This study presents a rapid and sequential radiochemical separation method for Pu and Am isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin and TRU resin. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the sample was allowed to evaporate to dryness after filtering the leaching solution with 0.45 micron filter. The Pu isotopes were separated in HNO3 medium with anion exchange resin. For leaching solution passed through anion exchange column, the Am isotopes were separated with TRU resin. The purified Pu and Am isotopes were measured by alpha spectrometer, respectively, after micro-precipitation of neodymium. The sequential radiochemical separation of Pu and Am isotopes in radioactive waste samples using anion exchange resin and TRU resin was validated with ICP-MS system.

We conducted safety assessments for the disposal of spent resin mixed waste after the removal of beta radionuclides (3H, 14C) in a landfill facility. The spent resin tank of Wolsong nuclear power plant is generated by 8:1:1 weight ratio of spent ion exchange resin, spent activated carbon and zeolite. Waste in the spent resin tank was classified as intermediate-level radioactive waste due to 14C. Other nuclides such as 60Co and 137Cs exhibit below the low-level radioactive waste criteria. The techniques for separating mixed waste and capturing 14C have been under development, with a particular focus on microwave-based methods to remove beta radionuclides (3H, 14C) from spent activated carbon and spent resin within the mixed waste. The spent resin and activated carbon within the waste mixture exhibits microwave reactivity, heated when exposed to microwaves. This technology serves as a means to remove beta isotopes within the spent resin, particularly by eliminating 14C, allowing it to meet the low-level radioactive waste criteria. Using this method, the waste mixture can meet disposal requirements through free water and 3H removal. These assessments considered the human intrusion scenarios and were carried out using the RESRAD-ONSITE code. The institutional management period after facility closure is set at 300 years, during which accidental exposures resulting from human intrusion into the disposal site are accounted for. The assessment of radiation exposure to intruders in a landfill facility included six human intrusion scenarios, such as the drilling scenario, road construction scenario, post-drilling scenario, and post-construction scenario. Among the six human intrusion scenarios considered, the most conservative assessment about annual radiation exposure was the post-drilling scenario. In this scenario, human intrusion occurs, followed by drilling and residence on the site after the institutional management period. We assumed that some of the vegetables and fruits grown in the area may originate from contaminated regions. Importantly, we confirmed that radiation doses resulting from post-institutional management period human intrusion scenarios remain below 0.1 mSv/y, thus complying with the annual dose limits for the public. This research underscores the importance of effectively managing and securing radioactive waste, with a specific focus on the safety of beta radionuclide-removed waste during long-term disposal, even in the face of potential human intrusion scenarios beyond the institutional management period.

The intermediate level spent resins waste generated from water purification for the the moderator and primary heat transport system during operaioin of heavy water reactor (HWR). Especially, moderator resins contain high level activity largely because of their C-14 content. So spent resins are considered as a problematirc solid waste and require special treatment to meet the waste acceptance criteria for a disposal site. Various methods have been studied for the treatment of spent resins which include thermal, destructive, and stripping methods. In the case of solidification methods, cement, bitument or organic polymers were suggested. In the 1990s, acid stripping using nitric acid and thermal treatment methods were actively investigated in Canada to remove C-14 nuclide from waste resin. In Japan, thermal distructive method was studied in the 1990s. Since 2005, KAERI developed acid stripping method using phosphate salt. However, acid stripping method are not suitable due to large amounts of 2nd waste containing acid solution with various nuclides. To solve this probelm, KAERI has been suggested the microwave treatment method for C-14 selective removal from waste resin in the 2010s. Pilot scale demonstration tests using radioactive waste resin generated from Wolsung unit 1 and unit 2 were successfully conducted and 95% of C-14 was selectively removed from the radioactive waste resin. In recent years, price of C-14 source is dramatically increased due to market growth of C-14 utilization and exclusive supply chain depending on China and Russia. High purity of C-14 were captured in HWR waste resin. Interest of C-14 recovery research from HWR waste resin is currently increased in Canada. In this study, microwave method is suggested to treat HWR waste resin with C-14 recovery process. Additionally, status of waste resin management and research trends of HWR waste resin treatment are introduced.

