As the importance of radioactive waste management has emerged, quality assurance management of radioactive waste has been legally mandated and the Korea Radioactive Waste Agency (KORAD) established the “Waste Acceptance Criteria for the 1st Phase Disposal Facility of the Wolsong Lowand Intermediate-Level Waste Disposal Center (WAC)”, the detailed guideline for radioactive waste acceptance. Accordingly, the Korea Atomic Energy Research Institute (KAERI) introduced a radioactive waste quality assurance management system and developed detailed procedures for performing the waste packaging and characterization methods suggested in the WAC. In this study, we reviewed the radioactive waste characterization method established by the KAERI to meet the WAC presented by the KORAD. In the WAC, the characterization items for the disposal of radioactive waste were divided into six major categories (general requirements, solidification and immobilization requirements, radiological, physical, chemical, and biological requirements), and each subcategories are shown in detail under the major classification. In order to satisfy the characterization criteria for each detailed item, KAERI divided the procedure into a characterization item performed during the packaging process of radioactive waste, a separate test item, and a characterization item performed after the packaging was completed. Based on the KAERI’s radioactive waste packaging procedure, the procedure for characterization of the above items is summarized as follows. First, during the radioactive waste packaging process, the characterization corresponding to the general requirements (waste type) is performed, such as checking the classification status of the contents and checking whether there are substances unsuitable for disposal, etc. Also, characterization corresponding to the physical requirements is performed by checking the void fraction in waste package and visual confirmation of particulate matter, substances containg free water, ect. In addition, chemical and biological requirements can be characterized by visually confirming that no hazardous chemicals (explosive, flammable, gaseous substances, perishables, infectious substances, etc.) are included during the packaging process, and by taking pictures at each packaging steps. Items for characterization using separate test samples include radiological, physical, and chemical requirements. The detailed items include identification of radionuclide and radioactivity concentration, particulate matter identification test, free water and chelate content measurement tests, etc. Characterization items performing after the packaging is completed include general requirements such as measuring the weight and height of packages and radiological requirements such as measurements of surface dose rate and contamination, etc. All of the above procedures are proceduralized and managed in the radioactive waste quality assurance procedure, and a report including the characterization results is prepared and submitted when requesting acceptance of radioactive waste. The characterization of KAERI’s radioactive waste has been systematically established and progressed under the quality assurance system. In the future, we plan to supplement various items that require further improvement, and through this, we can expect to improve the reliability of radioactive waste management and activate the final disposal of KAERI’s radioactive waste.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

Concrete waste generated in the result of dismantling a concrete structure in a radiation control area and refractory brick waste generated from uranium pellet sintering furnace are surface-contaminated by uranium particle of which the enrichment is below 5%. These wastes are hard to decontaminate so it was necessary to develop the process for its disposal. So, we developed the Process Control Plan (PCP) for disposal of radioactive concrete waste describing a whole sequence of disposal and inspecting procedures based on the KNF Radioactive Waste Quality Assurance Plan (KN-WQAP) established in 2021. Based on the PCP, we crushed the concrete waste by jaw-crusher. Then we sieved the crushed concrete waste and removed the particle of which size is below 0.3 mm, using sieve-vibrator where the 0.3 mm mesh-sized sieve is installed inside. Before conducting the crush-sieving method based on the PCP, we conducted Process Control Assessment (PCA) based on the KN-WQAP. The purpose of the PCA is to check whether the output of the process satisfies the Acceptance Criteria of Korea Radioactive Waste Agency (KORAD) so that we could confirm the validity of the PCP. The evaluation item of the PCA is a particulate size verification test. The test is passed only if the component ratio of a particle size below 0.2 mm is less than 15% and the particle size below 0.01 mm is less than 1%. The very first 3 drums passed the test, so we began applying the PCP to whole target drums. In the process of conducting the crush-sieving method in earnest, qualified inspectors based on KNWQAP participated conducting sampling, measuring and checking whether a foreign material was included. They tested samples and packaged drums regarding 5 spheres of general, radiological, physical, chemical and biological characteristic. KNF disposed concrete and refractory brick waste by the crush-sieving method so that KNF could take over 100 drums to KORAD in 2021. But, it is needed to be improved that a dust size below 0.3 mm is generated as a secondary waste which needs to be solidified for the final disposal and the work environment is not good enough because of the dust.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.

