To secure approval for a decommissioning plan in Korea, it is essential to evaluate contamination dispersion through groundwater during the decommissioning process. To achieve this, licensees must assess the groundwater characteristics of the facility’s site and subsequently develop a groundwater flow model. It is worth noting that Combustible Radioactive Waste Treatment Facility (CRWTF) is characterized by their simplicity and absence of liquid radioactive waste generation. Given these facility characteristics, the groundwater flow model for CRWTF utilizes data from neighboring facilities, with the feasibility of using reference data substantiated through comparative analysis involving groundwater characteristic testing and on-site modeling. To enable a comparison between the actual site’s groundwater characteristics and the referenced modeling, two types of hydraulic constant characterization tests were conducted. First, hydraulic conductivity was determined through long-term pumping and recovery tests. The ‘Theis’ and ‘Cooper-Jacob’ equations, along with the ‘Theis recovery’ equation, were applied to calculate hydraulic conductivity, and the final result adopted the average of the calculated values. Secondly, a groundwater flow test was conducted to confirm the alignment between the main flow direction of the referenced model and the groundwater flow in the CRWTF, utilizing the particle tracking technique. The evaluation of hydraulic conductivity from the hydraulic constant test revealed that the measured value at the actual site was approximately 1.84 times higher than the modeled value. This variance is considered valid, taking into consideration the modeling’s calibration range and the fact that measurements were taken during a period characterized by wet conditions. Furthermore, a close correspondence was observed between the groundwater flow direction in the reference model (ranging from 90° to 170°) and the facility’s actual flow direction (ranging from 78° to 95°). The results of reference data for the CRWTF, based on the nearby facility’s model, were validated through the hydraulic properties test. Consequently, the modeling data can be employed for the demolition plan of CRWTF. It is also anticipated that these comparative analysis methods will be instrumental in shaping the groundwater investigation plans for facilities with characteristics similar to CRWTF.

KAERI has developed a Radioactive Waste Information Management System (RAWINGS) to manage the life-cycle information from the generation to the disposal of radioactive waste, in compliance with the low- and medium-level radioactive waste acceptance criteria (WAC). In the radioactive waste management process, the preceding steps are to receive waste history from the waste generators. This includes an application for a specified container with a QR label, pre-inspection, and management request. Next, the succeeding steps consist of repackaging, treatment, characterization, and evaluating the suitability of disposal, for a process to transparently manage radioactive wastes. Since the system operated in 2021, The system is enhanced to manage dynamic information, including the tracking of the location of radioactive waste and the repackaging process. Small packages of waste could be classified as either radioactive or clearance waste during pre-inspection. Furthermore, waste generated in the past has already been packaged in drums, and a new algorithm has been developed to apply the repackaging when reclassification is required. All radioactive waste with the unique ID number on the specific container is managed within a database, the total amount and history of waste are managed, and statistical information is provided. This system is continuously be operated and developed to oversee life-cycle information, and serve as the foundational database for the Waste Certification Program (WCP).

In light of recent significant seismic events in Korea and worldwide, there is an urgent need to reevaluate the adequacy of seismic assessments conducted during facility construction. This study reexamines the ongoing viability of the Safety Shutdown Earthquake (SSE) criteria assessment for the Combustible Radioactive Waste Treatment Facility (CRWTF) site at the Korea Atomic Energy Research Institute (KAERI), originally established in 1994. To validate the SSE assessment, we delineated 13 seismic structure zones within the Korean Peninsula and employed two distinct methodologies. Initially, we updated earthquake occurrence data from 1994 to the present year (2023) to assess changes in the site’s horizontal maximum earthquake acceleration (g). Subsequently, we conducted a comparative analysis using the same dataset, contrasting the outcomes derived from the existing distance attenuation equation with those from the most recent attenuation equations to evaluate the reliability of the applied attenuation model. The Safety Shutdown Earthquake (SSE) criterion of 0.2 g remains unexceeded, even when considering recent earthquake events since the original evaluation in 1994. Furthermore, when applying various assessment equations developed subsequently, the maximum value obtained from the previously utilized ‘Donvan and Bornstein’ attenuation equation is 0.1496 g, closely resembling the outcome derived from the recently employed ‘Lee’ reduction equation of 0.1451 g. The SSE criteria for CRWTF remain valid in the current context, even in light of recent seismic occurrences such as the 2016 Gyeongju earthquake. Additionally, the attenuation equation employed in the evaluation consistently yields conservative results when compared to methodologies used in recent assessments. Consequently, the existing SSE criteria remain valid at present. This study is expected to serve as a valuable reference for confirming the SSE criterion assessment of similarly constructed facilities within KAERI.

