Decarbonization plays an important role in future energy systems for establishing a zero-carbon society. Hydrogen is believed to be a promising energy source that can be converted, stored, and utilized efficiently, leading to a broad range of possibilities for future applications. Hydrogen can be stored in various forms, including compressed gas, liquid hydrogen, hydrides, adsorbed hydrogen. Among these, liquid hydrogen has high gravimetric and volumetric hydrogen densities. There are a lot of previous studies on thermal behavior of MLI and VCS and optimization insulation system, but research on the insulation performance by varying the head shape of the tank has not been conducted. In this study, thermal-structural coupled analysis was conducted on the insulation system with VCS positioned between two layers of MLI for a liquid hydrogen storage tank. The analysis considered dome shapes (torispherical, circle, ellipses), and heat flux and temperature were derived from thermal analysis to predict insulation performance. Maximum equivalent stress and deformation were calculated from the structural analysis, and the optimal dome shape was proposed.

With the importance of permanent disposal of high-level radioactive waste (HLW) generated in Korea, the deep geological disposal system based on the KBS-3 type is being developed. Since the deep geological repository must provide the long-term isolation of HLW from the surface environment and normal habitats for humans, plants, and animals, it is essential to assess the longterm performance of the disposal facility considering thermal-hydraulic-mechanical-chemical (TH- M-C) evolution. Decay heat dissipated from HLW contained in the canister causes an increase in temperature in the adjacent area. The requirement for the maximum temperature is established in consideration of the possibility of bentonite degradation. Therefore, when designing the repository, the temperature in the region of interest should be identified in detail through the thermal evolution assessment to ensure that the design requirement is satisfied. In the thermal evolution analysis, it is needed to evaluate the temperature distribution over the entire area of the disposal panel to consider the heat generated from both a single canister and adjacent canisters. Computational fluid dynamics (CFD) codes are widely used for detailed temperature analysis but are limited to simulating a wide range. Accordingly, in this study, we developed an analytical solution-based program for efficiently calculating the temperature distribution throughout the deposition panel, which is based on threedimensional heat conduction equations. The code developed can assess the temperature distribution of engineered and natural barrier systems. Principal parameters to be inputted are as follows: (a) geometry of the panel (e.g. width, length, height, spacing between canisters), (b) geometry of the canister (e.g. diameter, height), (c) thermal properties of bentonite and host-rock, (d) initial conditions (e.g. residual heat, temperature), and (e) time information (e.g. canister emplacement rate, time-interval, period). Through the calculation for the conceptual problem of a deposition panel capable of accommodating 900 (i.e. 30×30) canisters, it was confirmed that the program can adequately predict when and where the maximum temperature will occur. It is expected that the overall temperature distribution within the panel can be obtained by the evaluation of the entire region using this program reflecting the detailed design of the repository to be developed in the future. In addition, the thermal evolution analysis considering the influence of other canisters can be performed by applying the results as boundary conditions in the CFD analysis.

In order to ensure the long-term safety of a deep geological repository, the performance assessment of the Engineered Barrier System (EBS) considering a thermal process should be performed. The maximum temperature at the side wall of a disposal canister for the technical design requirement should not exceed 100°C. In this study, the thermal modelling was conducted to analyze the effects of the thermal process from a disposal canister to the surrounding near-field host rock using the PFLOTRAN code. The mesh was generated using the LaGriT code and the material properties were assigned by applying the FracMan code. Initial conditions were set as the average geothermal gradient (25.7°C/km) and an average surface temperature (14.7°C) in Korea. The highest temperature was observed at the middle of the canister side wall. The temperature of the buffer was lower than that of the canister, and the temperature increase of the deposition tunnel and the host rock was insignificant due to the lower effect of the heat source. The result of the thermal evolution of the EBS represented the highest thermal effects in the vicinity of the canister. In addition, the thermal effects were largely decreased after 10 years of the entire simulation period. It demonstrated that the model took 3 years to heat up the buffer around the canister. The temperature at the canister side wall increased until 3 years and then decreased after that time. This is because that the radioactive decay heat from the heat source was emitted enough to raise the overall temperature of the EBS by 3 years. However, the decay heat rate of the canister decreased exponentially with the disposal time and then its decay heat was not emitted enough after 3 years. In conclusion, the peak temperature results of the EBS were lower than 70°C to meet the technical design requirement.