Currently, the most widely accepted disposal concept for long-term isolation of high level radioactive waste including spent nuclear fuels is to disposal in a deep geological repository designed and constructed with multiple barriers composed of engineered and natural barriers so that the waste can be completely isolated in a stable deep geological environment. In this concept, an important consideration is the heat generated from the waste due to the large amount of fission products present in the high level waste loaded in the disposal container. For safe and complete isolation of high level radioactive waste in the deep geology, the disposal concepts that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. In this paper, the deposition hole configuration and the decay heat dissipation area (surface area) of disposal container were considered for the efficient thermal management in the deep geological disposal concept. Heat transfer through the waste form, its container and surrounding components and the rock will be mainly by conduction. Heat transfer by radiation and convection can be negligible after backfilling. When considering heat conduction, according to Fourier’s law, if the thermal conductivity of the repository components is the same, the greater the heat dissipation area and the adjacent temperature gradient, the greater the conduction effect. Therefore, rather than the conventional concept of loading 4 PWR spent fuel assemblies per disposal container and placing one disposal container in a deposition hole, it is better to load one assembly per disposal container and place 4 disposal containers in a deposition hole. In this case, it was found that the disposal area could be reduced through efficient thermal management. Considering this thermal management method as an alternative to the concept of deep geological disposal, additional research is needed.

As the decommissioning of Kori Unit 1 progresses, securing technology for treatment and disposal of radioactive wastes that have not been disposed of so far, such as spent filters, is recognized as an urgent task. In this study, a method of confirming the disposal suitability of spent filters was presented by reviewing the waste characteristics as presented in the waste acceptance criteria (WAC). The waste characteristics to be satisfied to ensure disposal suitability of waste are largely classified into general requirements, solidification and immobilization requirements, radiological requirements, physical requirements, chemical requirements, and biological requirements. First, the general requirement is to prove that the prohibited waste form has not been introduced into items related to waste form and packaging, and to confirm the suitability of disposal through step-by-step packaging photos, generation information, X-ray inspection, and visual inspection. Second, in the solidification and immobilization requirements, spent filters are non-homogeneous waste, and if the total radioactivity concentration of nuclides with a half-life of more than 20 years is 74,000 Bq·g−1 or more, they must be immobilized. Third, in order to meet the characteristic criteria for nuclides and radioactivity concentration, sampling and scaling factors development are required and based on this, nuclides must be identified and demonstrated to be below the disposal concentration limits. Surface dose rate and surface contamination should be measured in accordance with standardized procedures and disposal suitability should be confirmed through document tests recording the measured values. Fourth, in order to satisfy the physical requirements of the particulate matter and filling rate characteristics, the spent filter must be immobilized, if necessary, thereby ensuring disposal suitability. Meanwhile, free water in the spent filter should be removed through pre-drying and dehydration, and the disposal suitability should be confirmed by applying a test. Fifth, the criteria for chelating agents should be checked for disposal suitability through operation records and component analysis of spent filters, and documents, that can prove harmful substances are removed in advance and no harmful substances are included in the package, should be provided. Lastly, in biological requirements, if the spent filters contain corrosive or infectious substances, they should be removed in advance and disposal suitability should be confirmed by providing documents that can prove that such substances are not included in the package.

Radioactive waste disposal facility in Korea, radioactive waste packaged in 200 L drums is placed in a concrete disposal container and disposed of at an underground silo type (cave) disposal facility. At this time, the disposal container cover is seated on the top of the disposal container, and if the disposal container and the cover are not completely combined, the container cover is raised up from the top of the disposal container, so safety problems may occur when stacking the disposal container. Therefore, various methods exist to secure a margin for the pure height inside the disposal container. The disposal container cover only covers the upper surface of the container to shield radiation, and structural performance is not required. Therefore, the method of processing the cover, such as a method of making the cover of the disposal container thin, is the easiest method to apply. In this study, a method to reduce the thickness of the cover of a concrete disposal container was devised, and structural performance under usability conditions such as lifting and seating was analyzed. In addition, the disposal container cover has a reinforced concrete form in which dissimilar materials (concrete and steel) are combined, an integrated analysis was performed to secure the reliability of the analysis results for this, and the analysis results were described. It was found that the proposed disposal container cover structure can improve usability by reducing the stress concentration phenomenon.

