The selection of fnancial assets within households has gained increasing prominence in the context of China’s economic development. This paper systematically reviews existing literature, both domestic and international, focusing on family heterogeneity and the infuence of external factors. It critically assesses the advancements made in understanding the determinants of family fnancial asset choices while also acknowledging the limitations in current research. Building upon this comprehensive review, the paper identifes three promising avenues for future exploration. Firstly, there is a need for research delving into the impact of the digital economic environment on family fnancial asset choices. Secondly, understanding how macroeconomic systems shape these decisions is crucial for a holistic perspective. Lastly, examining the heterogeneity among family members and its influence on financial asset choices offers a nuanced understanding of this intricate decision-making process. The insights presented in this paper are invaluable for both academic research and practical applications related to family fnancial asset selection.

After the major radioactivation structures (RPV, Core, SG, etc.) due to neutron irradiation from the nuclear fuel in the reactor are permanently shut down, numerous nuclides that emit alpha-rays, beta-rays, gamma-rays, etc. exist within the radioactive structures. In this study, nuclides were selected to evaluate the source term for worker exposure management (external exposure) at the time of decommissioning. The selection of nuclides was derived by sequentially considering the four steps. In the first stage, the classification of isotopes of major nuclides generated from the radiation of fission products, neutron-radiated products, coolant-induced corrosion products, and other impurities was considered as a step to select evaluation nuclides in major primary system structures. As a second step, in order to select the major radionuclides to be considered at the time of decommissioning, it is necessary to select the nuclides considering their half-life. Considering this, nuclides that were less than 5 years after permanent suspension were excluded. As a third step, since the purpose of reducing worker exposure during decommissioning is significant, nuclides that emit gamma rays when decaying were selected. As a final step, it is a material made by radiation from the fuel rod of the reactor and is often a fission product found in the event of a Severe accident at a nuclear power plant, and is excluded from the nuclide for evaluation at the time of decommissioning is excluded. The final selected Co-60 is a nuclide that emits high-energy gamma rays and was classified as a major nuclide that affects the reduction of radiation exposure to decommissioning workers. In the future, based on the nuclide selection results derived from this study, we plan to study the evaluation of worker radiation exposure from crud to decommissioning workers by deriving evaluation results of crud and radioactive source terms within the reactor core.

In Natural Analogue Study, Concrete is one of the important engineering barrier components in the Multi-thin wall concept of radioactive waste disposal and plays the most important role in disposal sites. The concrete barrier at the disposal site loses its role as a barrier due to various deterioration phenomena such as settlement, earthquake, and ground movement, causing the disposed waste to leak into the natural ecosystem. Some of the key factor is deterioration triggered by sulfate attack. Concrete deterioration induced by sulfate is commonly manifested in an extensive scale when a concrete structure makes contact with soil or water, aggravating its performance. In this study, an accelerated concrete deterioration evaluation experiment was performed using a total of three experimental methods to evaluate the reaction between concrete and water. The first experiment was a deterioration evaluation using Demi. Water, the second was a deterioration evaluation using KURT groundwater after extraction, and the last experiment was a concrete deterioration evaluation using KURT groundwater and sodium sulfate. For all of these experiments, accelerated concrete deterioration experiments were conducted after immersion for a total of 365 days, and specimens were taken out at 30-day intervals and characterization analysis such as SEM and EDS was performed. Experimental analyzes have shown that various chemical species are generated or destroyed over time. In the future, we plan to continue to conduct a total of three concrete deterioration evaluation experiments above, and additionally evaluate the chemical reaction between bentonite and concrete.

This study investigates the behavior of bentonite, used as a buffer material in deep geological disposal systems, in the context of pore morphology under the influence of field-collected groundwater conditions. The bentonite was processed into block form using cold isostatic press (CIP) and subsequently analyzed for its pore morphology in situ using synchrotron X-ray computed tomography (CT) within the field-collected groundwater environment. Bentonite buffers play a critical role in deep geological disposal systems by preventing contact between disposal containers and groundwater. Bentonite typically exhibits swelling upon contact with water, forming few layers of water molecules between its structural layers. However, the presence of ions such as K+ and Cl- can lead to a sharp reduction in swelling pressure. Loss of swelling pressure could negatively impact the integrity of future deep geological disposal systems, making its assessment crucial. This study involves processing various types of bentonite, including natural Na-type bentonite, into block forms and subjecting them to exposure in both deionized water and field-collected groundwater conditions. Internal pore morphology changes were measured using Xray CT technology.

