Radioactive waste is typically disposed of using standard 200 and 320 L drums based on acceptance criteria. However, there have been no cases evaluating the disposal and suitability of 200 L steel drums for RI waste disposal. There has been a lack of prior assessments regarding the disposal and suitability of 200 L steel drums for the disposal of RI waste. Radioactive waste is transported to disposal facilities after disposal in containers, where the drums are loaded and temporarily stored. Subsequently, after repackaging the disposal drums, the repackaged drums are transported to disposal facilities by vehicle or ship for permanent disposal. Disposal containers can be susceptible to damage due to impacts during transportation, handling, and loading, leading to potential damage to the radiation primer coating during loading. Additionally, disposal containers may be subject to damage from electrochemical corrosion, necessitating the enhancement of corrosion resistance. Metal composite coatings can be employed to enhance both abrasion resistance and corrosion resistance. The application of metal composite coatings to disposal containers can improve the durability and radiation shielding performance of radioactive waste disposal containers. The thickness of radioactive waste disposal containers is determined through radioactive shielding analysis during the design process. The designed disposal containers undergo structural analysis, considering loading conditions based on the disposal environment. This paper focuses on evaluating the structural improvements achieved through the implementation of metal composite coatings with the goal of enhancing corrosion and abrasion resistance.

도로 건설로 인한 서식지 파편화에 대한 저감방안으로 육교형 생태통로가 건설되고 있기는 하지만 효과성에 대해서는 아직도 논쟁이 있다. 생태통로의 효과성 평가를 위해 족적트랩, 카메라트랩과 같은 모니터링 방법이 실시되고는 있으나 얼마나 많은 개체가 이용하는지 정량적으로 평가하기에는 한계가 있다. 이에 본 연구에서는 생태통로와 인근 지역을 서식지로 이용할 가능성이 큰 소형포유류인 등줄쥐를 대상으로 포획-재포획 방법으로 개체 위치 파악을 통해 생태통로 이용 정도를 도출하고, 트랩 주변 환경 특성을 이용하여 등줄쥐의 생태통로 이용에 미치는 요인을 확인하였다. 등줄쥐의 생태통로 이용도는 격자 단위의 포획지점을 연결하여 이동 거리와 경로를 확인하였고, 생태통로 이용에 미치는 환경 특성은 트랩당 포획 횟수를 종속변수로 한 일반화 선형 모형(Generalized linear model)을 이용하였다. 연구결과, 등줄쥐의 이동 거리는 선행연구와 유사하게 나타났으며, 생태통로를 횡단하는 개체가 나타나지 않음에 따라 등줄쥐는 생태통로를 통로보다는 서식지로 이용함을 확인하였다. 등줄쥐가 생태통로를 이용하는 데 영향을 미친 환경 특성은 층위별 식생피복량(1~2m, 2~8m, 8m 이상), 교목 식생, 트랩 주변 최대 수목 흉고직경, 경사도가 유의하게 나타났다. 이에 따라 생태통로 조성 시 더 많은 교목과 관목을 식재하고, 높은 경사와 절토사면 생성을 방지하여 생태계 내 먹이원으로 이용될 수 있는 등줄쥐 이용도를 높인다면 생태통로의 효과성을 더 높일 수 있을 것으로 예상된다.

The disposal criteria of the domestic LILW disposal facility specifies that fluidized substances such as the spent resin, the evaporator bottom should be solidified in a physically stable solid form, such as cementation and polymerization. And the solidified form applies requirements for compressive strength, immersion test, thermal circulation test, radiation irradiation test, leaching test, and free standing water measurement test. On the other hand, it is specified that immobilization iss applied to wastes with a total radioactivity concentration of more than 74,000 of radionuclides with a half-life exceeding 20 years among non-homogeneous wastes such as spent filters and DAW, but the test requirements are not applied. Nevertheless, it is necessary for waste generator to establish quality control standards for the manufacture of immobilized solid form through reviewing overseas cases and domestic regulations and technical standards. The test requirements for solidified solid form require measurement of structural stability (compressive strength, immersion, thermal cycling, irradiation test), leachability (leaching test), and free standing water measurement. A characteristic of the immobilized solid form is that it is not mixed with the waste and that the cement medium surrounds the waste. Therefore, the structural soundness is higher than that of the solidified solid mixed with waste. In addition, even when in contact with water, the cement medium blocks the contact between waste and water, thereby preventing the spread of radionuclides. Therefore, considering the characteristics of these immobilized solid form, compressive strength test and free standing water measurement are applied for structural soundness. For other tests, it is determined that application is unnecessary.

There are generally two kinds of spent filter; one is spent filter media for mainly gaseous purification such as HEPA filter, the other is spent filter cartridge for liquid purification such as CVCS BRS cartridge type filter. The spent filter cartridge from liquid purification system has been storing in special shielding space in auxiliary building in NPPs since the beginning of 2006 according to the long term storage strategy for decaying short lived radionuclide and gaining the time for selecting practical treatment technology before final packaging. The spent filter cartridges generated Kori-1 reactor vary in their sizes as in length from 913 mm to 290 mm and range in radiation level from several hundred mSv per hour to below mSv per hour . It is high time that the spent filter cartridge is treated and packaged because LILW repository in Wolsung area is operating and Kori-1 reactor is scheduled to decommission. The spent filter cartridge is one of the wet solid wastes required of solidification. It is difficult for the spent filter cartridge to solidify because of their shape, structure, physical and chemical characteristics in addition to having high radiation level. NSSC notice defines that solidification of wet solid wastes include that solid material such as spent filter is encapsulated with cement, etc. as a form of macro-encapsulation. The radioactive waste acceptance criteria describes that non-homogeneous waste having above 74,000 Bq/g such as spent filter, dry active waste should be encapsulated with qualified material. Homogeneous waste such as spent resin, sludge, concentrated waste (liquid waste evaporator bottoms), etc. should be solidified complied with requirements except that spent filter which is allowed to encapsulate. It is needed to guide to the practice of these two requirements for spent filter. The sampling and test method is different between homogeneous solidification waste form and spent filter cartridge encapsulation waste form. For example, how core sample can be taken and how void space can be measured among spent filter cartridge in encapsulation waste form. The technical evaluation report for spent filter cartridge polymer encapsulation by US NRC has been reviewed and the technical position of US NRC was identified. As a result of review, improvement fields of waste acceptance criteria for spent filters are pointed out, and the technical position of US NRC for spent filter cartridge solidification is summarized. The recommendation on improvement directions for spent filter cartridge encapsulation is suggested.

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, 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.