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.

The present study investigated effects of microbial additives and silo density on chemical compositions, fermentation indices, and aerobic stability of whole crop rice (WCR) silage. The WCR (“Youngwoo”) was harvested at 49.7% dry matter (DM), and ensiled into 500 kg bale silo with two different compaction pressures at 430 kgf (kilogram-force)/cm2 (LOW) and 760 kgf/cm2 (HIGH) densities. All WCR forage were applied distilled water (CON) or mixed inoculants (Lactobacillus brevis 5M2 and Lactobacillus buchneri 6M1) with 1:1 ratio at 1x105 colony forming unit/g (INO). The concentrations of DM, crude protein, ether extract, crude ash, neutral detergent fiber, and acid detergent fiber of whole crop rice before ensiling were 49.7, 9.59, 2.85, 6.74, 39.7, and 21.9%, respectively. Microbial additives and silo density did not affect the chemical compositions of WCR silage (p>0.05). The INO silages had lower lactate (p<0.001), but higher propionate (p<0.001). The LOW silages had higher lactate (p=0.004). The INO silages had higher yeast count (p<0.001) and aerobic stability (p<0.001). However, microbial counts and aerobic stability were not affected by silo density. Therefore, this study concluded that fermentation quality of WCR silage improved by microbial additives, but no effects by silo density.

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.

대표적인 산업부산물인 폐주물사와 플라이애쉬의 재활용은 매우 시급한 현안중의 하나이다. 이러한 산업부산물의 재활용을 위해 저강도 콘크리트 개념을 적용한 유동성채움재(CLSM)로의 재활용을 위한 기본 연구가 지난 3년간 수행되었다. 본 연구에서는 폐주물사중 환경오염문제의 발생 가능성이 적은 것을 선별하여 CLSM용 잔골재로 이용하였다. 특히, 폐주물사에 화학약품처리를 하여 재활용 잔골재로 이용하는 방식사 역시 실험 재료로 이용하였다. 실험은 기존에 보편적으로 이용되는 뒤채움재 시공시 발생하는 사일로 토압에 대한 검증과 토압 경감효과에 대한 연구를 수행하였다. 실험결과 CLSM은 기존의 뒤채움재와는 그 성질이 매우 달라서, 옹벽배면과 절개면사이의 거리가 짧아서 발생하는 사일로 토압효과는 나타나지 않았다. 이는 기존의 뒤채움재와는 상당히 다른 재료이고, 또한, 기존에 이용되는 각종 토압론의 공식적용에 다소 신중해야 할 것으로 판단된다. 사용된 잔골재용 시료중에서는 방식사가 가장 큰 토압경감 효과를 나타냈고, 사용된 CLSM의 잔골재 특성에 따라 토압경감 효과에 다소 시간적인 차이는 있으나, 그 절대값은 어느 정도 일정한 값으로 수렴하는 것으로 나타났다.

대표적인 산업부산물인 폐주물사와 플라이애쉬의 재활용은 매우 시급한 현안중의 하나이다. 이러한 산업부산물의 재활용을 위해 저강도 콘크리트 개념을 적용한 유동성채움재(CLSM)로의 재활용을 위한 기본 연구가 지난 3년간 수행되었다. 본 연구에서는 폐주물사중 환경오염문제의 발생 가능성이 적은 것을 선별하여 CLSM용 잔골재로 이용하였다. 특히, 폐주물사에 화학약품처리를 하여 재활용 잔골재로 이용하는 방식사 역시 실험 재료로 이용하였다. 실험은 기존에 보편적으로 이용되는 뒤채움재 시공시 발생하는 사일로 토압에 대한 검증과 토압 경감효과에 대한 연구를 수행하였다. 실험결과 CLSM은 기존의 뒤채움재와는 그 성질이 매우 달라서, 옹벽배면과 절개면사이의 거리가 짧아서 발생하는 사일로 토압효과는 나타나지 않았다. 이는 기존의 뒤채움재와는 상당히 다른 재료이고, 또한, 기존에 이용되는 각종 토압론의 공식적용에 다소 신중해야 할 것으로 판단된다. 사용된 잔골재용 시료중에서는 방식사가 가장 큰 토압경감 효과를 나타냈고, 사용된 CLSM의 잔골재 특성에 따라 토압경감 효과에 다소 시간적인 차이는 있으나, 그 절대값은 어느 정도 일정한 값으로 수렴하는 것으로 나타났다.

