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.

The objective of this study is development of graphite-boron composite material as a replacement for metal canisters to Improve the heat dissipation and radiation shielding performance of dry spent nuclear fuel storage system and reduce the volume of waste storage system. KEARI research team plan to use the graphite matrix manufacturing technology to pelletize the graphite matrix and adjust the content of phenolic resin binder to minimize pore formation. Specifically, we plan to adjust the ratio of natural and synthetic graphite powder and use uniaxial pressing technology to manufacture black graphite matrix with extremely high radial thermal conductivity. After optimizing the thermal conductivity of the graphite matrix, we plan to mix it with selected boron compounds, shape it, and perform sintering and purification heat treatments at high temperatures to manufacture standard composite materials.

Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.

There have been a variety of issues related to spent nuclear fuel in Korea recently. Most of the issues are related to intermediate storage and disposal of spent nuclear fuel. However, recently, various studies have been started in advanced nuclear countries such as the United States to reduce spent nuclear fuel, focusing on measures to reduce spent nuclear fuel. In this study, a simple preliminary assessment of the thermal part was performed for the consolidation storage method which separates fuel rods from spent nuclear fuel and stores them. The preliminary thermal evaluation was analyzed separately for storing the spent fuel in fuel assembly state and separating the fuel rods and storing them. The consolidation storage method in separating the fuel rods was advantageous in terms of thermal conductivity. However, detailed evaluation should be performed considering heat transfer by convection and vessel shape when storing multiple fuel bundles simultaneously.

The Alkali-Metal Thermal to Electric Converter (AMTEC) can be used as a next-generation power generation technology related with a large thermal energy storage. In particular, this technology is expected for the higher efficiency by a cascade power generation with the thermoelectric generator(TEG), and the temperature distribution becomes a very important design parameter in this case. In this study, the temperature distribution of the AMTEC unit was analyzed through CFD analysis, and design points were discussed based on the results.

The conventional research trend on spent fuel was safety analysis based on mechanical perspective. Analysis of spent fuel cladding is based on the temperature of cladding and pressure inside cladding. To improve fuel cladding analysis, precise and accurate thermal safety evaluation is required. In this study a database which is about thermal conductivity and emissivity for the thermal modeling was established for a long-term safety analysis of spent fuel. As a result, we confirmed that the thermal conductivity of zirconium hydride was not accounted in conventional model such as FRAPCON and MATPRO. The conductivity of zirconium and its oxide was evaluated only as a function of temperature. However, the behavior of heat conductivity and emissivity is determined by the change of the material properties. The material properties depend on the microstructural characteristic. It can be seen that this conventional approach does not consider the microstructure change behavior according to vacuum drying process or burn-up induced degradation phenomena. To improve the thermal properties of spent nuclear fuel cladding, the measurement experiments of heat conduction and emissivity are required according to spent fuel experience and status such as the number of vacuum drying, cooling rate, burn up, hydrogen concentration and oxidation degree. In previous domestic reports and papers, we found that relative data between thermal properties and spent fuel experience and status does not exist. Recently, in order to understand the failure mechanism of hydrogen embrittlement, many studies have been conducted by accounting and spent fuel experience and status in a mechanical perspective. If microstructure information could be obtained from these studies, the modeling of thermal conductivity and emissivity will be possible indirectly. According to a recent abroad paper, it was confirmed that the thermal conductivity decreased by about 30% due to irradiation damage. The radiation damage effects on thermal conductivity also has not been studied in zirconium oxide and hydride. These un-revealed phenomena will be considered for the thermal safety model of spent fuel.

