학술논문
Design and Structural Safety Evaluation of Transfer Cask for Dry Storage System of PWR Spent Nuclear Fuel
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- 영문명
- 발행기관
- 한국방사성폐기물학회
- 저자명
- Taehyung Na Youngoh Lee Taehyeon Kim Yongdeog Kim
- 간행물 정보
- 『Journal of Nuclear Fuel Cycle and Waste Technology (JNFCWT)』Vol.21 No.4, 503~516쪽, 전체 14쪽
- 주제분류
- 공학 > 공학일반
- 파일형태
- 발행일자
- 2023.12.30
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국문 초록
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.
영문 초록
목차
해당간행물 수록 논문
- Design and Structural Safety Evaluation of Transfer Cask for Dry Storage System of PWR Spent Nuclear Fuel
- Design and Structural Safety Evaluation of Canister for Dry Storage System of PWR Spent Nuclear Fuels
- Assessing the Potential of Small Modular Reactors (SMRs) in Spent Nuclear Fuel Management: A Review of the Generation IV Reactor Progress
- Numerical Modelling of Radionuclide Migration for the Underground Silo at Near-Field
- Thermal Analyses of Deep Geological Disposal Cell With Heterogeneous Modeling of PLUS7 Spent Nuclear Fuel
- Review of Aging Management for Concrete Silo Dry Storage Systems
- Preliminary Selection of Safety-Relevant Radionuclides for Long-Term Safety Assessment of Deep Geological Disposal of Spent Nuclear Fuel in South Korea
- Development of a Methodology for Estimating Radioactivity Concentration of NORM Scale in Scrap Pipes Based on MCNP Simulation
- Comparison of Environmental Radiation Survey Analysis Results in a High Dose Rate Environment Using CZT, NaI(Tl), and LaBr3(Ce) Detectors
- Characterization of the Purified Ca-type Bentonil-WRK Montmorillonite and Its Sorption Thermodynamics With Cs(I) and Sr(II)
- Rolling Test Simulation of Sea Transport of Spent Nuclear Fuel Under Normal Transport Conditions
- The French Underground Research Laboratory in Bure: An Essential Tool for the Development and Preparation of the French Deep Geological Disposal Facility Cigéo
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