Structural stability of a waste form can be provided by the waste form itself (steel components, etc.), by processing the waste to a stable form (solidification, etc.), or by emplacing the waste in a container or structure that provides stability (HICs or engineered structure, etc.). The waste or container should be resistant to degradation caused by radiation effects. In accordance with the requirements for the domestic waste acceptance criteria, irradiation testing of solidified waste forms containing spent resin should be conducted on specimens exposed to a dose of 1.0E+6 Gy and other material 1.0E+7 Gy. Expected cumulative dose over 300 years is about 1.770E+6 Gy for spent resin and 0.770E+6 Gy for dried concentrated waste generated from NPPs generally. According to NRC Waste Form Technical Position, to ensure that spent resins will not undergo adverse degradation effects from radiation, resins should not be generated having loadings that will produce greater than 1E+6 Gy total accumulated dose. If it necessary to load resins higher than 1E+6 Gy, it should be demonstrated that the resin will not undergo radiation degradation at the proposed higher loading. This is the recommended maximum activity level for organic resins based on evidence that while a measurable amount of damage to the resin will occur at 1E+6 Gy, the amount of damage will have negligible effect on disposal site safety. Cementitious materials are not affected by gamma radiation to in excess of 1E+6 Gy. Therefore, for cement-stabilized waste forms, irradiation qualification testing need not be conducted unless the waste forms contain spent resins or other organic media or the expected cumulative dose on waste forms containing other materials is greater than 1E+7 Gy. Testing should be performed on specimens exposed to IE+6 Gy or the expected maximum dose greater than 1E+6 Gy for waste forms that contain ion exchange resins or other organic media or the expected maximum dose greater than 1E+7 Gy for other waste forms. This is suggestion as a review result that requirement for irradiation testing of solidified waste forms has something to be revise in detail and definitively.

The “shadow zone” is defined as a region below a flow obstacle, such as a vault, in unsaturated soils. Due to the capillary discontinuity of the cavity, water saturation on the top and side of the cavity is higher than the ambient saturation. On the bottom of the cavity, however, there is a region where water saturation is lower than ambient saturation. Undoubtedly, a shadow zone may also exist below a LILW disposal vault built in subsurface soils above the water table before the vault is fully degraded. During the degradation, flow in the shadow zone is controlled by the rate of water infiltrating the degrading vault. In this study, as one of the efforts to be made for enhancing safety margin by a realistic safety assessment of the engineered vault type LILW disposal facility, the shadow zone effect is investigated by a numerical parametric study using AMBER code. The conceptual model and data were excerpted from IAEA, ISAM Vault Test Case for the liquid release design scenario. It is assumed that the nearfield barriers degrade with time. In order to compare a visible shadow zone effect, the vault degradation period is assumed to be both 500 and 1,000 years, and the shadow zone depth to be varied according to unsaturated zone lithology. It can be seen that with a shorter shadow zone (2.7 m), radionuclides arrive at the water table earlier than with a full shadow zone (55 m) due to increased advection rate in the unsaturated zone. This effect tends to be more visible in the case of a longer degradation period. For radionuclides with short residence time relative to their half-lives in the unsaturated zone, such as Tc-99 and I-129, the radionuclides are shown to come out because they will arrive sooner, thereby allowing less peak release rate, when the shadow zone effect is considered. Once the vault is completely degraded and the infiltration rate of water flowing through the vault is equal to the ambient rate, the shadow zone effect disappears. In this example calculations using IAEA ISAM Vault Test Case input parameters, it might not be shown a significant shadow zone effect. Nevertheless, when the extent of the shadow zone is determined through more sophisticated hydraulic studies in the unsaturated soils surrounding the vault, the shadow zone effect would be checked up on the realistic near-field radionuclide transport modeling in order to contribute to gaining safety margins for post-closure safety assessment of the Wolsong 2nd phase LILW disposal facility.

