In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHM−1 burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

Code-compliant seismic design should be essentially applied to realize the so-called emulative performance of precast concrete (PC) lateral force-resisting systems, and this study developed simple procedures to design precast industrial buildings with intermediate precast bearing wall systems considering both the effect of seismic and blast loads. Seismic design provisions specified in ACI 318 and ASCE 7 can be directly adopted, for which the so-called 1.5S y condition is addressed in PC wall-to-wall and wall-to-base connections. Various coupling options were considered and addressed in the seismic design of wall-to-wall connections for the longitudinal and transverse design directions to secure optimized performance and better economic feasibility. On the other hand, two possible methods were adopted in blast analysis: 1) Equivalent static analysis (ESA) based on the simplified graphic method and 2) Incremental dynamic time-history analysis (IDTHA). The ESA is physically austere to use in practice for a typical industrial PC-bearing wall system. Still, it showed an overestimating trend in terms of the lateral deformation. The coupling action between precast wall segments appears to be inevitably required due to substantially large blast loads compared to seismic loads with increasing blast risk levels. Even with the coupled-precast shear walls, the design outcome obtained from the ESA method might not be entirely satisfactory to the drift criteria presented by the ASCE Blast Design Manual. This drawback can be overcome by addressing the IDTHA method, where all the design criteria were fully satisfied with precast shear walls’ non-coupling and group-coupling strength, where each individual or grouped shear fence was designed to possess 1.5S y for the seismic design.

국내 중·저준위 방사성폐기물은 영구적 격리를 위해 처분장에 매립하고 있으며 그 위치는 경주에 있다. 이러한 방 사성폐기물의 영구적인 격리를 위한 처분시설은 공학적 방벽과 자연 방벽으로 구성되어 있으며 자연 방벽을 특성을 파악 하기 위하여 한국원자력환경공단에서는 2006년부터 부지특성조사를 수행하였고, 이후 부지감시 및 조사계획에 따른 감시 를 수행하여 부지특성의 변화를 지속적으로 확인하고 있다. 중저준위 방폐장의 수리지화학적 환경은 자연 방벽의 평가를 위해 중요한 요소로 손꼽히고 있으나 동해와 가까운 경주의 지역적 특성상 해수의 영향을 반드시 고려해야 한다. 따라서 본 연구에서는 처분 부지의 지하수 관정 7개 및 관정의 심도별 수질 자료를 취합해 지하수 자료 총 30개를 해수 2개소와 비교 분석하여 수리지화학적 환경을 해석하였다. 분석 자료는 수질 10개 항목(온도, EC, HCO3, Na, K, Ca, Mg, Cl, SO4, SiO2)을 2017년 3분기부터 2022년 3분기까지 총 5년간 20회의 자료를 활용하였다. 특히, EC, HCO3, Na, Cl의 농도 변화 를 통해 연구 지역의 배경 농도 및 관정의 구간별 해수의 영향을 파악하였으며, 시계열 군집 분석을 통해 담수, 기수, 해 수의 분류를 시도하였다. 그 결과, 기존의 모니터링 방법으로는 확인하지 못한 부지내 수리지화학적 변화를 제시하였다.

In Korea, extensive industry-academia-research research has already established many facilities and technologies for materials and chemical experiments on non-radioactive substances. However, few facilities have been built to analyze the physical and chemical properties of substances irradiated through neutron irradiation. Korea is planning to decommission Kori-1 and Wolsong-1 in 2027. Extensive analysis of low-level and intermediate-level materials is required to begin decommissioning these nuclear power plants. The material’s composition and level can be identified by analyzing the structure’s characteristics, and a cutting and decontamination plan can be established based on this. In addition, by conducting a nuclide analysis on the waste generated after cutting, suitability for disposal can be secured, and stable treatment can be performed. Accordingly, the Korea Decommissioning Research Institute (KRID) plans to secure infrastructure (hot cells) to analyze the characteristics of intermediate-level decommissioning waste. The goal is to secure high-dose/high-radiation decommissioning waste processing technology through Korea’s first intermediate-level hot cell, support domestic nuclear power plant decommissioning projects, and secure and verify procedures related to nuclide analysis of intermediate-level using hot cells. In addition, by possessing these material properties and nuclide analysis technology, KRID can have a foundation to conduct continuous research on low- and intermediate-level radioactive materials and decommissioning. The purpose of KRID’s establishment is to use this foundation in the future to improve the technological level of the nuclear industry as a whole through linkage between industry, academia, and research institutes.