The development of the Internet and digital technology has given rise to a new form of cultural industry - digital literacy, and the trend of all-media convergence has also promoted the production and marketing of IP, and the space for the development of digital literacy under the IP strategy and the market demand are constantly expanding. Due to the problems of insufficient exploitation and utilisation, low technology level, content homogenisation and serious piracy of China's IPs, the feasible paths of building a digital cultural and creative industry chain, all-media marketing, establishing a digital copyright protection system, guiding the orderly competition of IPs, and carrying out international cooperation have been put forward with the combination of theories and practices, to improve the efficiency of IP production, content quality and digitalisation, stimulate the innovation of content and science and technology, protect digital copyrights, and provide a better service for China's high-quality digital copyright production and marketing. Protection of digital copyright, and provide a reference for the creation of high-quality digital cultural and creative IP in China.

Zinc injection into the coolant system of nuclear power plants is an effective method for reducing corrosion and improving performance. The effectiveness of this method is influenced by various factors such as zinc concentration and injection rate. This paper provides an overview of the factors affecting the effectiveness of zinc injection in nuclear power plants, with a focus on zinc concentration and injection rate, and discusses various research results on the effects of these factors on corrosion reduction and coolant system performance. Zinc concentration is an important factor affecting the effectiveness of zinc injection. The research results show that gradual increases in zinc concentration are more effective for coolant system stability. However, the concentration should not exceed the recommended levels as high zinc concentrations can have negative effects on the system. Injection rate is also an important factor affecting the effectiveness of this method. The research results show that gradual increases in injection rate are more effective for coolant system stability. However, excessive injection rates can have negative effects on the system such as overload of the zinc injection facility and chemical shocks within the coolant system, and therefore, should be optimized. In conclusion, zinc concentration and injection rate are important factors affecting the effectiveness of zinc injection in nuclear power plants. The optimal concentration and injection rate should be determined based on specific reactor conditions and system requirements, and efforts should be made to maximize corrosion reduction and performance improvement.

Korean innovative SMR has been implemented developing with improved safety/economy and i- SMR technology development project to secure a competitive edge in SMR. For nuclear power plants, according to the revision of the Nuclear Safety Act (2013.6), it is mandatory to be reflected in the aging management program of nuclear power plants, and the aging management and regulation of major nuclear power plants are being strengthened. For i-SMR, chemistry environment and management strategy is essential to mitigate corrosion and radiation fields, since it has compacted and integrated module designs. Since 1994, zinc injection into the reactor coolant system (RCS) has been applied more than 100 PWRs in the world to mitigate primary water stress corrosion cracking (PWSCC) and to reduce outof- core radiation fields. In domestic NPPs, 7 have been applying zinc injection and had up to 90% radiation field reductions. For this reason, SMR needs to apply zinc injection for chemistry strategy. Zinc target concentration will be 5~40 ppb at i-SMR, based on Ni-Fe-Cr materials as same as PWRs. Zinc injection location is in volume and purification control system between the volume control tank and charging P/P where the pressure is moderate. Zinc injection skid can consist of two micro-controllable pump (one for operation and one for stand-by) and one injection tank (batching tank for zinc solution). Zn, Ni, Si, Fe, and activated corrosion products should be monitored to identify zinc injection controls and trends. Flux mapping for core performance monitoring should be evaluated. The application of zinc will be essential and effective and bring sustainable reliability for corrosion control and mitigation strategy to meet the risk-free i-SMR development.

