. 분리된 공융염은 감압증류시험장내 세 지점에서 침적되거나 필터에 포집되으며, 침적되거나 포집된 공융염을 회수하는 것은 쉽지 않았다. 이 문제점을 해 결하기 위해 감압조건에서 온도구배를 이용하여 공융염 거동을 제어할 수 있는 공융염 진공증류/응축회 수 시스템을 개발하였으며, 이 장치를 이용하여 휘발된 공융염을 회수용기에서만 응축시켜 쉽게 회수할수 있음을 확인하였다.

비등온 및 등온조건에서의 열중량분석을 이용하여 다양한 압력조건(0.5 - 50 mmHg)에서 LiCl-KCl 공융염 증류속도를 우선 측정하였다. 비등온조건에서의 열중량분석결과로부터 온도의 함수로 표현될 수 있는 증류 속도식을 도출하였다. 이 속도식에 의해 계산된 휘발플럭스(flux)는 등온조건에서의 열중량분석을 통해 얻어 진 증류속도와 일치하였다. 1300 K 이하의 온도조건과 0.5 mmHg와 50 mmHg 사이의 감압조건에서 10-4- 10-5 mole cm-2 sec-1의 증류속도를 얻을 수 있다. 실험실규모 실험장치에서 50 mmHg의 압력과 1150 K 이상 의 온도 조건에서 한 시간 증류로 약 99%의 염이 분리되었다. 희토류 침전물내에 잔류하는 염을 증류에 의해 제거할 때 휘발시간이나, 온도를 증가시키는 것보다 휘발 계면적을 증가시키는 것이 효과가 더 큰 것으로 나타났으며, 휘발면적을 4.52 cm2에서 12.56 cm2로 증가시켜 한 시간 동안 증류하였을 때 99.95% 이상의 염이 분리되었다.

본 연구에서는 산화조건하 LiCl-KCl 공융염내에서 란탄계 염화물의 하나인 PrCl3의 열적거동을 살펴보았다. 먼저 산소를 주입하면서 PrCl3의 열중량분석(TGA; thermogravimetric analysis)을 실시하였고, 이 때 얻어진 결과들을 바탕으로, 산소분산법을 이용하여 온도에 따른 LiCl-KCl 공융염내 PrCl3의 산화실험을 수행하였다. PrCl3의 열중량분석 결과에 따르면, 약 380 ℃까지 PrCl3에서 염소의 해리가 급격하게 발생되었고 약 600 ℃에 서 PrCl3가 PrOCl로 전환되는 반응이 종료되는 것으로 확인되었다. 산소분산법에 의한 LiCl-KCl 공융염내 PrCl3의 열적거동은 산화조건에서 열중량분석시 나타난 PrCl3의 열적거동과 유사하였고, 발생된 PrOCl은 공 융염내에서 불용성 화합물로써 바닥으로 침전하였다. 산소분산법에 의한 공융염내 PrCl3의 PrOCl로의 전환 은 650 ℃ 이상의 온도에서 활발하게 진행되었고, 이 때 발생되는 배기가스내 Cl2의 농도분석을 통해 공융염내 PrCl3의 전환상태를 예측할 수 있을 것으로 판단된다