The primary heat transport system consists mainly of the in-core fuel channels connected to the steam generators by a system of feeder pipes and headers. The feeders and headers are made of carbon steel. Feeders run vertically upwards from the fuel channels across the face of the reactor and horizontally over the refueling machine to the headers. Structural materials of the primary systems of nuclear power plants (NPPs) are exposed to high temperature and pressure conditions, so that the materials employed in these plants have to take into accounts a useful design life of at least 30 years. The corrosion products, mainly iron oxides, are generated from the carbon steel corrosion which is the main constituent of the feeder pipes and headers of this circuit. Typical film thickness on CANDU-PHWR surface is 75μm or 30mg/cm2. Deposits on PHWR tends to be much thicker than PWR due to use of carbon steel and also for the source of corrosion products available on the carbon steel surface. Degradation of carbon steel for the feeder pipes transferring the primary system coolant by flow-assisted corrosion in high temperature has been reported in CANDU reactors including Point Lapreau, Gentully-2, Darlington and Bruce NPPs. The formation of Fe3O4 film on a carbon steel surface reduces the dissolution rate of steel substantially. The protectiveness of the Fe3O4 film over the carbon steel is affected by the environmental factors and the operational parameters of the feeder pipes, including the velocity, wall shear stress, solution pH, temperature, concentration of dissolved iron, quality of solution, etc. For effective chemical decontamination of these thick oxides containing radionuclides such as Co-60, it is necessary to understand the corrosion behaviors of feeder pipes and the characteristics of oxide formed on it. In this work, we investigated the growth of oxide films that develop on type SA-107 Gr. B carbon steel in high temperature water and steam environment by scanning electron microscopy (SEM) and glow discharge optical emission spectrometry (GD-OES) for the quantification and the solidstate speciation of metal oxide films. This study was especially focused to set the experimental tests conditions how to increase the oxide thickness up to 50 m by changing the oxidation conditions, such as solution chemistry and thermo-hydraulic conditions both temperature and pressure and so on.

Dry active wastes (DAWs) are combustible waste generated during the operation and decommissioning of nuclear facilities, and are known to be generated in the amount of approximately 10,000 to 40,000 drums (based on 200 L) per unit. It consists of various types of protective clothing, paper, and plastic bags, and is stored in radioactive waste storage facilities. Therefore, reducing the volume of DAWs is an important issue in order to reduce storage costs and utilize the limited space of waste storage facilities. Heat treatment such as incineration can dramatically reduce the volume of waste, but as the waste is thermally decomposed, CO2, a global warming gas, is generated and there is a risk of emissions of harmful gases including radionuclides. Therefore, a heat treatment process that minimizes the generation of CO2 and harmful gases is necessary. One of the alternatives to incineration is to carbonize DAWs, dispose of carbonized materials below the release standard as non-radioactive waste, and selectively separate and stabilize inorganic components, including radionuclides, from carbonized DAWs. In this study, 13 types of DAWs generated from nuclear power plants were selected and their thermal decomposition characteristics were investigated to design a heat treatment process that replaces incineration. As a result of TGA analysis, the temperature at which thermal decomposition of each waste begins is 260-300°C for cotton, 320-330°C for paper, 315-420°C for synthetic fiber, 350°C for latex gloves. The mass of most samples decreased to less than 1 % of the initial weight after heat treatment, and dust suit and latex gloves had residues of 13.83% and 13.71% of the initial mass, respectively. The metal components of the residue produced after heat treatment of the sample were analyzed by EDS. According to the EDS results, cotton contains Ca and Al, paper contains Ca, Al and Si, synthetic fiber contains Ca, Cu and Ti, latex gloves contain Ca and Mg. Additionally, ICP analysis was performed to quantify the inorganic components. These results are expected to be applicable to the processing of DAW generated at nuclear facilities in the future.

In the nuclear environment, sensors ensure safety, monitoring, and operational efficiency under various operating conditions. These sensors come in various forms, each tailored to specific purposes, including nuclear safety and security, waste treatment and storage, gas leak detection, temperature and humidity monitoring, and corrosion detection. Ensuring the longevity of sensors without the need for frequent replacements is a vital goal for researchers in this field. This paper explores materials that can act as shields to protect sensors from harsh environmental conditions (high radiation and temperatures) to enhance their lifetime. The types of material that had been explored were divided into categories: metal and non-metal. Fourteen types of metal and seven different plastic materials were studied and focused on their characteristics and current applications. Considering properties like melting point, intensity, and conductivity, plastic materials are chosen to be examined as sensor shielding material. A preliminary experiment was conducted to verify signal characteristics changes by shielding material. Metal material and plastic material each were placed in the middle of the granite and the target sensor. The result showed that when metal is between the granite and the sensor, the density and impedance are higher in granite than in the metal. This leads to signal attenuation and a shift in resonance frequency, while plastic does not. Therefore, PPS (Polyphenylene sulfide) and PAI (Polyamide-imide) have lower density and impedance than granite while also possessing heat, moisture, and radiation resistance for effective shielding.