To obtain confidence in the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. Thus, safety assessment models require uncertainty management as a key part of the confidencebuilding process. In application to the numerical modelling, the global sensitivity analysis is widely employed for dealing with parametric and conceptual uncertainties. In particular, the parametric uncertainty can be effectively reduced by minimizing the uncertainty of critical parameters in the safety assessment model. In this paper, the numerical model of each step disposal facility (Silo, Near surface, and Trench type) at Wolsong Low and Immediate Level Waste (LILW) Disposal Center is designed by using a two-dimensional finite element code (COMSOL Multiphysics). In order to determine the critical parameters for non-adsorbed nuclides such as H-3, C-14, Tc-99, we introduced the variance-based sensitivity analysis methodology of the global sensitivity analysis. In the case of Silo type, the density of waste is highly sensitive to the total leakage quantity of all nuclides. Additionally, the initial nuclide concentration of H-3 was identified as another important parameter of H-3. On the other hands, the mass transport coefficient showed a high contribution in C-14 and Tc-99. In other types of disposal facilities, the leaking properties of H-3 are significantly affected by the amount of infiltration water. However, C-14 and Tc-99 were found to be more sensitive to the density of waste.

본 논문에서는 우리나라의 중저준위 방폐물 처분을 위한 사일로 형식 지하동굴의 유한요소해석을 수행하였다. 사일로의 벽체부분 은 지름 25m의 원형구조이고, 높이는 35m이다. 사일로의 천장부분은 지름 30m의 돔 형식이고, 높이 17.4m의 규모이다. 사일로는 해 수면으로부터 –80m에서 –130m에 위치하고 있다. 중저준위 방폐물 처분 1단계 시설로 6개의 사일로가 건설되어 운영되고 있으나, 본 연구에서는 1개의 사일로에 대해서 고려하였다. SMAP-3D 프로그램을 사용하여 2차원 축대칭 유한요소모델과 3차원 유한요소모델 을 생성하였다. Generalized Hoek and Brown Model이 수치해석에 적용되었다. 다양한 측압계수(수평방향 현장응력과 수직방향 현장 응력의 비)의 변화에 따른 사일로 형식 지하동굴의 유한요소해석을 수행하였으며, 수치해석결과 및 분석결과가 제시되었다.

Numerical model was developed that simulates radionuclide (3 H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3 H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3 H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

본 연구는 경주 중·저준위처분장 2단계 표층처분시설의 폐쇄 후 안전성에 대한 불확실성을 예측·평가하기 위하여 수행되 었다. 다중덮개와 처분고의 건전/열화를 고려한 총4가지의 시나리오를 도출하여 강우침투 시 예상되는 처분시설 내부의 유 체 이동을 모사하였다. 강우 조건은 총 30년(1985~2014) 간의 월평균 데이터를 적용하였으며, 시뮬레이션 기간은 제도적 관 리기간인 300년으로 설정하였다. 처분덮개와 처분고 콘크리트 모두 건전성을 유지하는 조건의 기본 시나리오 평가 결과, 처 분시설 내부의 처분고를 완전히 포화시키지 못하는 것을 확인할 수 있었다. 다중 덮개층을 구성하는 8개 층의 각 매질의 모 세관 압력과 투과도 차이로 인하여 다중 덮개층이 효과적으로 차수·배수 역할을 하는 것으로 나타났다.

중·저준위 방사성폐기물 처분시설의 운영 중 사고로 인한 방사선적 영향을 평가하기 위해서는 운영 중 발생 가능한 사고 에 대한 타당성이 입증되어야 한다. 본 논문에서는 처분시설의 운영 중 사고분석 체계를 처분시설의 구성요소에 대한 안전 기능분석, 잠재위험요소분석, 위험도분석, 그리고 향후 조치대안으로 사고평가체계를 개선하였다. 이를 위하여 위험도분석 에 필요한 설계대안과 관리대안을 추가하여 설계-운영-평가가 연계되도록 하였다. 또한 운영 중 사고의 발생확률과 평가결 과의 심각성에 따라 운영중 사고에 대한 분류기준을 제안하여 처분시설 운영 중 대표 사고시나리오에 대한 정당성을 확보 하였다. 본 논문의 개선된 평가체계를 우리나라의 2단계 중·저준위 방사성폐기물 표층처분시설에 대한 처분시설 운영 중 사고분석의 사례에 대해 적용하였다.