Pressurized Heavy Water Reactors (PHWR) have stored ion exchange resins, which are used in deuteration, dehydrogenation systems, liquid waste treatment systems, and heavy water cleaning systems, in spent resin storage tanks. The C-14 radioactivity concentration of PHWR spent resin currently stored at the Wolseong Nuclear Power Plant is 4.6×10E+6 Bq/g, which exceeds the limited concentration of low-level radioactive waste. In addition, when all is disposed of, the total radioactivity of C-14, 1.48×10E+15 Bq, exceeds the disposal limit of the first-stage disposal facility, 3.04×10E+14. Therefore, it is currently impossible to dispose of them in Gyeongju intermediate- and low-level disposal facilities. As to dispose of spent resins produced in PHWR, C-14 must be removed from spent resins. This C- 14 removal technology from the spent resin can increase the utilization of Gyeongju intermediate- and low-level disposal facilities, and since C-14 separated from the spent resin can be used as an expensive resource, it is necessary to maximize its economic value by recycling it. The development of C-14 removal technology from the spent resin was carried out under the supervision of Korea Hydro & Nuclear Power in 2003, but there was a limit to the C-14 removal and adsorption technology and process. After that, Sunkwang T&S, Korea Atomic Energy Research Institute, and Ulsan Institute of Science and Technology developed spent resin treatment technology with C-14-containing heavy water for the first and second phases from 2015 to 2019 and from 2019 to the present, respectively. The first study had a limitation of a pilot device with a treatment capacity of 10L per day, and the second study was insufficient in implementing the technology to separate spent resin from the mixture, and it was difficult to install on-site due to the enlarged equipment scale. The technology to be proposed in this paper overcomes the limitations of spent resin mixture separation and equipment size, which are the disadvantages of the existing technology. In addition, since 14CO2 with high concentration is stored in liquid form in the storage tank, only the necessary amount of C-14 radioactive isotope can be extracted from the storage tank and be used in necessary industrial fields such as labeling compound production. Therefore, when the facility proposed in this paper is applied for treating mixtures in spent resin tanks of PHWR, it is expected to secure field applicability and safety, and to reflect the various needs of consumers of labeled compound operators utilizing C-14.

Radioactive waste generated during nuclear power plant decommissioning is classified as radioactive waste before the concentration is identified, but more than 90% of the amount generated is at a level that can be by clearance. However, due to a problem in the analysis procedure, the analysis is not carried out at the place of on-site but is transported to an external institution to identify concentration, which implies a problem of human error because 100% manual. As a way to solve this problem, research is underway to develop a mobile radioactive waste nuclide analysis facility. The mobile radionuclide analysis facility consists of a preparation room, a sample storage room, a measurement room, a pretreatment room, and a waste storage room, and is connected to an external ventilation facility. In addition, since the automation module is built-in from the sample pre-threatening step to the separation step, safety can be improved and rapid analysis can be performed by being located in the decommissioning site. As an initial study for the introduction of a mobile nuclide analysis facility, Visiplan, a preliminary external exposure evaluation code, was used to derive the analysis workload by a single process and evaluate the exposure dose of workers. Based on this, as a follow-up study, the amount of analysis work according to the continuous process and the exposure dose of workers were evaluated. As a result of the evaluation, the Regulatory dose limit was satisfied, and in future studies, internal and external exposure doses were evaluated in consideration of the route of movement, and it is intended to be used as basic data in the field introduction process.

Among the twenty six nuclear power plants in Korea, twenty four plants are currently in operation excluding the two permanently shut-down Kori #1 and Wolsung #1 plants. The decommissioning process includes many tasks such as cutting, decontamination, disposal and treatment. Among the tasks, because cutting is one of the tasks performed close to the target structure, there is a possibility for the workers to be exposed excessively to the radiation. There are representative large structures such as steam generators, nuclear reactors, reactor coolant pump, and pressurizer, made of metals, and radioactive concrete, made of concrete. Especially, compared to the trend of research to manage the radiation exposure of steam generators that are directly connected to pressurizers, the trend of research to manage the radiation exposure of pressurizers to workers is not satisfactory. Moreover, although there have been many studies on radioactive concrete, the studies to manage the radiation exposure to workers with a systematic cutting scenario are insufficient. In this study, radioactive concrete, a representative large structure made of concrete, was selected as the target for evaluation. The conditions for evaluation were cutting speed (1~10 m2/hr) and the time for cutting (permanent shutdown~30 years after the shutdown). A cutting scenario was developed by applying the situation for abrasive decontamination beforehand and Hot-to-Cold and Cold-to-Hot, and effort was made to derive a reasonable plan. The evaluation result derived were hourly radiation dose distribution of 1.19~0.103 mSv/hour and 1.29~0.0113 mSv/hour for a scenario without abrasive decontamination (in the order of Hot to Cold, Cold to Hot), and hourly radiation dose distribution of 0.547~0.0479 mSv/hour and 0.608~0.0522 mSv/hour for a scenario with abrasive decontamination. The maximum value of collective dose derived was 1.54E+04 mSv at the cutting time of permanent shutdown with cutting speed of 1 m2/hour in the Cold to Hot scenario before abrasive decontamination, and the minimum value derived was 5.15E+01 mSv at the cutting time of 30-year after the permanent shutdown with cutting speed of 10 m2/hour in the Hot to Cold scenario after abrasive decontamination.

