Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SFs with their high radiotoxicity, heat generating, and long-lived radioactivity are actually the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own methodologies and management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SF has also been conducted during last couple of decades since around 1997, during which several various alternative type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide. The world first Finnish repository which is expected to begin to operate sooner or later will be also this type. Since the characteristics of SF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development has been being roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in nearfield, far-field, and even biosphere in and around the conceptual repository system. Through the study such scenarios and models has been implemented to development of a safety assessment by utilizing GoldSim development tool for a rough quantitative comparison with existing disposal options and simple illustration purpose as well as for showing how to develop and implementation of the model to GoldSim templet.

Regulatory agencies require burn-up verification to ensure that dry storage casks using burn-up credit are not loaded with fuel with a reactivity greater than the allowable standard. Accordingly, in preparation for dry storage of SF, the reliability of the burnup was verified and action plans for fuel with confirmed errors were reviewed. Reliability verification was performed by comparing the actual burnup calculated with combustion calculation code (TOTE, ISOTIN) used in NPP and the design burnup calculated with the nuclear design code (ANC). As a result of comparing the differences between actual burnup and design burnup for 7,414 assemblies of SF generated from CE-type NPPs, the average deviation was confirmed to be 0.79% and 220 MWD/MTU. In the CE-type NPPs, no fuel showing large deviations was identified, and it was confirmed that reliability was secured. As a result of comparing the differences in 11,082 assemblies of SF generated from WH-type NPPs, the differences were not large, averaging 1.16% or 422 MWD/MTU. However, fuels showing significant differences were identified, and cause analysis was performed for those fuels. The cause analysis used a method of comparing the burnup of symmetrically loaded fuels in the reactor. For fuels that were not symmetrically loaded, a method was used to compare them with fuels with similar combustion histories. As a result of the review, it was confirmed that the fuel was under- or over-burned compared to symmetrically loaded fuel. For fuels for which clear errors have been identified, we are considering replacing them with the design burnup, and for fuels whose causes cannot be confirmed, we are considering ways to maintain the actual burnup.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

The utilization of methyl bromide (MB) for quarantine purposes has been hampered by its designation as an ozone-depleting substance under the Montreal Protocol. The International Plant Protection Convention's (IPPC) call for alternatives to MB and a reduction in its usage. There is an urgent need to explore and implement substitutes. Despite some substitute agents like EDN being developed for wood, EDN has been limited due to various factors such as occupational risks. This study focuses on evaluating the efficacy of Sulfuryl Fluoride (SF) as a viable alternative fumigant against Reticulitermes speratus, one of major wood destroying pests. Experimental trials conducted at ambient temperature (23°C) revealed promising results, with SF demonstrating LCT50 and LCT99 values of 30.87 mg·h/L and 42.53 mg·h/L, respectively. Under low-temperature conditions (5°C), SF remained effective but with slightly higher LCT50 and LCT99 values of 151.62 mg·h/L and 401.90 mg·h/L, respectively. The penetration test, conducted using R. speratus-infested pine wood cubes, further highlighted SF's efficacy, with LCT50 and LCT99 values of 31.59 mg·h/L and 53.34mg·h/L at 23°C, indicating powerful penetration capabilities. When tested at a loading ratio of 90% (v/v) at 5.0mg/L for 24 hours in a 500L chamber as a middle-scale trial, SF achieved a 100% mortality, showing its potential as a suitable replacement for MB. These findings suggest that SF could open new markets as an MB substitute and enhance safety at quarantine sites when applied to imported and exported timber.

Integrity evaluation scheme for Spent Fuel (SF) dry storage has been developed under transportation failure modes. This method especially considered the degradation characteristics of Spent Fuel (SF) during dry storage such as radial and circumferential hydride content, hydride volume fraction, oxide thickness, etc. Hydride and zircaloy cladding are considered as material composite system, using correlation models related to material properties. Critical Strain Energy Density (CSED) is compared with Strain Energy Density (SED), to evaluate cladding integrity. CSED serves as material characteristics, while SED can be considered as boundary condition. To calculate the CSED of cladding in the lateral failure mode, circumferential hydride concentration is used. SED is calculated considering both the bending moment and axial load. On the other hand, in the longitudinal failure case, fuel rod temperature, internal pressure, hoop stress, radial hydride concentration is used to calculate CSED. And pinch force (contact) was considered to evaluate SED. Model validations were conducted by comparing hot cell SF test and existing validated evaluation results. To separately handle normal transportation conditions from hypothetical accident conditions, SED according to stress-strain analysis results was separated into elastic and plastic regions. As a result of applying this scheme for 14×14 SF, failure probability of normal condition was zero, which is the similar result with DOE and same with EPRI. Regarding accident condition, lateral case showed similar result, but longitudinal case showed different but reasonable result, which was due to the different analysis conditions. The proposed methodology which was indigenously developed through this study is named as K-method.