The design and fabrication of suitable waste forms with high thermal and structural stability are essential for the safe management and disposal of radioactive wastes. In particular, the thermal properties and temperature distribution of waste form containing high heat-generating nuclides such as Cs and Sr can be used to evaluate its thermal stability, but also provide useful information for the design of canisters, storage systems, and repositories. In this study, a new program code-based thermal analysis framework has been developed to facilitate the characterization, design, and optimization of the waste form. Matlab was used as a software development platform because it provides powerful mathematical computation and visualization components such as the partial differential equation (PDE) toolbox for solving heat transfer problems using finite element method, the App Designer for graphical user interface (GUI), and the MATLAB Compiler for sharing MATLAB programs as standalone applications and web applications. The thermal analysis results such as temperature distribution, heat flux, maximum/ minimum temperature, and centerline/surface temperature profile are visualized with graphs and tables. To evaluate the effectiveness of the developed program, several design and optimization studies were carried out for the SrTiO3 waste form, selected as a stable form of strontium nuclide.

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.

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.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. Thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. COBRA-SFS is a legacy code based on subchannel analysis code, and its fidelity is verified for evaluating the thermal analysis for licensing a dry cask system. Herein, thermal analysis result based on CFD and COBRA-SFS codes is compared and the Dry Cask Simulator (DCS) is assessed as a benchmark experiment in this study. Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to address the degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT was used to simulate DCS. Simulations were conducted in various decay heat and helium pressure inside the canister. In realistic conditions of decay heat and helium pressure of actual dry cask system, CFD and COBRA-SFS analysis results gave good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also showed good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K. In high decay heat and high pressure condition, however, CFD and COBRA-SFS underestimated peak cladding temperature than experimental results.

The process basket assembly is an important module in pyroprocess, because pyroprocess is a batch process, so process materials are contained in a basket assembly and transferred with the basket. The basket assembly is composed of upper and lower assembly. The lower assembly is a basket or crucible which contains process materials, and it can have electrodes. The upper assembly mainly consists of heat shields, a flange, and connectors for supplying currents to electrodes of the lower assembly. During the electrolytic recovery process, the lower part is submerged into molten salt, whose temperature is about 500°C at least and the heat from salt is transferred to the upper assembly. And the heat affects the performance or durability of parts on the top of equipment and can raise cell temperature, which is an undesired situation. In addition, the handling equipment can pick the assembly when it is hottest, and during the transfer, the gripped part is under thermal and mechanical stresses. Because of this, the thermal effects from the heat should be required during equipment design stage. In this study, the thermal analyses of process basket assembly were conducted for 3 cases: the steady state of the basket assembly when it submerged in molten salt, the thermal and mechanical stresses when gripped by remote handling device, and the temperature changes under natural convection. These analyses were performed using Solidworks with flow simulation package, and the results will apply to improve the thermal resistant performance of the basket assembly.

This study presents a method for analyzing the surface temperatures of specific facilities, such as the 5 MWe reactor within the Yongbyon Nuclear Complex, to explore its potential utility in monitoring suspected nuclear-related activities in North Korea using thermal infrared (TIR) satellite imagery (Landsat series). TIR band data is utilized to derive surface temperatures in the specified areas, and the temperatures are analyzed on a monthly basis to examine any patterns within these regions. This research provides a pattern-of-life on temperature variation for the target areas through multiple TIR image datasets, offering additional information to analyze facilities’ operational status in remote and inaccessible regions.

