• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.390-391
• /
• 2005

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.161-162
• /
• 2005

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2019.05a
• /
• pp.55-56
• /
• 2019

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.11a
• /
• pp.79-80
• /
• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.05a
• /
• pp.397-398
• /
• 2018

• Nuclear Engineering and Technology
• /
• v.50 no.6
• /
• pp.890-898
• /
• 2018

Nondestructive inspection (NDI) is an integral part of structural integrity analyses of dry storage casks that house spent nuclear fuel. One significant concern for the structural integrity is stress corrosion cracking in the heat-affected zone of welds in the stainless steel canister that confines the spent fuel. In situ NDI methodology for detection of stress corrosion cracking is investigated, where the inspection uses a delivery robot because of the presence of the harsh environment and geometric constrains inside the cask protecting the canister. Shear horizontal (SH) guided waves that are sensitive to cracks oriented either perpendicular or parallel to the wave vector are used to locate welds and to detect cracks. SH waves are excited and received by electromagnetic acoustic transducers (EMATs) using noncontact ultrasonic transduction and pulse-echo mode. A laboratory-scale canister mock-up is fabricated and inspected using the proposed methodology to evaluate the ability of EMATs to excite and receive SH waves and to locate welds. The EMAT's capability to detect notches from various distances is evaluated on a plate containing 25%-through-thickness surface-breaking notches. Based on the results of the distances at which notch reflections are detectable, NDI coverage for spent nuclear fuel storage canisters is determined.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.243-244
• /
• 2005

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.1
• /
• pp.110-116
• /
• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.9 no.3
• /
• pp.191-198
• /
• 2011

In general, conventional criticality analyses for spent fuel transport/dry storage systems have been performed based on assumption of fresh fuel concerning the potential uncertainties from number density calculation of Transuranic and Fission Products in spent fuel. However, because of economic loss due to the excessive criticality margin, recently the design of transport/dry storage systems with Burnup Credit(BUC) application has been actively developed. The uncertainties in criticality analyses on transport/storage systems with BUC technique show strong dependance upon initial enrichment and burnup rate, whereas those in the conventional criticality evaluation based on fresh fuel assumption do not show such a dependance. In this study, regulatory-required uncertainties of the criticality analyses for BK 26 Cask, which is conceptually designed spent fuel transport cask with BUC corresponding to the limiting circumstances on nuclear power plants in Korea, are evaluated as a function of initial enrichment and burnup rate. Results of this study will be used as basic data for spent fuel loading curve of BK 26 Cask.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.4
• /
• pp.411-422
• /
• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.