This study presents a rapid and quantitative radiochemical separation method for Nb isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin after Fe coprecipitation. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the Nb isotopes were coprecipitated with Fe after filtering the leaching solution with 0.45 micron HA filter, while the Sr, Tc and Ni isotopes were in the solution. The Nb isotopes were separated in HCl medium with anion exchange resin. The purified Nb isotopes were measured using a low level liquid scintillation counter after installing quenching curve with standard Nb-94 isotopes. The separation method for Nb isotopes investigated in this study was applied to neutron dosimeter samples from the nuclear power plant after validating the Nb activity concentration with gamma spectrometry system.

The liquid radioactive waste system of nuclear power plants treats radioactive contaminated wastes generated during the Anticipated Operational Occurrence (AOO) and normal operation using filters, ion exchange resins, centrifuges, etc. When the contaminated waste liquid is transferred to an ion exchanger filled with cation exchange resin and anion exchange resin, nuclides such as Co and Cs are removed and purified. The lifespan and replacement time of the ion exchange resin are determined by performing a performance test on the sample collected from the rear end of the ion exchanger, and waste ion exchange resin is periodically generated in nuclear power plants. In the general industry, most waste resins at the end of their lifespan are incinerated in accordance with related laws, but waste resins generated from nuclear power plants are disposed of by clearance or stored in a HIC (High Integrity Container). Plasma torch melting technology can reduce the volume of waste by using high-temperature heat (about 1,600 degrees) generated from the torch due to an electric arc phenomenon such as lightning, and secure stability suitable for disposal. Plasma torch melting technology will be used to check thermal decomposition, melting, exhaust gas characteristics, and volume reduction at high temperatures, and to ensure disposal safety. Through this research, it is expected that the stable treatment and disposal of waste resins generated from nuclear power plants will be possible.

Inorganic and organic ion exchange materials were generally applied to liquid processes in nuclear reactor. In the case of heavy-water reactor (HWR), zeolite, active carbon, anion resin, and cation resin were used to treat liquid processes such as reactor primary coolant cleanup and liquid radioactive waste management system. Then, used ion exchangers were stored at storage tanks. Various kinds of nuclides were adsorbed in ion exchange materials. Especially, C-14, long half-life nuclide, was highly concentrated in anion resin, and waste resin was treated as intermediated level radioactive waste (ILW). Thermal and non-thermal methods such as pyrolysis, incineration, catalytic extraction, acid digestion, and wet oxidation have been studied for treating spent resin. However, destructive methods are not suitable due to massive off gas waste containing radioactive species. To solve this problem, various kinds of processes were developed such as acid stripping, PLO process, activity stripping, thermal treatment, and etc. In this study, microwave method is suggested to treat HWR waste resin. C-14 nuclide was selectively removed from waste resin without decomposition of main structure in waste resin. Radioactive waste resin generated from Wolsung HWR unit 1 and unit 2 was treated using microwave method and 95% of C-14 was successfully removed from the radioactive waste resin.

Abstract Recently, the circular economy aiming at elimination of waste and the continual use of resources has attracted a lot of attentions. In the circular system, the resource recovery uses the recycled wastes as the raw material to manufacture new valuable products. This work focuses on a low-cost process, in which an activated carbon (AC) adsorbent was prepared from waste cation exchange resin by calcination and HNO3 activation hydrothermal method. Surface structure and chemistry of AC were characterized by SEM, XRD, FTIR and Boehm titration. It was found that the acid treatment could increase the number of pores and the content of oxygen-containing functional groups on AC surface. In the adsorption experiment, Methylene blue (MB) was used to assess the adsorption capacity of AC. Experimental results showed that the highest efficiency of MB removal was achieved by AC with modification of 4M HNO3, showing the acidification effect on the adsorption capacity of AC. Adsorption isotherms of Langmuir and Freundlich were employed to fit the experimental data. It was shown that MB adsorption on AC is more consistent with Langmuir model that assumes a homogeneous adsorption. In the adsorption kinetic analysis, the adsorption was found to follow the pseudo-second-order model, indicating that adsorption of MB on acidified AC is dominated by chemical adsorption. The study revealed that the waste ion-exchange resin is a proper precursor of carbon adsorbent for MB dye. This low-cost method would specifically reduce the environmental cost of waste disposal.