To attenuate and control the spread of infectious disease, a body of research has been conducted to generate safe vaccines and to continue national-level surveillance. However, understanding on viability and persistence of avian influenza virus (AIV) in infected carcasses, and effective disposal approaches are still limited up to date. Here, using HA test and RT-PCR, we assessed active status of AIV and degradation of viral RNA in collected specimens at different sites and time points. First, AIV infectivity was recovered until day 2, and viral nucleic acids persisted to day 14 and 21 in inorganic and organic samples, respectively, in sealed vials incubated at room temperature. Second, AIV was totally inactivated in all examined specimens, and viral RNA was not detectable at all time points tested at least one month post-infection in AIV-inoculated carcasses buried directly in soil or fiber reinforced plastic (FRP) bin. Lastly, among different burial sites in South Korea, 6 out of 17 sampling sites in Jeonbuk province showed the presence of viral genetic materials, while the rest of the field samples displayed neither the presence of infective AIV nor detectable viral RNA. This study showed a linear relation between time and degradation degree of viral RNA in buried samples suggesting that burial disposal method is effective for the control or at least attenuation of spread of AI infection in infected animals although consistent monitoring is required to verify safety of disposal.

목적: 국내 콘택트렌즈 착용자들의 렌즈 사용 후 폐기처분 방법, 잘못된 폐기방법으로 발생될 수 있는 미세플라스틱에 의한 환경오염에 대한 인식을 알아보고자 하였다. 방법: 성인 261명(남자 124명, 여자 137명, 평균 나이 21.48±3.14세)을 대상으로 콘택트렌즈 주 구매처, 착용 콘택트렌즈 종류, 콘택트렌즈 착용 기간, 콘택트렌즈 폐기방법에 대한 정보나 교육을 받은 적이 있는지 여부, 콘택트렌즈 사용 후 폐기처분 방법, 환경오염 발생에 대한 인식에 대한 온라인 설문조사를 실시하였다. 결과: 콘택트렌즈 주 구매처는 안경원(50.0%), 렌즈샵(48.3%) 순으로 나타났으며, 콘택트렌즈 종류로는 매일착용 렌즈 (52.5%), 일회용 렌즈(38.5%)를 가장 많이 착용하고 있었다. 콘택트렌즈를 착용해 온 기간은 5년 이상(29.3%), 1년 미만(26.0%), 1년 이상~3년 미만(26.0%) 순으로 나타났으며, 1주일 중 콘택트렌즈를 착용하는 기간은 1~2일 착용(32.0%)이 제일 많았고, 1주일 내내(28.0%), 5-6일 착용(22.4%), 3-4일(17.6%) 순으로 응답하였다. 콘택트렌즈 구매처에서 폐기처분에 대한 교육이나 정보를 접했는지 여부는 “아니다(78.3%)”, “그렇다(21.7%)”로 나타났으며, 평소 학교나 공공기관, 인터넷 등 대중매체에서 콘택트렌즈 폐기처분에 대한 교육이나 정보를 접했는지 여부를 묻는 설문에는 교육받은 적 “없다(87.5%)”, “있다(12.5%)”로 나타났다. 폐기처분 방법으로는 매립용 쓰레기(45.6%), 재활용 쓰레기(29.6%), 싱크대나 화장실을 통한 배수구(16.8%) 순으로 응답하였다. 남성이 여성보다 폐기처분에 대한 교육이나 정보를 더 많이 접했지만(t=3.63189, p<0.00001), 여성이 환경오염에 대한 인식이 더 높은 것으로 나타났다(t=2.44269, p=0.01605). 결론: 하수처리 시설에서 분해되지 않고 미세플라스틱으로 변하는 콘택트렌즈로 인한 환경오염 문제를 줄이기 위해 콘택트렌즈 사용 후 올바른 폐기방법에 대한 정보를 제공하고 이에 대한 교육이 시급한 것으로 생각된다.