우리나라 열병합 발전소에서 운영되고 있는 최신 증기터빈의 출력과 효율 향상을 위한 첫 번째 기술적인 진보는 고온, 고압의 증기를 사용할 수 있는 소재 개발의 진척이라고 할 수 있다. 소재의 발전과 더불어 증기터빈의 내부효율 향상을 위한 설계적 노력의 결실 로 높은 효율의 증기터빈이 제작되었다. 오랜 기간 운전 중인 증기터빈의 내부효율은 기계적 수명의 한계로 점차 손실이 발생하고 효율 과 출력이 떨어지게 된다. 이러한 이유로 본 연구에서는 상용프로그램을 이용하여 열병합 발전소용 고압(HP)-중압(IP) 증기터빈의 증기유 로 성능해석을 수행할 수 있는 모델을 개발하고 성능계산 방법을 제시하고자 한다. 증기터빈의 복잡한 성능계산방식으로 인해 증기터빈 실무자들에게 실질적으로 유용한 참고문헌이 될 수 있도록 주요 변수들을 제시하였다. 또한 증기터빈 성능계산에 필요한 열정산도 분석 과 증기터빈 성능계산 결과의 적합성을 성능시험 결과와 비교 확인하였다.

In Korea, many characteristic component facilities and technologies in general experimental areas for non-radiative materials are owned by industry-academia research. Still, no characteristic analysis test technology has been developed for large, intermediate-level decommissioning waste emitted by neutron irradiation. Since Korea plans to decommission nuclear power plants in 2027, securing analysis technology for intermediate-level decommissioning waste is essential. Accordingly, the Korea Research Institute of Decommissioning (KRID) plans to secure an infrastructure (hot cell) to analyze the characteristics of intermediate-level dismantled waste. Afterward, we intend to stably dispose of the waste generated while decommissioning the current Gori Unit 1/Wolseong Unit 1 using the intermediatelevel dedicated hot cell. It aims to secure high-dose/high-radiation decommissioning waste handling technology through intermediate-level hot cells for the first time in Korea, supports domestic nucleardecommissioning projects, and secure and validate procedures related to material characteristics and nuclide analysis of intermediate-level waste. Furthermore, research on intermediate-level radioactive materials is expected to be carried out in cooperation with schools and research institutes.

This study presents a dry precast concrete (PC) beam-column connection, and its target seismic performance level is set to be emulative to the reinforced concrete (RC) intermediate moment resisting frame system specified in ACI 318 and ASCE 7. The key features include self-sustaining ability during construction with the dry mechanical splicing method, enabling emulative connection performances and better constructability. Test specimens with code-compliant seismic details were fabricated and tested under reversed cyclic loading, which included a PC beam-column connection specimen with dry connections and an RC control specimen. The test results showed that all the specimens failed in a similar failure mode due to plastic deformations in beam members, while the hysteretic response curve of the PC specimen showed comparable and emulative performances compared to the RC specimen. Seismic performance evaluation was quantitatively addressed, and on this basis, it confirmed that the presented system can fully satisfy all the required performance for the intermediate RC moment resisting frame.

The intermediate shaft of sliding type is assembled with coated shaft joint and tube joint. Since the intermediate shaft plays a role of absorbing displacement change due to vibration, the intermediate shaft must have a sliding force value in an appropriate range. In this study, an intermediate shaft assembly system for post-processing of defective intermediate shafts was developed. The intermediate shaft assembly system consists of a wear count prediction model and an automatic wear system. A wear count prediction model was created with the initial assembly sliding force, quality, and set values. As a result of applying the intermediate shaft assembly device, the sliding force of the intermediate shaft was induced within the set value range. And it was prevented from the intermediate shaft defect and eliminated manual work.