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

This study aims to evaluate the structural safety of a structural thermal barrier, installed inside the structure of a building and performed the role of a load-bearing element and an insulation simultaneously, contributing to the realization of net-zero buildings. To ensure the reliability of the analysis model, the analysis results derived from LS-DYNA were compared with the experimental results. Based on the results shown through the flexural experiment, the reliability of the thermal cross-section insulation structure model for slabs was validated. In addition, the effect of the UHPC block on the load support performance and its contribution to vertical deflection was verified.

본 연구에서는 육계사에 차열 페인트와 히트펌프의 적용에 따른 내부 온도 변화를 분석 하였다. 이를 위하여 환기율, 환기 방법, 시간별 환기 변화에 따른 실험 조건을 설정하였으며 육 계사 외부 및 내부 기온을 측정하였다. 그 결과, 차열 페인트를 도포한 육계사에서는 최대 1－2°C 실내 기온 상승을 억제하 는 효과가 나타났으며 히트펌프를 가동한 육계사에서는 외기 온도의 영향을 제일 적게 받는 환기율 0%일 때 내부 기온 감소 가 제일 크게 나타났다. 계사 내부의 온도가 외기 온도보다 높 을 경우에는 환기율을 높게 설정하여 환기팬을 이용한 냉방이 더욱 효과적이나 계사 내부 온도가 외기 온도와 유사하거나 낮을 경우에는 히트펌프를 이용하는 것이 가장 효과적일 것으 로 판단된다. 히트펌프 가동 시 외기 온도의 영향이 적은 환기 율을 0%로 설정하였을 때 내부 기온이 가장 큰 폭으로 감소하 였으나 실제 육계사에서는 분진, 이물질, 암모니아 등을 고려 하여 최소환기율 정도로 환기율을 설정한 후 히트펌프를 가동 하는 것이 가장 효율적일 것으로 판단된다. 본 연구는 실험 기 간이 짧아 데이터가 많지 않으며 실제 육계가 사육되고 있는 환경에서 실험을 진행한 것이 아니라는 한계가 있다. 향후 후 속 연구로 실제 닭이 사육되고 있는 환경에서의 히트펌프 효 과 분석과 히트펌프의 전력사용량, 냉방부하, 환기팬 가동시 간 등 다양한 환경인자를 포함한 연구가 진행되어야 할 것으 로 판단된다.

In this study, the thermal equilibrium of a motor operated in the sea and the temperature in the equilibrium were studied. To predict the equilibrium temperature in the sea, the cooling performance of the motor was studied by comparing results of analysis and experimental results in the air condition. By this study, the method of prediction of the cooling performance of a motor in various environments could be useful.

To analyze the radioactivity of 3H and 14C in miscellaneous radioactive wastes generated from nuclear power plants, a wet digestion method using sulfuric acid is currently used. However, sulfuric acid is classified as a special management material, and there is no disposal method for contaminated radioactive waste. Therefore, research on a thermal decomposition method that can analyze the DAW radioactive waste samples without using sulfuric acid is necessary. In this study, we will cover the final sample amount, sample injection method, and prevention of organic ignition to meet the minimum detection limit requirements of the analysis equipment. Through this research, optimal conditions for the thermal decomposition method for analyzing the radioactivity of 3H and 14C in DAW radioactive wastes generated from nuclear power plants can be derived.