The thermal evaluations for the conceptual design of the deep geological repository considering the improved modeling of the spent fuel decay heat were conducted using COMSOL Multiphysics computational program. The maximum temperature at the surface of a disposal canister for the technical design requirement should not exceed 100°C. However, the peak temperature at the canister surface should not exceed 95°C considering the safety margin of 5°C due to several uncertainties. All thermal evaluations were based on the time-dependent simulation from the emplacement time of the canister to 100,000 years later. In particular, the heat source condition was set to the decay heat rate and axial decay heat profile of the PLUS7 fuel with 4.0wt% U-235 and 45 GWD/MTU. The thermal properties of the granitic rock in South Korea were applied to the host rock region. For the reference design case, the cooling time of the SNF was set to 40 years, the distance between the deposition holes 8 meters and that between the deposition tunnels 30 meters. However, the peak temperature at the canister surface at 10 years was 95.979°C greater than 95°C. This design did not meet the thermal safety requirement and needed to be modified. For the first modified case, when the distance between the deposition tunnels was set to 30 meters, three cooling time cases of 40, 50 and 60 years and five distances of 6, 7, 8, 9 and 10 meters between the deposition holes were considered. The design with the distances of 9 and 10 meters between the deposition holes for the cooling time of 40 years and all five distances for 50 and 60 years were less than 95°C. For the second modified case, when the distance between the deposition holes was set to 8 meters, three cooling time cases of 40, 50 and 60 years and five distances of 20, 25, 30, 35 and 40 meters between the deposition tunnels were considered. The design with the distances of 35 and 40 meters between the deposition tunnels for the cooling time of 40 years, the distances of 25, 30, 35 and 40 meters for 50 years and all five distances for 60 years were less than 95°C. As a result, the peak temperature at the canister surface decreased as the cooling time and the distance between the deposition holes and the tunnels increased.

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 the 3rd revision of NUREG-0800, which was revised in 2007, the calculation method for decay heat in the design of the Ultimate Heat Sink (UHS) for a pressurized water reactor is recommended to be based on the ANSI/ANS-5.1 method. This method employs a more complex decay heat calculation formula compared to the one introduced in Branch Technical Position ASB 9-2, which was presented in the 2nd revision. While most of the variables for decay heat calculation in ANSI/ANS-5.1 can be inferred from the methods outlined in the appendices, determining the fractions of fission products is not straightforward despite their significant impact on the results. When reviewing documents that evaluate decay heat using the ANSI/ANS-5.1 method, it is observed that they often adopt a conservative approach by assuming that the fraction of the most influential fission product is 100%. In this study, the fractions of each fission product presented in LLNL’s 2016 report were used to calculate decay heat, and the results were compared with the ASB 9-2 method and ORIGEN code results. The comparison showed that ANS 5.1 tends to yield higher decay heat values than ANS 9-2, particularly at the reference time of 1M seconds, while ORIGEN-ARP generally produced lower values. Therefore, it is concluded that even when using the ANSI/ANS-5.1 method with the fractions of each fission product for decay heat calculations in spent nuclear fuel wet or dry storage facility assessments, it provides a sufficiently conservative thermal evaluation.