Treatment methods such as interim storage and immobilization are being considered to dispose of intermediate level waste (ILW), but some wastes that have been treated in the past may require repackaging. Re-packaging means to cover repackaging of waste that has already been packaged in a waste container and re-packaging is required for the following reasons: loss of shielding or containment, damage to external handling features, package out-of-specification, insufficient records and external policy. The re-packaging includes various methods such as non-intrusive treatment, overpacking of waste package, external treatment of waste container, repair waste container, injection of stabiliser, disassemble waste package, high temperature process, and dissolve waste package. The purpose of this paper is to evaluate the re-packaging possibility for various wastes by identifying the main repackaging methods among the above various re-packaging methods. 1) Disposal outside of the waste container is a viable technique for most packages, as coating with a portable spray gun for low dose rate packages or remotely using a robotic arm for high dose rate packages. 2) Waste container repair is divided into welding repair and patching of waste container according to the degree of damage. Weld repair and patching are important techniques that can be used to add additional shielding, repair damaged areas, and improve the integrity of lifting gears that may not be compliant. 3) In general, disassembly of waste packages has been applied to loose drummed waste. Packages and waste forms are physically disassembled, reduced in size, and placed in different new packages. For practical solution, grouted waste is repackaged by cutting using proprietary equipment such as diamond saws, wire saws, core drilling and rupture techniques. 4) High-temperature process involves cutting the waste package and placing the pieces in a hot bath of inorganic liquid or molten metal, and the process is applicable to all waste types. However, treatment of all gases produced, compliance with waste types and acceptance criteria. Finally, dissolving waste packages, which is generally considered impractical due to the variety of chemicals and radionuclides present in ILW, is a process that is easier to perform on raw ILW than conditioned waste. An example of waste being re-packaged is when old drummed waste is recovered from an old storage facility and the waste needs to be repackaged into a form that meets modern standards for interim storage and disposal.

The treatment of radioactive waste by melting has been mainly discussed with low-level waste (LLW). Considering that a large amount of waste in RV or RVI is intermediate-level waste (ILW), however, it is necessary to examine the possibility of treatment by melting of ILW. Different from LLW, melting of ILW with a high content of long-lived nuclides would lead to no free releasee, but has advantages in volume reduction, homogenization, and delay of release. In this paper, the possibility of melting as an alternative technology for the treatment of ILW in the future is reviewed by analyzing the benefits generated by melting ILW in the following aspects: 1) Similar to melting techniques of LLW, them of ILW are mostly based on well-known techniques, but it is necessary to review the feasibility of performing operations such as removal of solidified melt using remote equipment in abnormal situations such as loss of electricity. 2) It is necessary to specify radiation limits for the melting operation unless the ILW melting operation technique can guarantee that the risk of abnormal occurrence is very low. The main quantified radiation parameter is the ingot dose rate, which of 10 mSv/h is considered more reasonable. 3) Although the treatment of ILW by melting leads to a reduction in volume, the main characteristics of the waste still remain, and no waste can be disposed of for free release. Thus, the main potential benefits are improved long-term safety and reduced waste volume. 4) Reducing the surface-to-volume ratio of the molten material could reduce the amount of corrosive material per unit time and, consequently, increase long-term safety. Its effect on long-term safety is difficult to quantify precisely as it depends on several factors, such as the geometry of the original component or whether radionuclides were distributed on the surface of the original component or the induced radioactivity. 5) The volume reduction of ILW is estimated to be reduced by about 1/4 compared to the generated amount when assuming a disposal volume reduction factor of 3 and considering the dose reduction due to radioactive decay after long-term storage, however, due to the lack of knowledge about non-hazardous facility alternatives, it is difficult to evaluate cost-benefit. This is heavily influenced by both the final volume reduction and the potential to reduce the complexity of the repository’s technical barriers.

Korea Radioactive Waste Agency (KORAD), regulatory body and civic groups are calling for an infrastructure system that can more systematically and safely manage data on the results of radioactive waste sampling and nuclide analysis in accordance with radioactive waste disposal standards. To solve this problem, a study has been conducted on the analysis of the nuclide pattern of radioactive waste on the nuclide data contained in low-and intermediate-level radioactive waste. This paper will explain the optimal repackaged algorithm for reducing radioactive waste based on previous research results. The optimal repackaged algorithm for radioactive waste reduction is comprised based on nuclide pattern association indicators, classification by nuclide level of small-packaged waste, and nuclide concentration. Optimization simulation is carried out in the order of deriving nuclide concentration by small-packaged, normalizing drum minimization as a function of purpose, normalizing constraints, and optimization. Two scenarios were applied to the simulation. In Scenario 1 (generating facilities and repackaged by medium classification without optimization), it was assumed that there are 886 low-level drums and 52 very low-level drums. In Scenario 2 (generating facilities and repackaged by medium classification with optimization), 708 and 230 drums were assigned to the low-level and very low-level drums, respectively. As a result of the simulation, when repackaged in consideration of the nuclide concentration and constraints according to the generating facility cluster & middle classification by small package (Scenario 2) the low-level drum had the effect of reducing 178 drums from the baseline value of 886 drums to 708 drums. It was found that the reduced packages were moved to the very low-level drum. The system that manages the full life-cycle of radioactive waste can be operated effectively only when the function of predicting or tracking the occurrence of radioactive waste drums from the source of radioactive waste to the disposal site is secured. If the main factors affecting the concentration and pattern of nuclides are systematically managed through these systems, the system will be used as a useful tool for policy decisions that can prevent human error and drastically reduce the generation of disposable drums.