이 시험은 수직 사이로에 충진된 옥수수를 대상으로 M.B를 사용하여 자연중력하에서 소독할 때 M.B가스의 침투확산성 및 살충효과를 요약하면 다음과 같다. 1. 사이로 표층의 M.B가스 농도는 투약후 2시간후부터 100mg/l 이상에서 20mg/l 이하로 급격히 감소하였다. 2. 사이로 표층으로부터 3m 깊이 지점의 M.B가스 농도는 투약후 12시간이후 100mg/l 이상에서 30mg/l 이하로 감소되었으며 표층의 고농도 M.B가스가 불과 30분 이내에 3m 깊이 까지 침투되었음을 알 수 있었다. 3. 사이로 표층으로부터 고농도의 M.B가스는 13m 깊이 지점까지 1시간 이내에 침투되었으며 투약후 48시간까지 지속적으로 100mg/l 이상40mg/l 이상의 M.B가스가 검지되는 것으로 보아 자연중력하에서 옥수수를 소독하는 경우 13m 깊이 부근에 M.B 가스의 고농도 대를 형성하는 것으로 추측된다. 4. 사이로 기부(21m 지점)의 M.B가스는 에 불과했던 것으로 보아 M.B의 침투가 아주 적었던 것으로 판단되어진다. 5. M.B 가스의 횡 확산성은 수직 침투성에 비하여 매우 적었으며 사이로 길이 3m의 경우 투약방향으로부터 약 13m지점의 qdnl는 중심부의 약 , 그리고 깊이 13m 지점에서는 지점에서는 의 농도에 불과하였다. 6. 투약후 24시간만에 조사한 사이로의 표층 및 기부(21m 지점)에 삽입 했던 공시충(거짓쌀도둑)은 모두 사멸되었으며 48시간후에 조사한 3m, 7m, 및 13m 지점의 공시충도 모두 사충으로 발견되었다.

이 시험은 수직싸이로에서 대두를 훈증소독하는 경우 훈증제별로 침투확산력을 조사할 목적으로 실시한것으로써 그 결과를 요약하면 다음과 같다. 1. 대두의 표면에 M.B를 단독으로 투약한 경우 M.B Gas의 침투확산속도는 대만히 빨라 투약후 4시간 이내에 싸이로 기부에 60mg/l 이상의 M.B Gas가 검출되었으며 10시간 이후로부터는 M.B Gas의 농도가 차 감소하는 경향이었다. 2. M.B Gas의 Carrier로 Gas를 동시에 사용하였을 때는 M.B Gas의 침투확산속도가 M.B 단독 사용시보다 더욱 빨라져 투약후 1.5 시간이내에 싸이로 기부의 M.B Gas 농도는 70mg/l 이상을 나타내었다. 3. Phostoxin을 투약한 경우는 M.B를 투약한 경우 와는 반대로 침투확산력이 매우 미약하여 투약후 48시간동안 싸이로의 기부에서 10mg/l 이상의 농도를 검지할 수 없었다. 4. 소독효과를 조사하기 위해서 공시충으로 쌀바구미를 싸이로 기부에 삽입하였던바 M.B 및 구에서 는 완전한 살충효과를 얻었으나 Phostoxin 사용구에서는 대부분 생충으로 발견되었다.

The concrete silo dry storage system, which has been in operation at the Wolsong NPP site since 1992, consists of a concrete structure, a steel liner plate in the inner space, and a fuel basket. The silo system’s concrete structure must maintain structural integrity as well as adequate radiation shielding performance against the high radioactivity of spent nuclear fuel stored inside the storage system. The concrete structure is directly exposed to the external climatic environment in the storage facility and can be expected to deteriorate over time owing to the heat of spent nuclear fuel, as well as particularly cracks in the concrete structure. These cracks may reduce the radiation shielding performance of the concrete structure, potentially exceeding the silo system’s allowable radiation dose rate limits. For specimens with the same composition and physical properties as silo’s concrete structures, cracks were forcibly generated and then irradiated to measure the change in radiation dose rate to examine the effect of cracks in concrete structures on radiation shielding performance, and in the current state, the silo system maintains radiation shielding performance.

