The spent filters stored in Kori Unit 1 are planned that compressed and disposed for volume reduction. However, shielding reinforcement is required to package high-dose spent filters in a 200 L drum. So, in this study suggests a shielding thickness that can satisfy the surface dose criteria of 10 mSv·h−1 when packaging several compressed spent filters into 200 L drums, and the number of drums required for the compressed spent filter packaging was calculated. In this study, representative gamma-emitting nuclides in spent filter are assumed that Co-60 and Cs-137, and dose reduction due to half-life is not considered, because the date of occurrence and nuclide information of the stored spent filter are not accurate. The shielding material is assumed to be concrete, and the thickness of the shielding is assumed to 18 cm considering the diameter of the spent filter and compression mold. Considering the height of the compressed spent filter and the internal height of the shielding drum, assuming the placement of the compressed spent filter in the drum in the vertical direction only, the maximum number of packaging of the compressed spent filter is 3. When applying a 18 cm thick concrete shield, the maximum dose of the spent filter can packaged in the drum is 125 mSv·h−1, so when packaging 3 spent filters of the same dose, the dose of a spent filter shall not exceed 41 mSv·h−1 and not exceed 62 mSv·h−1when packing 2 spent filters. Therefore, the dose ranges of spent filters that can be packaged in a drum are classified into three groups: 0–41 mSv·h−1, 41–62 mSv·h−1, and 62–125 mSv·h−1based on 41 mSv·h−1, 62 mSv·h−1, and 125 mSv·h−1. When 227 spent filters stored in the filter room are classified according to the above dose group, 207, 3 and 4 spent filters are distributed in each group, and the number of shielding drums required to pack the appropriate number of spent filters in each dose group is 75. Meanwhile, 8 spent filters exceeding 125 mSv·h−1 and 5 spent filters that has without dose information are excluded from compression and packaging until the treatment and disposal method are prepared. In the future, we will segmentation of waste filter dose groups through the consideration of dose reduction and horizontal placement of compressed spent filters, and derive the minimum number of drums required for compressed spent filter packaging.

Currently, treatment and disposal suitability verification methods have not been established for radioactive waste, such as spent filters temporarily stored in each plant, so the WCP (Waste Certification Program) can be applied to verify the suitability of non-conforming waste at the site. In this study, WCP components such as certification organizations, certification methods, certification documents, and quality assurance (QA) plan that should be considered when developing WCP applicable to spent filter disposal were reviewed and presented. First, a certification organization consists of a certification organization that performs certification work, a certification support organization related to waste generation and treatment, and a quality control organization for waste certification. Especially, the support organization should support the implementation of WCP, so that spent filter processing procedures such as generation information management and immobilization can be properly packaged and transported. Second, in identifying the waste characteristics of the certification method, each characteristic identification procedure and certification method of the acceptance criteria should be described, evidence examining the suitability of general, radiological, physical, chemical, and biological requirements, and processes related to measurement and sampling should be established. In identifying characteristics, satisfaction of waste form, free water requirements, and whether it is subject to immobilization should be checked priorly, and a method of confirming particulate matter and securing filling rate when packaging compressed filters should be included. It is very important to develop a technology for verifying the safety and quality of the immobilized material because immobilization of the filters can be a processing method that satisfies various characteristic criteria. Meanwhile, it is essential to collect samples and develop scaling factors to identify the nuclides of filters and prove that they are below the concentration limits. For chemical and biological requirements, the characteristics are identified through generation information documents, corrective actions are taken and documented in case of nonconformance. Third, certification documents should include immobilization procedure manual, characteristic report, and characteristic test manuals such as free water, particulate matter and filling rate, radiation measurement method manual for packages, profile, and generation documents. Fourth, the QA plan should analyze the QA system of the plants, check the QA inspection details, establish general requirements for QA of spent filter disposal, and specify step-by-step certification work QA activities. In this study, considerations to ensure the disposal suitability at all stages from generation to disposal of spent filter were presented, and development of a WCP could contribute to preventing nonconformance.

Currently, in domestic nuclear power plants (NPP), the spent filters (SFs) used for the purpose of reducing and purifying the radiation of the primary cooling water system are temporarily stored in an untreated state. In order to dispose of SFs, radioactive nuclide analysis (RNA) of SFs is required to be conducted. As segmented gamma scanner (SGS) is already being used in Kori NPP, utilizing SGS for RNA of SFs would be practical and economical. In this paper, factors required to be considered to improve accuracy of SGSs for RNA of SFs are studied. The analysis of the nuclide inventory of the packaging drum for radioactive waste should be performed by the indirect drum nuclide analysis method. The material of the SFs is iron (SS304) on the outside, and paper on the inside. In addition, to meet disposal acceptance criteria, radioactive waste drums are packaged in thick grouting or shielding drums. Therefore, it is necessary to derive an appropriate correction method for high inhomogeneity and thick media. Considering these factors, evaluating radionuclides inventory plans to measure gamma rays in SGS mode. Correct the gamma ray measurement by examining the medium attenuation factor and error factors. In this way, the inventory of gamma nuclides is calculated, and the specific radioactivity of beta ray and alpha particle emitting nuclides other than gamma rays is planned to be calculated by applying scaling factors.