In general, systems are developed by repeatedly performing the processes of design, analysis, manufacturing, and performance testing. In particular, systems with temperature, pressure, and flow rate often utilize computational fluid dynamics tools at the design stage. In this paper, we aim to verify the reliability of the analysis results of Solidworks Flow Simulation, which is widely used in heat flow analysis at the design stage. A tube furnace was manufactured, various experiments were performed, and a study was conducted to compare the analysis results. The details of the experiment are as follows. First, an experiment was conducted in which the heater was heated to 900°C without insulating the exposed part of the tube. The detailed contents of the experiment are as follows; - Heating heater and measuring temperature without supplying flow inside the tube, - Tube flow supply (25°C, 15 lpm air) and heater heating/temperature measurement. Second, an experiment was performed in which the exposed part of the tube was insulated (thickness 50 mm) and the heater was heated to 900°C. The detailed contents of the experiment are as follows; - Insulate the outside of the tube except for the flanges at both ends of the tube, and heat the heater and measure the temperature without supplying flow inside the tube. - Insulate the outside of the tube except for the flanges at both ends of the tube, supply flow rate inside the tube (25°C, 15 lpm air) and measure heater heating/temperature. - Insulate the flange of the flow supply section, heat the heater and measure temperature without supplying flow inside the tube. - Insulate the flange of the flow supply section, heat the supply air (277°C, 15 lpm) and measure the temperature using a heating gun without heating the heater. - Insulate the flange of the flow supply section, supply heated air (277°C, 15 lpm) and measure heater heating/temperature. - Insulate the flange of the flow supply section and measure temperature according to heater heating (900°C) and supply temperature (25°C, 277°C 15 lpm). The following results were derived from the experimental and analysis results. - When the exposed part of the tube is insulated, the temperature inside the tube increases and the steady-state power decreases compared to non-insulated. - In areas with insulation, the temperature error between experiment and analysis results is not large. - When flow rate is supplied, there is a large temperature error in experiment and analysis results. - The temperature change after the center of the heater is not large for a temperature change of 15 lpm flow rate. From these results, it can be seen that Solidworks Flow Simulation has a significant difference from the experimental results when there is a flow rate in the tube. This was thought to be because the flow rate acts as a disturbance, and this cannot be sufficiently accounted for in the analysis. In the future, we plan to check whether there is a way to solve this problem.

The solid-state chemistry of uranium is essential to the nuclear fuel cycle. Uranyl nitrate is a key compound that is produced at various stages of the nuclear fuel cycle, both in front-end and backend cycles. It is typically formed by dissolving spent nuclear fuel in nitric acid or through a wet conversion process for the preparation of UF6. Additionally, uranium oxides are a primary consideration in the nuclear fuel cycle because they are the most commonly used nuclear fuel in commercial nuclear reactors. Therefore, it is crucial to understand the oxidation and thermal behavior of uranium oxides and uranyl nitrates. Under the ‘2023 Nuclear Global Researcher Training Program for the Back-end Nuclear Fuel Cycle,’ supported by KONICOF, several experiments were conducted at IMRAM (Institute of Multidisciplinary Research for Advanced Materials) at Tohoku University. First, the recovery ratio of uranium was analyzed during the synthesis of uranyl nitrate by dissolving the actual radioisotope, U3O8, in a nitric acid solution. Second, thermogravimetric-differential thermal analysis (TG-DTA) of uranyl nitrate (UO2(NO3)2) and hyper-stoichiometric uranium dioxide (UO2+X) was performed. The enthalpy change was discussed to confirm the mechanism of thermal decomposition of uranyl nitrate under heating conditions and to determine the chemical hydrate form of uranyl nitrate. In the case of UO2+X, the value of ‘x’ was determined through the calculation of weight change data, and the initial form was verified using the phase diagram for the U-O system. Finally, the formation of a few UO2+X compounds was observed with heat treatment of uranyl nitrate and uranium dioxide at different temperature intervals (450°C-600°C). As a result of these studies, a deeper understanding of the thermal and chemical behavior of uranium compounds was achieved. This knowledge is vital for improving the efficiency and safety of nuclear fuel cycle processes and contributes to advancements in nuclear science and technology.

Once discharged, spent nuclear fuel undergoes an initial cooling process within deactivation pools situated at the reactor site. This cooling step is crucial for reducing the fuel’s temperature. Once the heat has sufficiently diminished, two viable options emerge: reprocessing or interim storage. A method known as PUREX, for aqueous nuclear reprocessing, involves a chemical procedure aimed at separating uranium and plutonium from the spent nuclear fuel. This separation not only minimizes waste volume but also facilitates the reuse of the extracted materials as fuel for nuclear reactors. The transformation of uranium oxides through dissolution in nitric acid followed by drying results in uranium taking the form of UO2(NO3)2 + 6H2O, which can then be converted into various solid-state configurations through different heat treatments. This study specifically focuses on investigating the phase transitions of artificially synthesized UO2(NO3)2 + 6H2O subjected to heat treatment at various temperatures (450, 500, 550, 600°C) using X-ray Diffraction (XRD) analysis. Heat treatments were also conducted on UO2 to analyze its phase transformations. Additionally, the study utilized XRD analysis on an unidentified oxidized uranium oxide, UO2+X, and employed lattice parameters and Bragg’s law to ascertain the oxidation state of the unknown sample. To synthesize UO2(NO3)2 + 6H2O, U3O8 powder is first dissolved in a 20% HNO3 solution. The solid UO2(NO3)2 + 6H2O is obtained after drying on a hotplate and is subsequently subjected to heat treatment at temperatures of 450, 500, 550, and 600°C. As the heat treatment temperature increases, the color of the samples transitions from orange to dark green, indicating the formation of different phases at different temperatures. XRD analysis confirms that uranyl nitrate, when heattreated at 500 and 550°C, oxidizes to UO3, while the sample subjected to 600°C heat treatment transforms into U3O8 due to the higher temperature. All samples exhibit sharp crystal peaks in their XRD spectra, except for the one heat-treated at 450°C. In the second experiment, the XRD spectra of the heat-treated UO2 consistently indicate the presence of U3O8 rather than UO3, regardless of the temperature. Under an oxidizing atmosphere within a temperature range of 300 to 700°C, UO2 can be oxidized to form U3O8. In the final experiment, the oxidation state of the unknown UO2+X was determined using Bragg’s law and lattice parameters, revealing that it was a material in which UO2 had been oxidized, resulting in an oxidation state of UO2.24.