The low and intermediate-level radioactive waste generated in Korea is disposed of at Wolsong Disposal Facility. For the safety of a disposal facility, it must be assessed by considering some abnormal scenarios including human intrusion as well as those by natural phenomena. The human intrusion scenario is a scenario that an incognizant man of the disposal facility will be occurred by the drilling. In this paper, the well usage scenario was classified into the human intrusion event as the probability of the well drilling is very low during the man’s lifecycle and then was assessed by using conservative assumptions. This scenario was assessed using the dilution factor of contaminants released from a disposal facility and then it was introduced the applied methodology in this study. The assessed scenario using this methodology is satisfied the regulatory limits.

본 연구에서는 중·저준위방사성폐기물 처분시설(이하 처분시설)에서 발생하는 기체의 이동현상을 예측하기 위한 2차원 수 치 모델링을 수행하였다. 또한, 기체 이동 모델링에서 주요 입력변수로 적용되는 사일로 콘크리트의 기체침투압(gas entry pressure)와 기체 투과도(gas permeability)를 실측하여, 모델링 입력변수로 적용하였다. 사일로 콘크리트의 기체침투압(gas entry pressure)와 기체 투과도(gas permeability)는 각각 0.97±0.15 bar 및 2.44×10-17 m2로 측정되었다. 기체 이동 모델링 결과, 사일로 내부에서 발생하는 수소 기체는 기상으로 이동하지 않고 지하수에 용해되어 지하수와 함께 생태계로 이동하는 것을 알 수 있다. 또한, 폐쇄 후 약 1,000 년 후 부터 사일로 상부부터 수소기체 밀도가 증가하기 시작하는 것으로 예측되었 다. 따라서, 사일로 내부에서 발생된 기체는 기상으로 사일로 내부에 축적되지 않으며, 이로 인해 사일로 콘크리트의 내구성 에 영향을 미치지 않을 것으로 판단된다.

중저준위 방사성폐기물 처분장의 안전성 평가를 위하여 지하 사일로와 그 주변의 굴착손상영역 (EDZ) 및 단열암반을 고려한 지하수유동해석과 핵종이동해석의 통합모델을 개발하였다. 사일로를 다중방벽개념으로 고려하여 사일로를 구성하는 3개의 특성지역 (waste, buffer, concrete)으로 구분하여 해석하였고, EDZ는 사 일로 주변과 건설운영 터널 주변의 손상영역을 고려하였다. 단열암반의 불균일성은 분리단열 (discrete fractures)로 부터 해석된 불균일한 지하수 유속계로 도출하였고, 그 결과를 핵종의 이동경로를 모사하는데 사 용하였다. 현 모델은 핵종누출에 따른 사일로 배치의 최적화와 안전성의 정량화를 도출하는데 사용가능하다

To validate the previous conceptual design of cover system, construction of the engineered barrier test facility is completed and the performance tests of the disposal cover system are conducted. The disposal test facility is composed of the multi-purpose working space, the six test cells and the disposal information space for the PR center. The dedicated detection system measures the water content, the temperature, the matric potential of each cover layer and the accumulated water volume of lateral drainage. Short-term experiments on the disposal cover layer using the artificial rainfall system are implemented. The sand drainage layer shows the satisfactory performance as intended in the design stage. The artificial rainfall does not affect the temperature of cover layers. It is investigated that high water infiltration of the artificial rainfall changes the matric potential in each cover layer. This facility is expected to increase the public information about the national radioactive waste disposal program and the effort for the safety of the planned disposal facility.