Spent ion exchange resins have been generated during the operation of nuclear facilities. These resins include radioactive nuclides. It is needed to fabricate them into a stable form for final disposal. Cement solidification process is a useful method for the fabrication of them into a waste form for final disposal. In this study, proper conditions for the fabrication of them into a stable waste form were determined using the cement solidification process. In-drum waste forms were then produced at the conditions, where the stability of representative samples was evaluated for final disposal. The samples were satisfied to the Waste Acceptance Criteria for low and intermediate level radioactive waste disposal sites. This result can be utilized to derive optimal conditions for the fabrication of spent ion exchange resins into a final disposal form.

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.

In a nuclear power plant, the activated corrosion products are deposited on the reactor coolant system. The activated corrosion products must be removed to reduce the radiation exposure to workers before maintaining or decommissioning of the nuclear power plant. In order to remove the remove the activated duplex oxide layer generated on the reactor coolant system in the pressurized water reactor (PWR), the Cyclic SP (Sulfuric acid/Permanganate)-HyBRID (Hydrazine Based Reductive metal Ion Decontamination) process developed by KAERI (Korea Atomic Energy Research Institute) can be used. After applying the Cyclic SP-HyBRID process, a sulfate-rich waste powder containing the radionuclide is generated, and the radioactive powder has to be stabilized for final disposal. In the previous study, it was confirmed that the low-temperature sintering method can be applied to immobilize the sulfate-rich waste powder. Thus, immobilization of the Cyclic SP-HyBRID process waste powder was carried out by the low-temperature sintering method using a low melting point glass, and the physicochemical and radiological characteristics of a waste form were evaluated in this study. As a result, the compressive strength of the waste form increased with increasing sintering temperature and sintering time. It is considered that the result was caused by the difference in the band gap between the bismuth borate and zinc borate, which are the products during the sintering process. It was verified that the physical stability was maintained after the 107 Gy of irradiation test. In addition, it was confirmed that the radioactive metal hydroxides contained in the waste powder converted to metal oxide forms, which have the lower solubility, at the sintering temperature. Finally, the waste form was evaluated as a low-level radioactive waste from the concentration of radionuclides contained in the waste form.

Strong acidic wastewater containing a radionuclide is generated from the decontamination of radioactively contaminated wastes or equipment. This wastewater is generally treated though a precipitation process using an alkali (alkali earth) hydroxides. In this precipitation process, a significant amount of alkali (alkali earth) sulfates are generated, which is the reason for the increase in the radioactive waste generation. In this study, a method for separating only radionuclides and metal ions from the wastewater was evaluated. For this reason, precipitation behaviors of radionuclides and metal ions by hydrazine injections were investigated using equilibrium calculations. In addition, behaviors of hydrazine decomposition after removal of radionuclides and metal ions were analyzed for recycling the wastewater.

Radioactively contaminated metal components from a nuclear power plant must be decontaminated to reduce the risk of radiation exposure to workers, which can be cleaned using a foam decontamination used to reduce the amount of wastewater significantly. Metal components with a fixed radioactive contamination can be effectively decontaminated using a foam consist of 0.5wt% nonionic surfactant, 0.5 M H2SO4, and 0.2 M Ce(SO4)2. However, strongly acidic wastewater is generated from the decontamination method, which contains a high concentration of the nonionic surfactant and ionic materials with radioactive nuclides. This wastewater must be treated as a stable form. In this study, an integrated process of precipitation and low pressure distillation was evaluated for the treatment of wastewater. It was confirmed that the surfactant and ionic materials were effectively removed from the wastewater through the integrated process.