This study examined the heat balance in the electrolytic reducer during oxide reduction of pyroprocessing. The adoption of carbon anodes instead of conventional platinum anodes in the oxide reduction process has made it possible to apply high currents, and it has been observed that the temperature of the molten salt of in the reactor rises rapidly when applying high currents, so it is important to maintain an optimal operational temperature range. In this study, salt resistant heat, reaction heat, and decay heat were identified as factors affecting heat balance during the operation of oxide reduction process. Equations describing the relationships among these factors were established. Then using this, a correlation was developed to understand the relationship between applied current and the molten salt temperature in the reactor observed in the actual operation of the carbon anode electrolytic reducer of KAERI. Furthermore, this study proposed strategies to mitigate excessive temperature elevation during oxide reduction operation. A comparative assessment of these approaches was conducted. Considering KAERI electrolytic reducer operation environment, among the considered cooling strategies, the cooling effectiveness was calculated to be highest in the following order: heat transfer to extra salt, convection, conduction, argon gas bubbling.

In electric vehicles, the core is a secondary cell battery. Raw material is pulverized by the grinding disc in the Classifier Separator Mill (CSM) and rises through the Classifier Wheel. Both require characteristics to withstand high-speed rotation, including abrasion, corrosion, and shock. Our study analyzes the impact of RPM and heat source on temperature, convergence, and durability. In conclusion, high heat increases flow, while high RPM reduces the maximum temperature but may harm durability. Proper RPM settings enhance durability.

This study explores strategies to address the variability of heat energy using renewable energy and heat grids to achieve carbon neutrality. Renewable energy source introduce fluctuations that can impact the stability of power systems, while heat grids provide a systematic infrastructure for efficient supply of heat energy and power generation. Jeju Island is chosen as a case study, focusing on balance control based on groundwater-based geothermal energy to optimize the distribution of heat energy. The results demonstrate that the 3rd control method is the most effective in maintaining the target temperature in greenhouses, and specific temperature settings and objectives are proposed for each control method based on crop requirements. These research findings contribute to achieving carbon neutrality and reducing power consumption.

본 연구에서는 강한 내열성을 가지는 비병원성 Bacillus atrophaeus, Bacillus subtilis, Geobacillus stearothermophilus 포자의 열 저항성을 분석하여 레토르트 식품 제조 시 직 접적인 멸균 여부 판정에 사용 가능성을 평가하고자 하였 다. B. subtilis 포자의 D121-value는 2.9±0.1분이었으며, Zvalue는 43.0±1.4oC로 나타났다. G. stearothermophilus 포 자의 D121-value는 4.3±0.1분이었으며, Z-value는 25.0±1.6oC 로 나타났다. B. atrophaeus 포자의 D121-value는 3.7±0.1분 이었으며, Z-value는 35.8±1.4oC로 나타났다. B. subtilis, G stearothermophilus와 B. atrophaeus 포자의 D121-value는 모 두 레토르트 식품 멸균 확인에 사용되는 C. botulinum 포 자의 D121-value 보다 높은 값을 나타내었다. 이러한 결과 를 종합하여 볼 때 레토르트 식품 멸균 시 병원성 포자형 성균인 C. botulinum 대신 B. subtilis, G. stearothermophilus, B. atrophaeus 포자를 사용할 수 있을 것으로 판단된다. 또 한 기존 세균발육 실험에 소요되는 13일보다 단시간인 2- 3일에 멸균 여부를 확인할 수 있을 것으로 판단된다.