Radioactive waste must be stored for at least 300 years and must bear astronomical costs. In addition, unexpected potential risk factors are also a considerable burden. In the case of low-level radioactive waste, combustible and liquid low-contamination radioactive waste can be treated relatively easily through high-temperature plasma which the volume can be reduced by 1/250 and the weight by 1/30. It is possible to permanently dispose of the ash leached after plasma treatment in a more stable manner compared to the conventional methods. Types of low-level combustible radioactive waste, including paper, vinyl, clothing, filters, and resins, account for more than 30% of the total waste volume. Furthermore, high-temperature plasma treatment of low-level radioactive waste from petrochemical plants and medical institutions have many advantages, namely astronomical cost savings, securing free space in existing storage facilities, and improving the image of nuclear energy. Korea is preparing to decommission the Kori No. 1 nuclear power plant, and small and mediumsized enterprises and related organizations are conducting various studies to incinerate radioactive waste. In foreign countries, Britain began incineration technology in the 1970s, and Plasma Energy Group, LLC, headquartered in Florida, USA, physically changed the molecular structure of the material by combining plasma chambers and plasma arcs and obtained a patent application in 1992. Germany was approved for operation in 2002, and Switzerland completed a trial run of a plasma technologybased facility in 2004. Important radionuclides in terms of radioactive gas waste treatment include inert gases, radioiodine, and radioactive suspended particles. Gas waste is compressed in a compressor through a surge tank in the gas waste treatment system and filters at each stage. after that, the shortlife nuclide is naturally collapsed for 30 to 60 days in the storage or activated carbon adsorbent in the attenuation tank and released through HEPA filters. The radioactive concentration at discharge is monitored and managed using continuous monitoring equipment, and the oxygen concentration is managed in the gas waste treatment system to prevent explosion risk. The problem of radioactive waste disposal is not only a problem for people living in the present era, but also a big social issue that brings a burden to future generations While interest in plasma treatment is increasing from the decommissioning of the Kori Unit 1. in Korea, it is showed that there is a lack of systematic management and research especially in the radioactive volatile gases fields, that’s why I propose some ideas as follows. First, the government and related institutions should invest to the continuous radioactive monitoring system to produce and distribute continuous radioactive monitoring facilities with an affordable price. Second, it is recommended that radioactive waste incineration would be connected to the GRS system of the plant’s gas radwaste treatment system, and radioactive volatile materials should be monitored through continuous monitoring system. Third, radioactive volatile materials generated according to the temperatures and times during plasma incineration treatment are different. Therefore, prior classification of each expected radioactive volatile substance must be performed before incineration.

When decommissioning and operating nuclear power plants, a lot of radioactive waste in concentrated waste powder, slurry, sludge, and powder is generated. The radioactive waste, non-conformity for disposal, cannot be treated or disposed of, but is currently being stored instead. To dispose of the waste, the waste can be solidified by mixing with an appropriate solidification agent. However, when the solidification agent and powder particles are mixed as in the conventional method, the final volume of the waste form to be disposed of increases. In order to solve this problem, in this study, volume reduction was achieved, compared to the existing powder, by applying the roll compaction technology to mold the radioactive waste into compressed pellets. Soil, concrete, concentrate waste, and contaminated soil powder were used as test materials, and pellets were prepared under different operating conditions. Subsequently, a compressive strength test was performed to confirm the integrity and optimal process conditions of the manufactured pellets. However, in order to perform the compressive strength test, the upper and lower surfaces of the pellets must be horizontal, but the pellet has the shape of two tetrahedrons joined together. Hence, test specimens for measuring compressive strength were prepared by making a surface treatment jig. The compressive strength test showed a high strength of 5.20~28.20 MPa. The process conditions showing high compressive strengths were selected as the optimal process conditions. Finally, the volume reduction ratios were calculated by measuring the weight, density and volume of the manufactured pellets. The degrees of volume reduction of the manufactured pellets compared to the existing powder were checked. When the roll gap was 0 mm, the average reduction ratios of the test materials were 3.7 for the soil, 4.0 for the concrete, 4.6 for the concentrate waste, and 3.8 for the contaminated soil. When roll gap was 1 mm, the ratios were 2.7 for the soil, 2.9 for the concrete, 3.4 for the concentrate waste, and 2.8 for the contaminated soil. Therefore, from a conservative point of view (Roll gap = 1 mm), when powdered waste is formed into pellets, it means that the volume is reduced by 1/2.7 for soil, 1/2.9 for concrete, 1/3.4 for concentrated waste, and 1/2.8 for contaminated soil.