Trades are classified as a direct trade, in which an exporter and an importer directly conclude a contract to execute a transaction, and an indirect trade, in which a transaction is conducted through a third party. A license issued by NSSC is required for the indirect trade of Trigger List Items although the items do not cross the Korea’s borders. This study would summarize characteristics for each type of the indirect trade, and suggest things considered from the perspective of regulations. Sensitive items such as weapons of mass destruction could be illegally transferred through black trade, and the role of brokers is important. Therefore, Korea controls indirect trade of Trigger List Items in a more conservative approach, although NSG guidelines does not include it as a control scope. Advanced countries in export control field such as the US, UK, and Canada also controls on indirect trade. One of the indirect trade types is an intermediate trade in which the goods are imported from a foreign country and exported to another foreign country without being brought into the trader’s country. Another type is a merchandising trade which is a form of brokerage between exporter from a foreign country and importer from another foreign country. Both types have one thing in common that the goods are not crossed the trader’s country. The difference is that an intermediate trader directly participates in the contract and earn a difference between import and export amount, whereas a merchandising trader just arranges the transaction and earns a brokerage fee. The profits from the intermediate trade are considered as export records, while merchandising trade profits are not considered export records. In other words, only the intermediary trade is considered as an export. Also, the license types are different for each of them. An export license should be issued for the intermediate trade of Trigger List Items, whereas a brokerage license should be issued for the merchandising trade of Trigger List Items. The definition of export in the Foreign Trade Act includes intermediate trade for only goods, but technology is missing, although the sub-regulation specifies the intermediate trade including both. The technology need to be added as it can be the subject of intermediate trade in spite of intangible characteristics. Also, outreach activities are more needed as nuclear industry awareness on export control for both trades is low.

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate- level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

이 연구는 해양경찰조직의 발전을 위해 해양경찰공무원이 인식하는 권한 부여 리더십이 조직시민행동과의 관계를 검증 목적 으로 연구를 진행하였다. 리더십에는 다양한 유형이 있지만 해양경찰공무원은 직무 특성상 부여된 권한이 실무에서 더욱 효과적인 문제 해결에 도움이 될 것이며, 비록 비공식적 행동이지만 결과적으로 조직의 발전으로 연결되는 조직시민행동은 최근 조직발전과 관련된 연 구로 많이 진행된다는 점에서 접근하였다. 이 연구에서는 해양경찰공무원 168명을 대상으로 구조화된 설문지를 통해 연구가 진행되었으 며 연구 결과 해양경찰공무원의 직무는 위험성, 돌발성 등으로 인해 해양경찰공무원은 상호 의존적 직무 성격이 중요하므로 중간 관리자 의 목표에 대한 설명과 구성원에 대한 다양한 행동은 해양경찰 조직성과에 긍정적 결과로 연결된다는 점에서 정책적 제언을 하였다.

경주 방폐물 처분시설의 1단계 시설로 건설된 지하 사일로 구조는 2014년에 10만 드럼 규모로 완공되어 현재 운영중에 있다. 지하 사일로 구조는 지름 25m, 높이 50m로써 방폐물을 저장하는 실린더부분과 돔 부분으로 구성되어 있으며, 돔부분은 운영터널과 연결 되는 하부 돔 부분과 상부 돔 부분으로 구분할 수 있다. 지하 사일로 구조의 벽체는 철근콘크리트 라이너이고, 두께는 약 1m이다. 본 논문에서는 지하 사일로 구조의 건설과정 및 운영과정의 단계별 유한요소해석을 수행하였다. SMAP-3D 프로그램을 사용하여 2차원 축대칭 유한요소해석을 수행하였다. 2차원 축대칭 유한요소모델의 신뢰성을 검토하고자 3차원 유한요소해석도 수행하였다. 본 논문 에서는 지하 사일로 구조의 구조거동을 분석하고 구조적 안전성을 검토결과를 제시하였다.