본 연구에서는 막여과공정, 오존접촉공정, 활성탄흡착공정, 자외선소독공정 등이 계층형으로 배치된 정수처리 실증시설의 설계 및 운영을 통해 수직형 정수 처리시설의 최적 설계/운영방안을 도출하고자 하였다.

The predictive capacity of wastewater treatment facility in the industrial park was estimated by the traditional method and on-the-spot survey such as certification of wastewater report and the invoices of water supply and ground water supply. The ratios of a converted wastewater to supplied industrial water between traditional method and on-the-spot survey in the estimation methods were different. By using traditional method, the business type of clothes, accessary and fur production had 77.18 % of waste ratio of wastewater and 10.72 m3/day·1000 m2 unit mass of wastewater as the highest among 9 business types. With the respect to the on-the-spot survey, food manufacturing business type had 75 % of waste ratio of wastewater and 8.35 m3/day·1000 m2 unit mass of wastewater as the highest values. The amount of wastewater from on-the-spot survey method was 541 m3/day less than one from traditional method.

본 연구는 축산분뇨 처리장(바이오가스 플랜트) 액비의 고도처리에 적합한 공정을 도출하기 위한 기초 연구이다. 액비를 고도처리하기 위하여 나노여과 및 역삼투(reverse osmosis) 공정을 각각 사용하였고, 전처리공정으로서 담체를 첨가하지 않은 MBR과 담체를 첨가한 MBR을 각각 적용하여 비교하였다. 액비의 질소는 주로 암모니아성 질소의 형태로 존재하였다. MBR의 운전에서 담체(biomedia) 유무에 따른 COD, T-N 및 T-P의 제거효율에서 큰 차이는 없었으나, 담체를 첨가한MBR의 TMP는 담체를 첨가하지 않은 MBR에 비해 매우 서서히 증가하였다. 전처리 공정으로 담체를 첨가한 MBR 공정을사용한 경우, NF에 의한 COD, T-N 및 T-P의 제거효율은 각각 99.8, 86.5% 및 99.8%이었으며, RO에 의한 제거효율은 각각99.9, 86.8% 및 99.9%이었다. MBR과 NF/RO 공정을 이용하여 처리한 액비의 최종 수질은 분뇨처리장 방류수 수질기준과비교하였을 때, COD와 T-P는 방류수 수질기준을 만족하였으나, T-N은 수질기준에 부적합하였다. 따라서 T-N에 대한 방류수 기준을 만족시키기 위해서는 MBR 조업 cycle의 조정 또는 나노여과/역삼투에 의한 2차 재처리 등의 개선이 필요한 것으로 판단된다.

본 연구에서는 공공하수처리시설의 효율적인 관리방안 제시를 위하여 2008년~2010년까지 28개소의 공공하수처리시설에 대하여 복합악취와 지정악취물질(22종)을 대상으로 악취실태조사 및 원인분석을 실시하였다. 조사결과 전처리 공정과 슬러지 처리공정에서 주로 고농도의 악취가 발생되고 있었으며, 황화수소와 메틸머캅탄 등의 황화합물류가 주요 악취원인물질로 조사되었다. 공공하수처리시설에서 발생되는 악취는 유입수의 성상에 따라 차이가 있으며, 유입수에서의 복합악취는 67배~66,943배, 황화수소는 ND~66.87 ppm으로 조사되었다. A 하수처리시설 유량 조정조에서의 복합악취와 황화수소는 교반시 각각 3,000배, 6.23 ppm, 비교반시 각각 300배, 0.20 ppm으로 조사되었다. 유입 분배조와 생슬러지 분배조는 하수와 슬러지 이송 파이프 라인의 낙차에 의해 내부에 양(+)압이 형성되므로 파이프 라인의 연장과 악취포집설비를 정상적으로 설치․운영하여 내부를 음(-)압 상태로 유지할 필요가 있다.

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.

