국내 중·저준위 방사성폐기물은 영구적 격리를 위해 처분장에 매립하고 있으며 그 위치는 경주에 있다. 이러한 방 사성폐기물의 영구적인 격리를 위한 처분시설은 공학적 방벽과 자연 방벽으로 구성되어 있으며 자연 방벽을 특성을 파악 하기 위하여 한국원자력환경공단에서는 2006년부터 부지특성조사를 수행하였고, 이후 부지감시 및 조사계획에 따른 감시 를 수행하여 부지특성의 변화를 지속적으로 확인하고 있다. 중저준위 방폐장의 수리지화학적 환경은 자연 방벽의 평가를 위해 중요한 요소로 손꼽히고 있으나 동해와 가까운 경주의 지역적 특성상 해수의 영향을 반드시 고려해야 한다. 따라서 본 연구에서는 처분 부지의 지하수 관정 7개 및 관정의 심도별 수질 자료를 취합해 지하수 자료 총 30개를 해수 2개소와 비교 분석하여 수리지화학적 환경을 해석하였다. 분석 자료는 수질 10개 항목(온도, EC, HCO3, Na, K, Ca, Mg, Cl, SO4, SiO2)을 2017년 3분기부터 2022년 3분기까지 총 5년간 20회의 자료를 활용하였다. 특히, EC, HCO3, Na, Cl의 농도 변화 를 통해 연구 지역의 배경 농도 및 관정의 구간별 해수의 영향을 파악하였으며, 시계열 군집 분석을 통해 담수, 기수, 해 수의 분류를 시도하였다. 그 결과, 기존의 모니터링 방법으로는 확인하지 못한 부지내 수리지화학적 변화를 제시하였다.

In general, radioactive waste with high radioactivity is made into a solid form with performance such as leaching restriction, shape retention, and structural stability so that radioactive waste does not affect humans and the environment as much as possible. This should be applied equally to radioactive waste, whether homogeneous or heterogeneous. The requirements are stipulated in the “Low and Intermediate Level Radioactive Waste Delivery Regulations” notice of the Korea Nuclear Safety and Security Commission. On the other hand, the waste acceptance criteria for domestic disposal facilities require immobilization of heterogeneous waste when the activity concentration is above a certain level, but do not provide specific immobilization performance requirements. In this study, the immobilization requirements applied to heterogeneous radioactive waste in various overseas countries operating low and intermediate-level radioactive waste disposal facilities were studied. First, the IAEA’s safety standards for radioactive waste immobilization, domestic regulations, and disposal facility waste acceptance criteria were reviewed. Countries operating surface disposal facilities such as the United States, France, Spain, and Japan and countries operating underground disposal facilities such as Sweden and Finland were divided to review the current status of immobilization application to heterogeneous waste in overseas countries. When reviewing overseas cases, each country’s disposal methods, types of disposal waste, and waste treatment criteria were also reviewed. It was found that the immobilization requirements for heterogeneous radioactive waste vary depending on the disposal method and the type of barrier used to ensure disposal safety in each country. The common point is to surround heterogeneous radioactive waste within a concrete lining of a certain thickness, and to apply the thickness, compressive strength, and diffusion coefficient of the concrete lining as immobilization performance requirements. Through this study, the immobilization performance requirements for heterogeneous radioactive waste in various overseas countries that stably operate low- and intermediate-level radioactive waste were confirmed, which is expected to contribute to specifying the performance requirements for immobilization of heterogeneous radioactive waste in domestic disposal facilities.

Korea Radioactive Waste Agency (KORAD), regulatory body and civic groups are calling for an infrastructure system that can more systematically and safely manage data on the results of radioactive waste sampling and nuclide analysis in accordance with radioactive waste disposal standards. To solve this problem, a study has been conducted on the analysis of the nuclide pattern of radioactive waste on the nuclide data contained in low-and intermediate-level radioactive waste. This paper will explain the optimal repackaged algorithm for reducing radioactive waste based on previous research results. The optimal repackaged algorithm for radioactive waste reduction is comprised based on nuclide pattern association indicators, classification by nuclide level of small-packaged waste, and nuclide concentration. Optimization simulation is carried out in the order of deriving nuclide concentration by small-packaged, normalizing drum minimization as a function of purpose, normalizing constraints, and optimization. Two scenarios were applied to the simulation. In Scenario 1 (generating facilities and repackaged by medium classification without optimization), it was assumed that there are 886 low-level drums and 52 very low-level drums. In Scenario 2 (generating facilities and repackaged by medium classification with optimization), 708 and 230 drums were assigned to the low-level and very low-level drums, respectively. As a result of the simulation, when repackaged in consideration of the nuclide concentration and constraints according to the generating facility cluster & middle classification by small package (Scenario 2) the low-level drum had the effect of reducing 178 drums from the baseline value of 886 drums to 708 drums. It was found that the reduced packages were moved to the very low-level drum. The system that manages the full life-cycle of radioactive waste can be operated effectively only when the function of predicting or tracking the occurrence of radioactive waste drums from the source of radioactive waste to the disposal site is secured. If the main factors affecting the concentration and pattern of nuclides are systematically managed through these systems, the system will be used as a useful tool for policy decisions that can prevent human error and drastically reduce the generation of disposable drums.

