A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

The treatment of waste generated during operation as a part of preparation for decommissioning is coming to the fore as a pending issue. Non-fuel waste stored in the spent fuel pool (SFP) of PWRs in Korea includes Dummy fuel, damaged fuel rod storage container, reactor vessel specimen cask, spent in-core instrumentation, spent control element assemblies, spent neutron source assemblies, burnable poison rods, etc. In order to treat such waste, it is necessary to classify radioactive waste level and analyze kinds of nuclide in accordance with legal requirements. In order to solve the problem, the items that KHNP-CRI is trying to conduct like followings. First, KHNP-CRI will identify the current status of non-fuel waste stored in the SFP of all domestic nuclear power plants. In order to consider the treatment of non-fuel waste, it is essential to know what kind of items and how many items are stored in the SFP. Second, to identify the dimension and characteristics of non-fuel waste stored in the SFP would be conducted. The configuration of non-fuel waste is important information to handle them. Third, the way to handle non-fuel waste would be deduced including analysis of their dimension, whether the equipment should be developed to handle each kind of non-fuel waste or not, how to transport them. In order to classify radioactive waste level and analyze the nuclide for the non-fuel waste, handling tools and the cask to transport them into the facility which nuclide analysis is able to be performed would be required. Fourth, the nuclide analysis technology would be identified. Also, domestic holding technology would be identified and which technology should be developed to classify the radioactive waste level for the non-fuel waste would be deduced. This preliminary study will provide KHNP-CRI with the insight for the nuclide analysis technology and future work which is following action for the non-fuel waste. Based on the result of above preliminary study, the feasibility of the research for the treatment of non-fuel waste would be evaluated and research plan would be established. In conclusion, the treatment of non-fuel waste stored in the spent fuel pool of domestic PWR should be considered to prepare the decommissioning. KHNP-CRI will identify the quantity, the dimension and kinds of non-fuel waste in the SFP of domestic PWR. Also, the various nuclide analysis technology would be identified and the technology which should be developed would be defined through this preliminary study.

Spent nuclear fuel (SNF) characterization is important in terms of nuclear safety and safeguards. Regardless of whether SNF is waste or energy resource, the International Atomic Energy Agency (IAEA) Specific Safety Guide-15 states that the storage requirements of SNF comply with IAEA General Safety Requirement Part 5 (GSR Part 5) for predisposal management of radioactive waste. GSR Part 5 requires a classifying and characterizing of radioactive waste at various steps of predisposal management. Accordingly, SNF fuel should be stored/handled as accurately characterized in the storage stage before permanent disposal. Appropriate characterization methods must exist to meet the above requirements. The characterization of SNF is basically performed through destructive analysis/non-destructive analysis in addition to the calculation based on the reactor operation history. Burnup, Initial enrichment, and Cooling time (BIC) are the primary identification targets for SNF fuel characterization, and the analysis mainly uses the correlation identified between the BIC set and the other SNF characteristics (e.g., Burnup - neutron emission rate) for characterizing. So further identification of the correlation among SNF characteristics will be the basis for proposing a new analysis method. Therefore, we aimed to simulate a SNF assembly with varying burnup, initial enrichment, and cooling time, then correlate other SNF properties with BIC sets, and identify correlations available for SNF characterization. In this study, the ‘CE 16×16’ type assembly was simulated using the SCALEORIGAMI code by changing the BIC set, and decay heat, radiation emission characteristics, and nuclide inventory of the assembly were calculated. After that, it was analyzed how these characteristics change according to the change in the BIC set. This study is expected to be the basic data for proposing new method for characterizing the SNF assembly of PWR.

Recently, the deep geological disposal system isolating a spent nuclear fuel (SNF) is considered a disposal method of high-level radioactive waste for the safety of humans or the natural environment. The one of important requirements for maintaining the thermal stability of these systems is that the temperature of the buffer does not exceed 100°C even though the decay heat is emitted from highlevel radioactive wastes loaded in the disposal container. In 2007, a deep geological disposal system based on the Swedish disposal concept was developed for the SNF in Korea. To respond to the development process, the thermal stability of the deep geological disposal system developed for the disposal of domestic pressurized light water reactor (PWR) SNFs with discharged burn-up of 55 GWD/MTU was evaluated in 2019. The thing is that the recent fuel activity is pursuing to operate further high burn-up fuel conditions, and it leads to emergency core cooling system (ECCS) revision for extending the license for up to 60 or more than 60 GWD/MTU in the world. In this regard, this study evaluates numerically the thermal stability of the deep geological disposal system for the high burn-up PWR SNF having large decay heat compared to previous conditions for two different length disposal containers classified according to the length of PWR SNFs discharged from domestic nuclear power plants. A finite element analysis using a computational program was used to evaluate the thermal design requirements. Results show that both types of disposal containers would increase the temperature which reduces or fails to meet the safety margin of the disposal system. This study suggests that the design of the previous disposal system is needed to be further developed for the high burn-up PWR SNF which would be used in future nuclear power plant systems.