The background of the development is to contribute to the reduction of radioactive waste, recycling of resources and effectively purifying the air in the workplace. Ultimately, it affects the reduction of internal exposure of workers. According to the standard procedure of KHNP,「Use and Management of Respiratory Protection Equipment」, the expiration date of mask filter is indicated by the manufacturer before opening. It is 1 year from the date of first combination after opening. We have developed an air purifying equipment that can recycle and reuse expired mask filter waste in nuclear power plant. In order to confirm the performance, we observed air pollution level by operation time. The location was measured at 3 locations including the decontamination product warehouse in NPP, and the size of the measurement locations were 53 m3, 150 m3, 180 m3, and 900 m3. As a result of measurement, significant air purification effect was found in 53 m3 and 150 m3. Decontamination effect of 80% was shown after 1 hour of operation, and 20% of decontamination effect was shown gently for 3 hours thereafter. On the other hand, there was no significant decontamination effect in the 180 m3 and 900 m3 spaces. Significant results were derived by statistical methods. Statistical procedure involves the collection of data leading to test of the relationship between two statistical data sets, or a data set and synthetic data drawn from an idealized model. The basic composition and product characteristics was as follows: Blower, filter fixing unit, Air purifier outlet round shape, Differential pressure gauge, inverter (200 V, 3π, 200 W). The developed product weigh is 25 kg. This is lighter than the existing product weighing 100 kg. It is judged that it is suitable for convenient use. Because the area where the major air pollution level occurs is isolated by a room in NPP. This developed product has a greater significance in that it recycles radioactive waste within the radiation management area rather than air purification efficiency.

A porous photocatalyst concrete filter was successfully produced to remove NOx, by mixing TiO2 photocatalyst with lightweight aerated concrete. Ultra Fine Bubbles were used to form continuous pores inside the porous photocatalytic concrete filter, which was mixed via a bubble generation experiment. The optimal mixing condition was determined to be with 4% of the bubble generation agent B. NO removal specimens were prepared for various photocatalytic loading conditions, and the specimen containing 3% P-25 removed NO at a concentration of 1.03 μmol in 1 h. The NO removal rate of the porous photocatalytic concrete filter prepared in this study was 10.99%. This photocatalytic filter performance was more than 9 times the amount of NO removed by a general photocatalytic filter. The porous photocatalyst concrete filter for removing NOx developed in this study can be applied to various construction sites and the air quality can be solved by reducing NOx contributing to the formation of fine particles.

본 연구에서는 stainless steel로 재질로 된 금속평막모듈을 이용하여 고플럭스가 유지되면서 처리수의 안정화 방 안을 모색하였다. 이 모듈은 기공사이즈가 13 μm 단위여서 플럭스가 60 LMH에서 100 LMH까지 고플럭스로 운전이 가능하 다 그러나 SS가 초기 운전 시 30~50 ppm 정도 유출되지만 SS가 응집핵으로 작용하므로 응집이 가능하게 된다. 기존 고분자 막 여과수는 응집핵이 없어서 coagulation은 되지만 floculation이 안되므로 추가적으로 응집보조제인 clay나 벤토나이트를 투 여하게 되는데 본 연구에서는 이런 응집보조제 필요 없이 SS 누출만으로 floculation이 되므로 총인처리와 처리수질이 안정성 을 도모하고자 하였다. 최종적으로 안정적인 처리수에 고플럭스가 가능한 Metal필터 운전이 MBR 시스템에서 적용가능한지 타당성을 연구하고자 하였다.

In this study, the removal performance of high-concentration H2S and complex odors was evaluated for bio-filters installed in sewage treatment plants and manure treatment plants. The amount of odor generation according to temperature was found to be higher in summer than in spring. It was found that the longer the empty bed retention time of the bio-filter, the higher the odor removal efficiency. Therefore, in order to effectively remove odors, it is necessary to maintain a sufficiently long residence time when designing a bio-filter. In addition, a case in which a bio-filter and a wet pre-treatment system were combined to remove high-concentration odors was studied. The result showed that the wet pre-treatment was effective in removing high-concentration odors. In particular, most of the hydrogen sulfide could be adequately removed by wet pre-treatment.