Radionuclide analysis methods must be secured in the event of emergencies such as the discovery of unknown nuclear material or nuclear accidents in neighboring countries or Korea. Most institutions in Korea are in their early stages of radionuclide analysis method development and do not even have Radiation Controlled Areas where they can handle the samples safely. Some institutions such as the Korea Atomic Energy Research Institute have the ability to perform radionuclide analysis for nuclear facilities or verification of nuclear activities. In Korea, it is necessary to secure nuclide analysis technology to enable independent verification in times of emergency or need. This paper analyzes uranium as the target nuclide using alpha spectrometer and TIMS. Alpha spectrometer detects alpha particles emitted from uranium samples and measures the concentration of uranium isotopes. This method has a high selectivity that distinguishes it from other elements, and accurate measurements can be made even when uranium samples are mixed with other elements. In addition, there is minimal interference from other radioactive isotopes in the sample, and the sample preparation is simple, resulting in relatively short analysis times. In contrast, TIMS detects ionized uranium ions by heating the uranium sample. This method may have potential interference from other elements and may take relatively longer analysis times. However, TIMS has high sensitivity and accuracy and can detect various elements other than uranium, making it suitable for various analyses. Therefore, when analyzing uranium, it is recommended to select and use the appropriate device according to the purpose, as both alpha spectrometer and TIMS have their pros and cons. Furthermore, by using both devices in parallel, more accurate and reliable results can be obtained. This paper aims to compare the analysis methods of alpha spectrometer and thermal ionization mass spectrometry, which are widely used for nuclide analysis in unknown nuclear materials.

Spent nuclear fuels released from the reactor are stored in cooling pools and then stored in dry storage casks. During the transition from the wet storage to dry storage cask, a vacuum drying process is used to remove residual water in the cask. During the vacuum drying process, gas pressure is reduced to below 400 Pa to promote evaporation and water removal. KAERI is developing a PWR single assembly (PLUS7) test equipment to simulate the thermal flow in spent fuel assembly. In this study, the thermal conductivity of air at low pressure was derived to perform the thermal analysis of the canister in vacuum. In addition, thermal analyses were performed for the canister with backfill gases of helium, air, and a vacuum in the vertical orientation using the COBRA-SFS code. At low pressure, the thermal conductivity of air depends on pressure and temperature. The reduced thermal conductivity, kr (W/m-K) was calculated using the curve fit for air at reduced pressure in thin gaps presented in the General Electric Fluid Flow Handbook. 􀝇􀯥/􀝇􀬴 = 􀬵 􀬵􀬾 􀮼􀯍/􀯉􀰋 Where, k0 is the thermal conductivity at atmospheric pressure (W/m-K), P is the reduced (vacuum) pressure (Pa), δ is the gap size (m), T is the temperature (K), and C is the Lasance constant (7.657E-5 N/m-K). The thermal conductivity of air decreases as the pressure decreases. The reduced thermal conductivity of air at pressures of 400 Pa and 40 Pa was calculated to be 0.97 and 0.77, respectively. For the analysis in vacuum, no enhancement of the convective heat transfer was assumed (Nu=1.0). For the helium backfill, the peak cladding temperature was the lowest and the axial temperature profile was the flattest due to the higher thermal conductivity and lower density of the helium. For the vacuum backfill, the peak cladding temperature was the highest and temperature gradient was the sharpest due to the only radiative heat transfer effect in the fuel assembly.

This study deals with the maximum thermal load analysis and optimal capacity determination method of tank culture system for applying seawater source heat pump to save energy and realize zero energy. The location of the fish farm was divided into four sea areas, and the heat load in summer and winter was analyzed, respectively. In addition, two representative methods, the flow-through aquaculture system and the recirculation aquaculture system were reviewed as water treatment methods for fish farms. In addition, the concept of the exchange rate was introduced to obtain the maximum heat load of the fish farms. Finally, power consumption for heat pumps was analyzed in the view point of sea areas, tank capacity, and exchange rate based on the calculated maximum thermal load.

In this study, we performed thermal safety design of the electric module of a heat-loaded equipment with consideration of its heat dissipation performance. Initially, we calculated the heat dissipation of natural convection to choose a cooling method. Based on this, we found that some modules required forced convection and selected an air-cooling method with an outdoor temperature of 43 degrees Celsius, which is the maximum temperature in Korea. Prior to module production, we performed thermal analysis of each module and proceeded with a design to increase the thermal conductivity of the module as a primary step, and subsequently proceeded with Heat Sink design to maximize the heat dissipation performance. After considering various constraints according to the system requirements and designing the cooling path, we experimentally and analytically secured thermal safety at the operating temperature of the equipment.