본 논문은 에너지를 실시간으로 저장할 수 있는 저장장치 중 열에너지 저장 콘크리트를 대상으로 재료의 미세구조와 물성(열전도 도)의 상관관계를 분석하는 연구를 수행하였다. 에너지 저장 콘크리트의 열전도 성능을 증가시키기 위해 혼화재인 그라파이트 (graphite)를 사용하였다. 그라파이트가 시멘트 질량의 10%와 15%를 치환한 시편과 일반 콘크리트(OPC) 시편을 제작하여 그라파이 트의 혼입에 따른 미세구조 변화 및 열전도도의 영향을 마이크로 스케일에서 분석하였다. 마이크로-CT를 활용하여 OPC와 그라파이 트를 사용한 콘크리트의 공극률을 비교하였으며, 확률함수를 사용하여 미세구조 특성을 정량화하였다. 미세구조 특성 차이가 열전도 도에 미치는 영향을 확인하기 위해 3차원 가상 시편을 제작하여 열해석을 수행하였으며, 이를 열평판법을 사용하여 측정한 열전도도 실험 결과와 비교하였다. 열해석 수행 시 그라파이트 재료가 지닌 열전도도 성능을 반영하기 위하여 해석 결과와 실험 결과를 기반으 로 고체상의 열전도도를 역해석을 통해 계산하였으며, 그라파이트가 시편의 열전도도에 미치는 영향에 대해 분석하였다.

Bird screen meshes are installed at the air inlet and outlet ducts of spent fuel storage casks to inhibit the intrusion of debris from the external environment. The presence of these screens introduces an additional resistance to air flow through the ducts. In this study, a porous media model was developed to simplify the bird screen meshes. CFD analyses were used to derive and verify the flow resistance factors for the porous media model. Thermal analyses were carried out for concrete storage cask using the porous media model. Thermal tests were performed for concrete casks with bird screen meshes. The measured temperatures were compared with the analysis results for the porous model. The analysis results agreed well with the test results. The analysis temperatures were slightly higher than the test temperatures. Therefore, the reliability and conservatism of the analysis results for the porous model have been verified.

Recently, uncertainty of demand and supply for energy has expanded as energy market change rapidly. Rapid accumulation storing of cold heat is a method for decreasing the gap between demand and supply of time, quantity and quality of heat energy. It is stored rapidly when electric power is on. And cold heat energy in heat is phase change material (PCM) heat storage is used when electric power is off. Porous media for cold heat is PCM. The present study has been conducted for improving performance of cold heat storage in refrigerator using characteristics on laten heat and phase change of PCM. This study provide basic data to improve performance of cold heat using characteristic of PCM. Also, this study predicts and measures specific points in PCM heat storage during process of phase change.

플랜트의 유황 저장 탱크는 강재로 구성되며, 탱크 저면은 앵커볼트에 의해 Ring Wall 형상의 콘크리트 기초와 연결된다. 탱크 내 유황이 내부 열원에 의해 고온상태를 유지하기 때문에, 유황 저장 탱크는 상온의 유체를 저장하는 다른 탱크에 비해 큰 체적팽창을 겪게 된다. 일반적으로 탱크 기초의 구조설계는 기초의 내외부의 온도차를 하중으로 적용한 구조해석이 수행 되는데, 이 방법은 탱크의 열팽창 특성이 앵커볼트에 의해 집중하중 형태로 콘크리트 기초에 전달되는 현상을 고려할 수 없 다. 이는 온도하중의 영향을 과소평가하게 되며, 앵커볼트에 인접한 콘크리트의 균열을 야기한다. 본 연구는 앵커볼트에 의 한 온도 하중전달 메커니즘을 고려한 하중 평가식을 제안함으로써, 콘크리트 기초에 작용하는 하중을 보다 합리적으로 결정 하고자 한다. 이를 위해 탱크 바닥판과 앵커볼트가 포함된 유한요소모델을 이용해 앵커볼트 개수 증감에 따른 온도하중의 변화를 분석하였으며, 분석결과를 이론해와 결합해 앵커볼트에 의해 전달되는 하중을 평가할 수 있는 명시적인 형태로 해를 제시하였다. 제안된 식의 유효성을 확인하기 위해 실제 플랜트 현장의 유황 저장 탱크 설계에 적용하였으며, 실무적으로 사 용 가능함을 보였다.