현행 규제요건에 따르면 국내에서 발생된 모든 폐밀봉선원은 자체처분 대상, 극저준위 또는 중·저준위 방사성폐기물에 해 당하며, 기본적으로 방사능 농도를 기준으로 한 처분방식 제한규정을 준수해야 한다. 본 연구에서는 이러한 분류체계 이외 에 IAEA 및 국외 폐밀봉선원 사용국의 방사성폐기물 분류체계, 폐밀봉선원 고유 특성 등에 대한 검토 및 분석결과를 토대 로 반감기 및 A/D 값(각 선원의 방사능(A)을 작업자 및 일반 대중에 대한 잠재적 위험도를 의미하는 방사성핵종 고유의‘D 값’을 활용하여 정규화한 수치로 선원의 상대적 위험을 평가하는 기초적인 기준으로 사용)에 대한 기준을 추가적으로 적용 하여 국내 폐밀봉선원 분류체계에 대한 방안을 제시한 후, 각 범주에 대한 처분방식을 도출하였다. 다양한 처분시점을 상정 한 국내 폐밀봉선원 특성 분석 및 처분방안별 대상 수량·체적 평가결과를 통해 본 연구에서 도출된 처분방안을 처분 예상 시기와 무관하게 2015년 3월말 기준으로 임시저장 중인 모든 폐밀봉선원에 대해 적용할 수 있음을 확인하였다. 단, 방사능 량을 확인할 수 없거나 비방사능 또는 A/D 값을 산출할 수 없는 선원에 대해서는 본 연구결과를 적용할 수 없으므로 처분방 안 이행을 위해서는 사전에 비방사능, 체적 등의 선원 고유 특성이 반드시 확인되어야 한다.

Carcasses of pigs were trench buried using either general soil or mature compost as a cover material and the malodorous substances discharged were observed about a year. With the soil burial method, the speed of decay was shown to be dominantly affected by the ambient air temperature. However the compost burial method’s decaying process took place quickly, even ambient air temperature was dropped; it holds the temperature of 40oC or higher. With the compost burial method, there was a period where, the temperature inside the pig carcasses and the temperature of cover-material layer were strongly reversed. From this discovery, level of decay process could be speculated. With the soil burial method there was a trend when malodorous substances concentration was high, the level of concentration in the cover soil was also tends to be high. However, the compost burial method had different result. When malodorous substances concentration was high the level of concentration in the compost cover layer was observed to be lower. This indicates compost burial method shown to intercept and absorb malodorous substances. Furthermore, the compost burial method appears to be able to contribute to deactivate the pathogens by quickly decompose the carcasses at a high temperature.

자연환기는 처분장의 작업 환경 및 위생, 부유 방사성 핵종의 노출 등과 같은 안전문제에 있어 자연환기 자체만으로는 기계적 강제 환기에 비해서 덜 효과적이지만 처분장 내의 수분제거, 작업 환경 조성과 관련하여 라돈 (Rn) 가스의 희석과 같은 향후 처분장의 장기적 환경을 위해서는 중요한 역할을 할 수 있고, 환풍기와 같은 환기 설비를 이용해야하는 기계환기에 비해 경제적으로 매우 효과적 일수 있다. 본 논문에서는 지하 처분장의 건설 및 운영 기간동안 자연환경 조건에 따라 처분장에 스스로 생기는 자연 환기의 타당성에 대하여 기술하였다. 자연 동굴을 통한 자연환기 유사에 의해 밝혀진 증거들과, 수직갱을 갖는 산악 도로터널에서의 자연 환기 측정, 그리고 주어진 자연환기 압력에 의한 공기 발생량 계산 등을 통해서 자연 환기는 한국형 지하 방사성 폐기물 처분장에 잠재적으로 매우 유익함을 알 수 있다. 효과적으로 유도된 자연 환기는 방사성 폐기물 처분장 내에 발생하는 열과 습도, 그리고 라돈 가스를 제어하기에 경제적으로 좋은 방법이 될 수 있다. 자연환기를 통해 처분장의 전반적 열적 특성은 개선될 수 있고, 수분으로 포화된 공기는 효과적으로 건조되고 그 건조상태 유지 기간은 확장 될 수 있을 것이다.