경주 방폐물 처분시설의 1단계 시설로 건설된 지하 사일로 구조는 2014년에 10만 드럼 규모로 완공되어 현재 운영중에 있다. 지하 사일로 구조는 지름 25m, 높이 50m로써 방폐물을 저장하는 실린더부분과 돔 부분으로 구성되어 있으며, 돔부분은 운영터널과 연결 되는 하부 돔 부분과 상부 돔 부분으로 구분할 수 있다. 지하 사일로 구조의 벽체는 철근콘크리트 라이너이고, 두께는 약 1m이다. 본 논문에서는 지하 사일로 구조의 건설과정 및 운영과정의 단계별 유한요소해석을 수행하였다. SMAP-3D 프로그램을 사용하여 2차원 축대칭 유한요소해석을 수행하였다. 2차원 축대칭 유한요소모델의 신뢰성을 검토하고자 3차원 유한요소해석도 수행하였다. 본 논문 에서는 지하 사일로 구조의 구조거동을 분석하고 구조적 안전성을 검토결과를 제시하였다.

In this study, the well-known non-destructive acoustic emission (AE) and electrical resistivity methods were employed to predict quantitative damage in the silo structure of the Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC), Gyeongju, South Korea. Brazilian tensile test was conducted with a fully saturated specimen with a composition identical to that of the WLDC silo concrete. Bi-axial strain gauges, AE sensors, and electrodes were attached to the surface of the specimen to monitor changes. Both the AE hit and electrical resistance values helped in the anticipation of imminent specimen failure, which was further confirmed using a strain gauge. The quantitative damage (or damage variable) was defined according to the AE hits and electrical resistance and analyzed with stress ratio variations. Approximately 75% of the damage occurred when the stress ratio exceeded 0.5. Quantitative damage from AE hits and electrical resistance showed a good correlation (R = 0.988, RMSE = 0.044). This implies that AE and electrical resistivity can be complementarily used for damage assessment of the structure. In future, damage to dry and heated specimens will be examined using AE hits and electrical resistance, and the results will be compared with those from this study.

A safety assessment of radioactive waste repositories is a mandatory requirement process because there are possible radiological hazards owing to radionuclide migration from radioactive waste to the biosphere. For a reliable safety assessment, it is important to establish a parameter database that reflects the site-specific characteristics of the disposal facility and repository site. From this perspective, solubility, a major geochemical parameter, has been chosen as an important parameter for modeling the migration behavior of radionuclides. The solubilities were derived for Am, Ni, Tc, and U, which were major radionuclides in this study, and on-site groundwater data reflecting the operational conditions of the Gyeongju low and intermediate level radioactive waste (LILW) repository were applied to reflect the site-specific characteristics. The radiation dose was derived by applying the solubility and radionuclide inventory data to the RESRAD-OFFSITE code, and sensitivity analysis of the dose according to the solubility variation was performed. As a result, owing to the low amount of radionuclide inventory, the dose variation was insignificant. The derived solubility can be used as the main input data for the safety assessment of the Gyeongju LILW repository in the future.

As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.

한국원자력환경공단은 처분시설 내 1단계 인수·저장구역의 인수검사 공간 및 드럼 취급 공간 부족에 대한 문제를 해결하기 위하여 방폐물검사건물을 건설하여 저장·처리능력을 확충할 예정이다. 본 연구에서는 MCNP 코드를 이용하여 방폐물검사 건물 내 저장구역에서 취급하는 해체 방사성폐기물 대상 신형처분용기를 대상으로 작업종사자의 피폭선량을 평가하였다. 평가결과, 시설 내 저장 가능한 최대 용기 개수(304개)와 방사선작업에 대한 연간 예상 작업시간(약 306시간)에 대하여 연간 집단선량은 총 84.8 man-mSv로 계산되었다. 시설 내 총 304개의 신형처분용기(소형/중형 타입)가 저장 완료된 시점에서 인수검사, 처분검사를 위한 작업종사자의 투입인력은 총 25명, 작업종사자 당 예상피폭선량은 연평균 3.39 mSv로 산출 되었다. 소형용기 취급 시 작업종사자의 고방사선량 작업에 따른 작업효율과 방사선적 안전성 확보를 위해서는 콘크리트 라이너의 두께를 증가시키는 추가적인 차폐가 필요할 것으로 평가되었다. 향후 본 연구를 바탕으로 실측기반의 해체폐기 물의 선원항과 특성을 활용하여 방사선작업 당 작업시간 및 투입인력을 산출함으로써 작업종사자의 최적의 방사선작업조건을 도출할 수 있을 것으로 사료된다.