In a nuclear facility, the base metal can be radiologically contaminated during the operation. They must be decontaminated to reduce the radiation exposure to workers before decommissioning of the nuclear facility. In order to decontaminate the nuclear facility, it is possible to apply a perfluorocarbon (PFC) based emulsion consisted of surfactant and decontamination reagent. The PFC has high resistance for the radiation decomposition, and PFC based emulsion can be easily stabilized using the ultrasonication method. During decontamination process, a dispersion stability of the emulsion affects to the decontamination performance because the decontamination reagents dispersed in the emulsion contact contaminated surface. In this study, the dispersion stability the PFC based emulsion was evaluated following the composition of the emulsion and dispersion condition such as temperature, ultrasonication time. It was confirmed that the concentration of surfactant is highly related to the dispersion stability from the result of Turbiscan analysis using the multiple light scattering method. It was also verified that the droplet size of the decontamination reagent in the stable emulsion was smaller than that in the unstable emulsion. This phenomena can be explained by the relationship between the interfacial tension and droplet size. Finally, the recovering test of the PFC from the spent PFC-based decontamination emulsion was conducted using distillation method. The distillation test was performed using vacuum distillation unit, and the distillation temperature was 80°C. From the distillation test, about 95 % of PFC was recovered by distillation. From this result, it is considered that PFC-based decontamination emulsion reduces the volume of the secondary waste.

Sulfate-rich waste powder containing a radioactive nuclide is generated from chemical decontamination process and radioactive liquid waste treatment using ion exchange resin. The radioactive sulfate-rich waste powder should be stabilized for final disposal. The techniques for immobilization of the radioactive sulfate-rich waste powder such as hydraulic cement, geopolymer, and iron phosphate glass have been applied, however, there are limitation in these techniques. Firstly, the hydraulic cement cannot applied to the wastes containing high concentration of sulfate because the expansion, cracks, and disintegration can be happened in the waste form. Geopolymer has a low density although they can be used as a good binder. The iron phosphate glass can be utilized, however, a considerable amount of SO2 gas is emitted due to the high sintering temperature. In this study, immobilization of radioactive sulfate-rich waste powder was carried out to resolve above problems by applying low temperature sintering method using a low-melting glass. As a result, it was confirmed that the waste form has a high bulk density. The compressive strength of the waste form was over 40 MPa, which is higher than the acceptance criteria (≥ 3.44 MPa). From ANS 16.1 test, it was verified that the waste form met the acceptance criteria of the leachability index (≥ 6). It was also confirmed that the waste form was chemically durable through product consistency test (PCT). In addition, the chemical stabilities of waste forms were compared following the sintering condition and the composition of the waste forms. The difference of the chemical stability was explained by difference in the abundance of chemical form obtained from the sequential extraction test.

Activated corrosion products deposited on the reactor coolant system in a nuclear power plant should be removed to reduce the radiation exposure to workers. Chemical decontamination processes using organic acids have been widely applied to remove the activated corrosion products. However, they are highly corrosive to the base metal and generate a considerable amount of ion exchange resin waste, which is hard to be treated. In order to resolve this problem, KAERI has been developed a chemical decontamination process using chelate-free inorganic acid, HyBRID (Hydrazine Based Reductive metal Ion Decontamination) process. Especially, the Cyclic SP (Sulfuric acid/Permanganate)- HyBRID process was suggested as the decontamination process for applying to the remove the double oxide layer generated on the reactor coolant system in the pressurized water reactor (PWR). During the Cyclic SP-HyBRID process, the process is continuously applied without discharging or recharging of the decontamination process solution from the primary circuit. Thus, it is necessary to include the removal processes of the decontamination reagents middle of the Cyclic SP-HyBRID process, e.g., ‘Mn removal step’ for removing the permanganate ions and ‘hydrazine decomposition step’ for decomposition of the remaining hydrazine. During these removal processes, the metal ions can also be removed from the process solution. In this study, the behaviors of metals were investigated during the Cyclic SP-HyBRID process. The concentration changes of metal ions in the process solution were analyzed using atomic absorption (AA) spectroscopy. The metal precipitates generated during the process were characterized using X-ray diffraction (XRD) and Fourier Transform Infrared (FT-IR) spectroscopy. From the results of the analysis, it was observed that the metal ions dissolved in the process solution were converted into metal hydroxides and precipitated at the Mn removal process. It was confirmed by equilibrium calculation result that the OH− ions generated at the Mn removal can react with the metal ions and form the metal hydroxides. It is considered that this removal behaviors of the metals can contribute the decontamination performance.