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.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

Pressurized Heavy Water Reactors (PHWR) have stored ion exchange resins, which are used in deuteration, dehydrogenation systems, liquid waste treatment systems, and heavy water cleaning systems, in spent resin storage tanks. The C-14 radioactivity concentration of PHWR spent resin currently stored at the Wolseong Nuclear Power Plant is 4.6×10E+6 Bq/g, which exceeds the limited concentration of low-level radioactive waste. In addition, when all is disposed of, the total radioactivity of C-14, 1.48×10E+15 Bq, exceeds the disposal limit of the first-stage disposal facility, 3.04×10E+14. Therefore, it is currently impossible to dispose of them in Gyeongju intermediate- and low-level disposal facilities. As to dispose of spent resins produced in PHWR, C-14 must be removed from spent resins. This C- 14 removal technology from the spent resin can increase the utilization of Gyeongju intermediate- and low-level disposal facilities, and since C-14 separated from the spent resin can be used as an expensive resource, it is necessary to maximize its economic value by recycling it. The development of C-14 removal technology from the spent resin was carried out under the supervision of Korea Hydro & Nuclear Power in 2003, but there was a limit to the C-14 removal and adsorption technology and process. After that, Sunkwang T&S, Korea Atomic Energy Research Institute, and Ulsan Institute of Science and Technology developed spent resin treatment technology with C-14-containing heavy water for the first and second phases from 2015 to 2019 and from 2019 to the present, respectively. The first study had a limitation of a pilot device with a treatment capacity of 10L per day, and the second study was insufficient in implementing the technology to separate spent resin from the mixture, and it was difficult to install on-site due to the enlarged equipment scale. The technology to be proposed in this paper overcomes the limitations of spent resin mixture separation and equipment size, which are the disadvantages of the existing technology. In addition, since 14CO2 with high concentration is stored in liquid form in the storage tank, only the necessary amount of C-14 radioactive isotope can be extracted from the storage tank and be used in necessary industrial fields such as labeling compound production. Therefore, when the facility proposed in this paper is applied for treating mixtures in spent resin tanks of PHWR, it is expected to secure field applicability and safety, and to reflect the various needs of consumers of labeled compound operators utilizing C-14.

The type of radioactive waste that may occur in the process of NPP dismantling can be classified into solid, liquid, gas, and mixed waste. Most of the radioactive waste generated during the dismantling of a NPP is metal solid waste, but liquid radioactive waste is also a very important factor in terms of radiation environmental impact assessment. In the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to design liquid radioactive waste processing system of Radioactive Waste Treatment Facility (RWTF). Depending on the amount of liquid radioactive waste generated, the type of liquid radioactive waste processing system included in the RWTF is different. In addition, in order to apply to the domestic RWTF, it is important to secure the site area occupied by the each system, the liquid radioactive waste treatment capacity of the system, and how to secure circulating water used for dilution and discharge of liquid radioactive waste. Therefore, this review aims to suggest an optimal method for the treatment system for liquid radioactive waste included in RWTF of Wolseong.

The decommissioning of Korea’s nuclear power facilities is expected to take place starting with the Kori Unit 1 followed by the Wolsong Unit 1. In Korea, since there is no experience of decommissioning, considerations of site selection for the waste treatment facilities and reasonable selection methods will be needed. Only when factors to be considered for construction are properly selected and their effects are properly analyzed, it will be possible to operate a treatment facility suitable for future decommissioning projects. Therefore, this study aims to derive factors to be considered for the site selection of treatment facilities and present a reasonable selection methodology through evaluation of these factors. In order to select a site for waste treatment facilities, three virtual locations were applied in this study: warehouse 1 to warehouse 3. Such a virtual warehouse could be regarded as a site for construction warehouses, material warehouses, annexed building sites, and parking lots in nuclear facilities. If the selection of preliminary sites was made in the draft, then it is necessary to select the influencing factors for these sites. The site of the treatment facility shall be suitable for the transfer of the waste from the place where the dismantling waste is generated to the treatment facility. In addition, in order for construction to take place, interference with existing facilities and safety should not be affected, and it should not be complicated or narrow during construction. Considering the foundation and accessibility, the construction of the facility should be economical, and the final dismantling of the facility should also be easy. In order to determine one final preferred plan with three hypothetical locations and five influencing factors, there will be complex aspects and it will be difficult to maintain consistency as the evaluation between each factor progresses. Therefore, we introduce the Analytic Hierarchical Process (AHP) methodology to perform pairwise comparison between factors to derive an optimal plan. One optimal plan was selected by evaluating the three virtual places and five factors of consideration presented in this study. Given the complexity and consistency of multiple influencing factors present and prioritizing them, AHP tools help users make decisions easier by providing simple and useful features. Above all, it will be most important to secure sufficient grounds for pairwise comparison between influencing factors and conduct an evaluation based on this.