The manufactured nuclear fuel assembly is loaded into the nuclear reactor after the core design, and is finally discharged to the wet storage pool after depletion for 3 cycles. The discharged spent nuclear fuel is transported and stored in a dry storage system at the on-site of the nuclear power plant, which is cooled by natural convection, and undergoes final disposal or reprocessing through an intermediate dry storage facility. In this series of processes, the characteristics of the final product, the spent fuel, vary depending on the environmental conditions, so it is essential to manage each history data to verify the long-term integrity of the spent nuclear fuel. In this paper, safety information on spent nuclear fuel is described in order to establish technical requirements that should be considered in each stage of storage, transport, reprocessing, and disposal of spent nuclear fuel. Comprehensive safety information on spent nuclear fuel is basically calculated from basic information that considers characteristic information that can be obtained through the manufacture and design of nuclear fuel assemblies, operation history in a nuclear reactor, and location history in a wet storage pool. It can be divided into secondary production information (SF Burnup, Nuclide Inventory, etc.) and tertiary integrity-related information obtained through cladding inspection during spent fuel storage. KHNP produces this multi-layered information according to the production stage and manages it through the comprehensive management system of the spent nuclear fuel, and safety information with some errors is not only improved through re-verification but also continuously updated. In this paper, the spent nuclear fuel safety information was derived based on various information calculated in the entire process of being discharged and managed in a wet storage pool, including new fuel manufacturing information and depletion history. Such safety information will be used as basic data for long-term safe management of spent nuclear fuel, and will be continuously produced and managed. In the future, additional discussions will be held on the safety information of the spent nuclear fuel through consultation with KORAD and regulatory agencies.

사용후핵연료의 효율적 관리를 위하여 한국원자력연구원에서 수행 중인 파이로 공정으로부터 발생되는 폐기물 처리기술에 대한 최근 연구동향을 종합적으로 고찰하였다. 파이로 폐기물 처리기술은 처분 대상 폐기물의 감용 및 포장, 저장과 최종 처분에 적합한 고화체 제조를 목표로 하고 있다. 한국원자력연구원에서 수행 중인 파이로 폐기물 처리 기술개발 접근 방향은 공정 흐름으로부터 발생한 폐기물내 주요 핵종들을 분리하고 회수한 물질 등을 재사용함으로서 폐기물 발생량을 최소로 하며 동시에 분리한 핵종을 별도로 고화처리하는 것이다. 폐기물 처리 주요 기술 특성은 먼저 전해환원용 원료물질 제조를 위하여 전처리 고온 열처리 공정을 사용하며, LiCl 과 LiCl-KCl 염으로부터 핵종을 분리하고 회수염의 재사용 및 핵종 함유량을 증대시킨 최종 고화체 제조 기술을 개발하는 것이다. 따라서 실험실 규모 실험 결과를 토대로 최근에는 공정 용량 증대를 위한 자료 확보를 목적으로 공학규모 시험을 수행 중에 있다.

본 논문은 국내 원전의 습식저장조에 저장 중인 경수로형 사용후핵연료를 금속겸용용기를 이용해 건식으로 운영하기 위한 운영공정을 개발하는 것이다. 국내 경수로형 원전의 사용후핵연료는 1990년대 초부터 습식으로 소내에서 운반을 한 경험은 많으나 건식으로 운전한 경험은 전혀 없는 실정이다. 이에 따라 금속겸용용기를 운영할 수 있는 세부 운영공정을 개발하 였으며 주요 운영공정에서 금속겸용용기의 주요 구성품 및 사용후핵연료의 안전성이 유지됨을 확인하였다. 단기운영공정은 총 21시간 내에 이루어지도록 절차를 수립하였고 단계별로 허용운전 시간(15시간 습식공정, 3시간 배수공정, 그리고 3시간 진공공정)도 제시하였다.

최근 국내 원전의 경수로 사용후핵연료 습식 저장시설의 포화시점이 다가옴에 따라 운반 및 저장용기를 이용한 건식저장시스템 개발이 활발하게 수행되고 있다. 일반적으로 사용후핵연료 운반 및 저장용기 설계를 위한 차폐해석 시 장전 가능 연료 중 가장 보수적인 연료를 설계기준연료로 선정하여 해석을 수행한다. 그러나 실제 금속 운반용기에 장전되는 사용후핵연료 는 해석평가에 적용된 설계기준연료에 한정되지 않고 다양하기 때문에 초기농축도, 연소도, 최소냉각기간의 특성을 고려한 차폐평가를 통하여 장전가능 여부가 결정된다. 이에 본 연구에서는 금속 겸용용기에 장전 가능한 연료를 대상으로 국내 운반기준을 만족하는 최소냉각기간의 결정을 위한 차폐해석 방법을 기술하였다. 특히 발생량이 많은 초기농축도 3.0~4.5wt% 의 사용후핵연료는 차폐해석 구간을 세분화하여 평가하여 연구결과의 활용에 효율성을 높이고자 하였다. 차폐평가를 통해 2008년까지 국내 원전에서 발생한 장전대상연료 중 약 81%의 사용후 핵연료를 금속겸용용기로 운반할 수 있는것으로 평가 되었다. 본 연구결과를 통해 금속 겸용용기의 운반조건에 장전 가능한 연료의 특성을 제시함으로써 운반 시 운영절차의 개 발을 위한 기술적 근거 수립에 도움이 되고자 한다.