본 연구는 사료내 비테인, 글라이신, 그리고 콜린의 혼합 첨가가 고온 스트레스 환경에서 노령 산란계의 생산성, 난품질, 면역 반응 및 혈액성상에 미치는 영향을 조사하고자 수행되었다. 총 336마리의 86주령 로만 갈색종 노령 산란계를 6처리 7반복, 반복당 8수씩 임의 배치하였다. 대조구는 모든 영양소 및 에너지 요구량을 충족하거나 초과하도록 배합하였다. 대조구를 제외한 사료 처리구는 0.2% 비테인, 0.62% 글라이신, 그리고 0.32% 콜린을 단독, 두 가지 혼합, 혹은 세 가지 혼합으로 사료내 첨가하였다. 실험은 8주 동안 진행되었으며, 모든 산란계는 매일 8시간 동안 평균 온도 31.7±1.7℃, 습도 57%의 고온 스트레스 조건에서 사양되었고, 이외 시간에는 평균 온도 27±1.3℃, 습도 57%에서 사양하였다. 실험 결과, 비테인, 글라이신 및 콜린의 첨가는 생산성, 난품질, 그리고 면역 반응에 유의적인 영향을 미치지 않았다. 그러나, 0.2% 비테인과 0.62% 글라이신을 혼합 첨가한 처리구에서 혈청 알라닌 아미노전이효소 농도가 유의적으로 감소했다. 하지만, 다른 혈청 지표들은 처리간 유의적인 차이가 관찰되지 않았다. 결론적으로, 현재 수준에서 사료내 비테인, 글라이신, 그리고 콜린의 혼합 첨가는 고온 스트레스 환경에서 사양되는 노령 산란계의 생산성, 난품질, 면역 반응 및 혈액 성상에 긍정적인 영향을 미치지 않는다고 판단된다.

긴털가루진드기(Tyrophagus putrescentiae)는 주로 실내에 서식하며 알레르기 및 호흡기 질환을 일으키는 주 요 해충이다. 주로 침구류에서 발생하며, 사람이나 동물로부터 떨어진 각질과 비듬 또는 저장 곡식 등을 주요 먹이원으로 한다. 암컷 한 마리가 일생에 200~300개의 알을 산란하기 때문에 빠르게 번식할 수 있으며, 육안으로 확인할 수 없는 크기이기 때문에 관찰 및 관리가 불가능하다. 하지만 긴털가루진드기로 인해 발생하는 알레르기 에 관한 연구는 환자의 치료에 초점이 맞춰져 있기 때문에, 근본적으로 알레르기를 발생시키는 원인을 해결할 수 있는 방안이 부족한 상황이다. 본 연구에서는 가정에서 발생할 수 있는 긴털가루진드기를 효과적으로 방제하 기 위한 통합적 방제 전략을 제시한다. 열, 살비제(Chlorfenapyr, Spiromesifen, Pyflubumide), 초음파, Essential oil(Lemongrass, Peppermint)을 사용하여 살비 및 기피 효과를 확인하였다. 고온 노출은 100% 살비 효과를 보여주 었으며, 살비제는 3시간 동안 90% 이상의 살비 효과를 보여주었다. 긴털가루진드기가 초음파에 노출되었을 때 23%의 기피 행동을 관찰하였으며, Essential oil(Lemongrass, Peppermint)을 유인용 먹이에 처리했을 때, 긴털가루 진드기 97%의 접근을 차단하였다. 본 연구를 통해 실내 알레르기의 가장 중요한 원인으로 주목받는 긴털가루진 드기에 대한 체계적인 방제 전략을 적절한 환경에 적용시킴으로써, 집먼지진드기로 인한 알레르기 및 호흡기 질환 발생을 감소시키고 안전한 주거 환경을 조성할 것으로 기대한다.

Hydrogen is considered as one of the most promising future energy carriers due to its noteworthy advantages of renewable and high calorific value. The long-term storage of liquid hydrogen with low heat leakage is essential for future deep space exploration. Because of low critical temperature and volatility, liquid hydrogen tank poses severe requirements to multi-layer insulation (MLI). In order to reduce heat leak into tank, vapor cooled shield (VCS) was set up to cool MLI by retrieving the heat of discharged cryogenic gas hydrogen. This paper presents an parametric study on insulation system in liquid hydrogen storage vessel with MLI and VCS. Thermal model was developed, and heat transfer analysis by varying VCS position was conducted. Temperature and heat flux distributions along time passing were derived, and effect of VCS position on insulation performance was investigated.