원전 일차계통 HyBRID 제염공정에서 발생되는 제염폐액에는 황산이온과 방사성 핵종을 포함한 금속이온 및 발암성 물질의 하이드라진을 포함하고 있어 이를 안전한 수준으로 처리할 수 있는 기술개발이 필요하다. 본 연구에서는 모의 제염폐액 내 황산 및 금속이온의 제거와 하이드라진의 분해시험을 실시하여 황산이온, 금속이온 및 하이드라진을 효과적으로 제거할 수 있는 HyBRID 제염폐액 처리공정을 도출하였으며, 1 L 규모에서의 반복실험과 Pilot 규모(300 L/batch)에서의 평가시험을 통해 도출한 HyBRID 제염폐액 처리공정의 성능 재현성과 적용성을 검증하였다.

사용후핵연료을 건식처리하는 파이로프로세싱 중 전해정련 및 제련공정 후 발생되는 우라늄과 초우라늄 및 희토류 등의 염화물을 함유한 LiCl-KCl 공융염에는 특히 희토류 함량이 높기 때문에 유효자원으로 활용이 가능한 형태의 우라늄과 초우라 늄의 분리/회수가 쉽지 않다. 이러한 문제점을 해결하기 위해 본 연구에서는 LiCl-KCl-UCl3-NdCl3 시스템에서 산화제(K2CO3) 를 이용하여 UCl3를 산화물 형태로 전환한 후 전기화학적 방법을 이용하여 NdCl3를 금속형태로 분리하는 실험을 실시하였 다. 실험에 앞서, 이론적 평형계산을 수행하여 우라늄 염화물을 산화물로 전환하기 위한 실험조건을 결정하였다. 상기의 실 험에서 LiCl-KCl 내 UCl3는 첨가제의 주입량이 이론적 반응당량에 근접하였을 때 거의 대부분이 염내에서 염화물 형태로 존 재하지 않는 것으로 나타났다. 이후 액체금속음극을 이용하여 NdCl3를 금속형태로 전착시켰으며, 전착실험 후 투명한 용융 상의 LiCl-KCl 공융염과 갈색의 우라늄 산화침전물이 존재함이 확인되었다. 이러한 결과들을 통해 LiCl-KCl-UCl3-NdCl3 시스 템에서 우라늄 및 희토류를 각각 분리할 수 있는 방안을 수립할 수 있을 것으로 판단된다.

KAERI에서는 파이로프로세싱에서 발생하는 금속폐기물의 부피 및 무게 감량을 위해 고방사성 장반감기 핵종을 포함하는 anode sludge내 NM의 고화매질로써 폐피복관과 첨가금속을 재활용하는 연구를 진행하고 있다. 본 연구에서는 Cr 함량을 조절한 Zr-17Cr-8NM, Zr-22Cr-8NM, Zr-27Cr-8NM 합금을 유도용융을 통해 제조하였고, 전기화학적 부식시험을 실시하여 부 식특성을 평가하였다. 모든 조성에서 기존 연구 중인 Zr계 합금고화체 조성보다 우수한 부식특성을 나타냈다. 또한 Zr-22Cr- 8NM 시편의 부식시험 후 침출용액 조성 분석 결과, 500 mV 전압 조건 이하에서는 NM 침출이 없었고 이를 통해 우수한 화 학적 안정성을 갖는 합금고화체 조성을 확보하였다.