Cs 이온에 대해 선택성을 갖는 ferrocyanide-음이온 교환수지를 제조하여 모의 제 염폐액 내에 존재하는 Cs 이온에 대한 흡착실험을 수행하였다. 제조된 이온교환 수지가 citric acid를 주제염제로 하는 제염폐액 내에 존재하는 Cs+ 이온에 대한 흡착능력은 상용 양이온교환수지에 비해 4배 이상 효과적인 것으로 나타났다. 모의 제염폐액과 선택성 이온교환수지를 접촉시킨 후 360분이 경과하면 금속이온에 대한 흡착반웅이 평형에 도달하였다. 본 연구범위에서 Co 이온농도가 필요이상 증가하게 되면 Cs 이온의 흡착율은 감소하였다. 과산화수소와 히드라진을 사용한 선택성 폐 이온교환수지의 재생실험 결과 전기중성화조건을 만족시키기 위해 Cs 이온이 수지로부터 용출됨을 확인하였고 열화없이 재 사용가능성을 확인하였다.

반응성 페놀수지 폐액을 처리하기 위해 중공사막 모듈을 이용한 투과증발 막 탈수공정을 연구하였다. 이 공정의 거동을 예측하기 위한 모사모델을 확립하였고 여기에 사용되는 중요 기본 파라메타들을 평판형 막을 사용하여 직접 구하여 사용함으로써 공정모사의 정확성을 얻을 수가 있었다. 이들을 모사치와 중공사 투과증발 막으로 부터 직접 측정한 각 투과특성들을 비교한 결과 서로 잘 일치함을 보여 본 모사모델의 타당성을 입증하였다. 사용된 중공사막은 중공사 안쪽에 활성층이 도포되어 있으며 공급액은 중공사 내부로 공급하였다. 공급액의 막내에서의 흐름속도에 따라 온도분포가 결정되며 이에 따라 막 투과특성이 달라짐을 모사결과로부터 얻을 수가 있었다. 공급액 온도증가는 막을 통한 탈수 투과 속도를 증가시킬 뿐 아니라 반응속도 증가로 인하여 물 생성속도도 증가시킴으로써 공급액 저장조 내의 수분 함량은 이들 상반된 공정들에 의해 결정이 됨을 보였다. 투과압력이 공급액 증기압보다 훨씬 작은 범위에서 증가할 경우 투과추진력인 공급액과 투과부의 투과물 활성도비 감소가 크지 않아 투과특성을 약간 저하시킨다. 그러나 투과압력이 공급액의 증기압에 접근할 경우 활성도비 감소가 현저하게 일어나 투과특성저하가 급격히 일어난다.

To prevent environmental pollution caused by leakage of leachate from waste landfill, vinyl acetate-ethylene (VAE) resin is applied to liner and cover materials to improve their performance. Styrene, styrene butadiene rubber, and VAE are widely used as polymer resins that have excellent water resistance and durability. Further, VAE resin is known to have additional advantages such as adhesion to nonpolar materials and resistance to saponification as a copolymer. In this study, the effect of VAE content on the properties of liner and cover materials was studied. The water and air content ratios, bending and compressive strengths, water absorption ratio, and coefficient of permeability of these materials were measured. The liner and cover materials with 4 wt% VAE showed good properties.