The 300 concrete silo systems installed and operated at the site of Wolsong nuclear power plant (NPP) have been storing CANDU spent nuclear fuel (SNF) under dry conditions since 1992. The dry storage system must be operated safely until SNF is delivered to an interim storage facility or final repository located outside the NPP in accordance with the SNF management policy of the country. The silo dry storage system consists of a concrete structure, liner steel plate in the inner cavity, and fuel basket. Because the components of the silo system are exposed to high energy radiation owing to the high radioactivity of SNF inside, the effects of irradiation during long-term storage must be analyzed. To this end, material specimens of each component were manufactured and subjected to irradiation and strength tests, and mechanical characteristics before and after irradiation were examined. Notably, the mechanical characteristics of the main components of the silo system were affected by irradiation during the storage of spent fuel. The test results will be used to evaluate the long-term behavior of silo systems in the future.

The Wolsong Nuclear Power Plant (NPP) operates an on-site spent fuel dry storage facility using concrete silo and vertical module systems. This facility must be safely maintained until the spent nuclear fuel (SNF) is transferred to an external interim or final disposal facility, aligning with national policies on spent nuclear fuel management. The concrete silo system, operational since 1992, requires an aging management review for its long-term operation and potential license renewal. This involves comparing aging management programs of different dry storage systems against the U.S. NRC’s guidelines for license renewal of spent nuclear fuel dry storage facilities and the U.S. DOE’s program for long-term storage. Based on this comparison, a specific aging management program for the silo system was developed. Furthermore, the facility’s current practices—periodic checks of surface dose rate, contamination, weld integrity, leakage, surface and groundwater, cumulative dose, and concrete structure—were evaluated for their suitability in managing the silo system’s aging. Based on this review, several improvements were proposed.

To ensure 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. This paper addresses the development of the safety assessment model for the underground silo of Wolseong Low-and Immediate-Level Waste (LILW) disposal facility in Korea. As the simulated result, the nuclides diffused from the waste were kept inside the silo without the leakage of those while the integrity of the concrete is maintained. After the degradation of concrete, radionuclides migrate in the same direction as the groundwater flow by mainly advection mechanism. The release of radionuclides has a positive linear relationship with a half-life in the range of medium half-life. Additionally, the solidified waste form delays and reduces the migration of radionuclides through the interaction between the nuclides and the solidified medium. Herein, the phenomenon of this delay was implemented with the mass transfer coefficient of the flux node at numerical modeling. The solidification effects, which are delaying and reducing the leakage of nuclides, were maintained the integrity of the nuclides. This effect was decreased by increasing the half-life and the mass transfer coefficient of radionuclides.

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.

In our previous study, we developed a CFD thermal analysis model for a CANDU spent fuel dry storage silo. The purpose of this model is to reasonably predict the thermal behavior within the silo, particularly Peak Cladding Temperature (PCT), from a safety perspective. The model was developed via two steps, considering optimal thermal analysis and computational efficiency. In the first step, we simplified the complex geometry of the storage basket, which stored 2,220 fuel rods, by replacing it with an equivalent heat conductor with effective thermal conductivity. Detailed CFD analysis results were utilized during this step. In the second step, we derived a thermal analysis model that realistically considered the design and heat transfer mechanisms within the silo. We developed an uncertainty quantification method rooted in the widely adopted Best Estimate Plus Uncertainty (BEPU) method in the nuclear industry. The primary objective of this method is to derive the 95/95 tolerance limits of uncertainty for critical analysis outcomes. We initiated by assessing the uncertainty associated with the CFD input mesh and the physical model applied in thermal analysis. And then, we identified key parameters related to the heat transfer mechanism in the silo, such as thermal conductivity, surface emissivity, viscosity, etc., and determined their mean values and Probability Density Functions (PDFs). Using these derived parameters, we generated CFD inputs for uncertainty quantification, following the principles of the 3rd order Wilks’ formula. By calculating inputs, A database could be constructed based on the results. And this comprehensive database allowed us not only to quantify uncertainty, but also to evaluate the most conservative estimates and assess the influence of parameters. Through the aforementioned method, we quantified the uncertainty and evaluated the most conservative estimates for both PCT and MCT. Additionally, we conducted a quantitative evaluation of parameter influences on both. The entire process from input generation to data analysis took a relatively short period of time, approximately 5 days, which shows that the developed method is efficient. In conclusion, our developed method is effective and efficient tool for quantifying uncertainty and gaining insights into the behavior of silo temperatures under various conditions.