The dose was evaluated for the workers transporting the spent resin drums from a spent resin mixture treatment facility. The treatment technology of spent resin mixture waste based on microwave was developed to compensate for the shortcoming of the existing one. The mechanism of the facility for the treatment is divided into separation, desorption, condensation and adsorption process. The treated spent resin that has passed through the microwave reactor flows into the spent resin storage tank. As the treatment time elapses, if spent resin accumulates in the spent resin storage tank, it is moved to the drum of the volume of 200 L. The drum must be moved by the worker, in which case radiation exposure to the drum transport worker occurs. It requires the dose evaluation for drum transport workers in terms of radiation safety. Dose evaluation was performed in consideration of the change in the composition ratio and weight of the spent resin mixture, where the working time for transportation was considered from 10 to 120 minutes in 10-minute increment. In the case of 100 kg of the spent resin mixture, the dose range was derived as 4.62×10−3 – 5.90×10−2 mSv for the 100 kg of spent resin, 4.72×10−3– 5.58×10−2 mSv for the 80 kg of spent resin and 20 kg of zeolite and activated carbon, and 5.38×10−3 – 6.32×10−2 mSv for the 60 kg of spent resin and 40 kg of zeolite and activated carbon. In the case of 150 kg of the spent resin mixture, the dose range was derived as 6.83×10−3 – 8.20×10−2 mSv for the 150 kg of spent resin, 7.13×10−3 – 8.22×10−2 mSv for the 120 kg of spent resin and 30 kg of zeolite and activated carbon, and 8.28×10−3 – 8.86×10−2 mSv for the 90 kg of spent resin and 60 kg of zeolite and activated carbon. The estimated maximum doses for each weight (100 kg and 150 kg of mixture) were confirmed to be 3.16×10−1% and 4.43×10−1% of the annual average dose limit of 20 mSv for radiation workers.

Starting with the permanent shutdown of Kori Unit 1, the first waste treatment facility in Korea will be built on the Kori site. In this facility, major process such as decontamination, cutting, radiation measurement and volume reduction of decommissioning waste are performed, and radioactive liquid waste is generated by the waste treatment process and personnel decontamination. The generated liquid waste is finally discharged to the sea through radioactive monitoring system after sufficient treatment to meet the standard radiological effluent control. Whereas the treated liquid waste is additionally diluted through the circulation water discharge conduit and discharged to the sea in the operating nuclear power plants, there is no circulation water in the waste treatment facility. Therefore, a new discharging method for dilution after treatment should be considered. In this paper, the treatment concept and discharge method of radioactive liquid waste system in waste treatment facility are reviewed.

The permanent shutdown of Wolseong 1, PHWR (Pressurized Heavy Water Reactor) was decided. Accordingly, there is need for C-14 treatment technology to spent resin generated by PHWR in classified Medium Level Radioactive Waste by C-14 specific activity. However, spent resin by PHWR is mixed and stored with activated carbon and zeolite (mixture), not a single storage, and separation from the mixture must be carried out in advance for C-14 treatment in the spent resin. This study developed a C-14 treatment facility that combined with the technology of separating spent resin from spent resin mixture by PHWR NPP and the technology of C-14 treatment for disposal. The C-14 treatment facility consists of spent resin separation (Part 1) and treatment of separated spent resin. (Part 2) Part 1 is applied with a process of separating the mixed and stored spent resin from the spent resin mixture by applying a drum screen method. In the case of Part 2, spent resin treatment process for desorbing and collecting C-14 nuclides in the separated spent resin using microwave reactor was applied. Except for the adsorbent used to collect C-14 detached in the process of separating and treating spent resin, no additional material is introduced into the facility, and thus secondary waste is significantly reduced. In addition, pollution prevention banks at the bottom of the facility and a sealed automated circulation system were applied to prevent unexpected leakage and diffusion of radioactive materials and ensure stability of workers. Currently, the C-14 treatment facility has been verified for spent resin separation and spent resin treatment using simulated spent resin mixture, and the facility will be demonstrated and verified for field applicability. According to derived results, it is believed that it will be possible to apply the C-14 treatment facility when decommissioning of PHWR.