Data commentary is an important text type in research articles; however, its discourse model is often challenging to access because it is embedded in the upper genres such as textbook, weather forecast, and journal article. This study aims to establish a discourse model of data commentary, with a focus on academic research papers in Economics and Business administration journals. To accomplish this, this study employs Move analysis and SF-MDA(Systemic Functional-Multimodal Discourse Analysis) to investigate the moves of data commentaries and the metafunctional meanings of each step. The results indicate that the data commentary discourse model consists of three moves: (1) summarizing the topic and methodology, (2) representing figure and numbers, and (3) analyzing and commenting on results. Additionally, 22 steps are identified for each move that creates metafunctional meaning: ideational, interpersonal, and textual.

The saturation rates of the spent fuel (SF) wet storage at the Kori Nuclear Power Plant (NPP), Hanbit, and Hanul are 83.3%, 74.2%, and 80.8% as of the fourth quarter of 2021. The storages of Kori NPP and Hanbit NPP are expected to be saturated in 2031, and Hanul is expected to be saturated in 2032. Therefore, the construction of an interim storage facility to store the SF temporarily stored in the NPP was planned, and preparations for the safe transport of the SF are required. In this paper, radiological preliminary assessment using NRC-RADTRAN in the process of sea transport of SF from the wet storage or ISFSI of the Hanbit NPP to the optional interim storage facility was performed. Since domestic SF transport vessels are not currently in operation, the specifications of the UK Pacific Grebe vessel which can carry up to 20 casks were used. The transport cask used the specifications of KORAD-21, a transport container developed in Korea. Because it can carry more SF assemblies than the existing KN-18. In addition, a land transport safety test was conducted in 2020 and a sea transport test is planned. The sea transport route was entered by referring to the transport route of domestic low and intermediate level waste. The accidents rate was calculated using statistics on maritime accidents from 2017 to 2021. The probability accidents along the transportation route were evaluated as 3.152E -10. When transporting to an interim storage facility, the SF expected to be the main transport target was selected as WH 17X17, combustion 45,000 MWD/MTU, and concentration of 4.5%. The source term was calculated and entered according to this data and the release fraction was entered with reference to the DOE report. In the case of normal transport without accident, the individual dose of the crew member and public residents were estimated to be 0.0525% and 0.000492% of the annual limit of 1 mSv/yr for the general public. Under the accident conditions, the annual individual doses of residents were 0.0011%, 0.0023%, 0.0034%, and 0.0046% of the annual limit of 1 mSv/yr when carrying 5, 10, 15, and 20 casks. Currently, the site of the interim storage facility has not been precisely determined, but a preliminary radiation assessment through sea transport resulted in a significantly lower than the limit. Combined scenario sea transport followed by land transport will be carried out in the next stage of study.

Some Spent Fuel Pools (SFPs) will be full of Spent Nuclear Fuels (SNFs) within several years. Because of this reason, transporting the SNF from SFP to interim storage facilities or permanent disposal facilities should be considered. There are two ways to transport the SNF from a site to other site, one is the land transportation with truck or train, and the other is the maritime transportation with ship. The maritime transportation has some advantages compared with the land transportation. The maritime transportation method uses safer route which is far from populated area than land transportation method, and transport more weight than land transportation method. However, the cask should be loaded into the ship for the maritime transportation, and there is a possibility of a drop accident of the cask onto the ship. Therefore, it is necessary to evaluate the structural integrity of the cask and ship for the drop accident during the loading process. To evaluate the structural integrity of the cask and ship, it is necessary to determine the analysis conditions that caused the greatest damage in the drop accident. There may be various conditions such as the drop angle of the cask, the initial falling speed, the drop position onto the ship, the size of the ship, etc. This study set the drop angle of the cask and the drop position onto the ship as the simulation variables, which have high possibility to occur during cask drop. However, the others are excluded since they are controllable by worker. In this paper, various drop angle (0, 15, 30, 45, and 70 degree) of the cask were simulated to define the greatest damage condition. KORAD-21 cask model was used for Finite Element Analysis (FEA), and FEA was performed to simulate a horizontal drop (1 m drop). The strain-hardening material properties for the deck were used as HT36 steel. The Cowper-Symonds constitutive model for HT36 was used to consider the strain rate effect. A Tie-down structure for supporting the cask was modeled with the cask model which contained inner structures like canister, basket, etc. Structural integrity of the cask and tie-down structure were evaluated using the von-Mises stress and equivalent plastic strain (PEEQ), and one of the ship deck was evaluated using deflection of ship deck and equivalent plastic strain. Compared with each cask drop angle conditions, 45 degree of the cask drop angle showed the highest deflection and PEEQ values, but did not exceed ultimate strain of HT36. In the ship deck, the corner of deck showed the highest PEEQ value in all simulation cases. As the result, the 45 degree of the cask drop angle condition results was more conservative than other conditions, and the corners of deck failure was able to evaluate ship safety.