CANDU Spent Fuel (CSF) dry storage system, SILO, has been operated from 1992 at Wolsung under 50 year operating license. As of 2023, this system has been operated for over 30 years and its licensed remaining operation time is less than 20 years. When it faces the final stage of operation, it has only two options; moving to a centralized away-from-reactor storage or extending its license atreactor. These two options have an inevitable common duty of confirming the CSF integrity by a “demonstration test”. Since the degradation of CSF and structural materials in the SILO are critically dependent on temperature, two important goals of the ‘DEMO test’ were set as follows. 1. Design of ‘DEMO SILO’: Development of internal monitoring technology by transforming SILO design. 2. Accurate measurement and evaluation of the three-dimensional temperature distribution in the ‘DEMO SILO’ Based on operating real commercial SILO dimension, a conceptual “DEMO SILO” design has been developed from 2022. Because, unlike with commercial Silo, ‘Demo Silo’ must be disassembled and assembled, and have penetration holes. Safety evaluation technologies like structural, thermal and radiation protection analysis also have been developed with design work. ‘Demo SILO’ should evaluate an accurate 3D temperature distribution with minimal number of thermocouples and penetration holes to avoid disruption of internal flow and temperature distribution. For this reason, a ‘Best Estimate Thermal-Hydraulics evaluation system for SILO’ is under development and it will be essential for ensuring temperature prediction accuracy. Construction of a full-scale test apparatus to validate this technology will begin in 2024. In order to supply power to many heaters and monitor temperature gradient inside of this apparatus, it has modular design concept by dividing its whole body to axial 9 sub-bodies which looks like a donut containing a basket at center position.

On-site storage facility using concrete silo dry storage systems for spent nuclear fuel at Wolsong NPP site came into operation in 1992 and was expanded four times, and a total of 300 silo dry storage systems are currently in operation. The design lifetime of silo dry storage systems has been licensed for 50 years. As the dry storage systems are subject to time constraints for a limited lifetime, countries operating the dry storage systems are working to ensure the long-term integrity of dry storage systems and IAEA also recommends that the dry storage systems be assessed for long-term storage. To demonstrate the long-term integrity due to material degradation during the licensed design lifetime, the structural integrity of silo dry storage systems was evaluated by considering the material degradation characteristics of concrete. The concrete compressive strength results measured so far by the rebound hammer method, which is an internationally standardized nondestructive test method for converting hardness into compressive strength using the correlation between rebound number and strength at the time of a Schmidt hammer strike, were analyzed in accordance with Wolsong NPP’s procedure to quantify the degradation characteristics, and the prediction of concrete strengths for 20 years and 50 years after construction of the silo dry storage systems was determined, respectively. Based on these residual compressive strengths, structural analyses of the silo dry storage systems were carried out under normal, off-normal and accident conditions of the related regulations, and the structural integrity of silo dry storage systems was reevaluated. It was confirmed the silo dry storage systems are able to maintain structural integrity up to the design lifetime of 50 years even if the concrete is deteriorated.

The structural integrity of concrete silos is important from the perspective of long-term operation of radioactive waste repository. Recently, the application of acoustic emission (AE) is considered as a promising technology for the systematic real-time health monitoring of concrete-like brittle material. In this study, the characteristics of AE wave propagation through concrete silo of Gyeongju radioactive waste repository were evaluated under the effects of groundwater and temperature for the quantitative damage assessment. The attenuation coefficients and absolute energies of AE waves were measured for the temperature cases of 15, 45, 75°C under dry and saturated concrete specimens, which were manufactured based on the concrete mix same as that of Gyeongju concrete silo. The geometric spreading and material loss were taken into account with regard to the wave attenuation coefficient. The attenuation coefficient shows a decreasing pattern with temperature rise for both dry and saturated specimens. The AE waves in saturated condition attenuate faster than those in dry condition. It is found that the effect of water content has a greater impact on the wave attenuation than the temperature. The results from this study will be used as valuable information for estimating the quantitative damage at the location micro-cracks are generated rather than the AE sensor location.