• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.559-570
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• 2023

The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4597-4606
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• 2023

Thermal analysis of the CANDU spent fuel dry storage canister is very important to ensure the integrity of the spent fuel. The analyses have been conducted using a conservative approach, with a particular focus on the peak cladding temperature (PCT) of the fuel rods in the canister. In this study, we have performed a realistic thermal analysis using a computational fluid dynamics (CFD) code. The canister contains 9 fuel bundle baskets. A detailed analysis of even a single basket requires significant computational resources. To overcome this challenge, we replaced each basket with an equivalent heat conductor (EHC), of which effective thermal conductivity (ETC) is developed from the results of detailed CFD calculations of a fuel bundle basket. Then, we investigated the effects of some conservative models, ultimately aiming at a realistic analysis. The results revealed: (i) The influence of convective heat transfer in the basket cannot be ignored, but it's less significant than expected. (ii) Modeling of the lifting rod is crucial, as it plays a decisive role in axial heat transfer at the center of the canister and significantly reduces the PCT. (iii) Convection within the canister is very important, as it not only reduces the PCT but also shifts its location upwards.

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

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.166-177
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• 1993

A simplified thermal analysis method to evaluate the maximum temperature of the CANDU 37-element fuel bundle within a fuel basket in a given spent fuel dry storage canister has been presented along with the results of sample analyses performed to examine the parametric effects of the ambient conditions on the maximum fuel temperature within a canister. To solve the multi-dimensional heat transfer problem of the complex geometry of rod bundles within a canister where three modes of heat transfer are superimposed, the CANDU spent fuel bundles stored in the dry storage canister are first replaced by equivalent concentric fuel cylinders. The simplified axi-symmetric two-dimensional multi-mode heat transfer problem of the equivalent fuel cylinders is then analyzed with an existing computer code, HEATING5, using additional input data and heat transfer correlations. A comparison between the predicted temperature profile and the mock-up test results shows that the agreement is quite satisfactory.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.90-91
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• 2019

• The KSFM Journal of Fluid Machinery
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• v.19 no.4
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• pp.19-28
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• 2016

Objective of this study is to compare the inherent characteristics of natural convection flow inside the canister of spent fuel dry storage system with different backfill gases by utilizing computational fluid dynamics (CFD) code. Four working fluids were selected for comparison study. Helium currently used backfill gas for canister, air, nitrogen, and argon are frequently used as coolant in many heat transfer applications. The results indicate that helium has very distinct conductive behavior and show very weak natural convective flow compared to the others. Argon showed the strongest natural convective flow but also the worst coolability. Air and nitrogen showed similar characteristics to each other. However, due to difference in Prandtl number, nitrogen showed more effective natural convective flow. These results suggest that experimental validation for the nitrogen is needed to investigate the potential coolability other than currently commercially used helium.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.359-370
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• 2023

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.317-327
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• 2024

This study investigated the chloride-induced stress corrosion cracking (CISCC) behavior expected to occur in welds of austenitic stainless steel, which are considered candidate materials for dry storage containers for spent nuclear fuel. The behavior was studied by varying temperature, relative humidity (RH), and chloride concentration. 304L-ER308L welded plates were processed into U-bend specimens and exposed to a cyclic corrosion chamber for 12 weeks. The CISCC behavior was then analyzed using electron microscopy. A previous study by the authors confirmed that CISCC occurred in ER308L at 60 ℃, 30% RH, and 0.6 M NaCl via selective corrosion of δ-ferrite. When the temperature was lowered from 60 ℃ to 50 ℃, CISCC still occurred. However, when the humidity was reduced to 20% RH, CISCC did not happen. This can be attributed to the retardation of the deliquescence of NaCl at lower humidity, which was insufficient to promote CISCC. Furthermore, increased chloride concentration to 1.0 M resulted in the absence of CISCC and widespread surface corrosion with severe pitting corrosion because of the increase in thin film thickness.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.403-409
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• 2017

For dry storage of the spent nuclear fuel (SNF) stored in the storage pool of a nuclear power plant, essentially all moisture must be removed to prevent corrosion of the assembly and canister internals and/or degradation of fuel cladding integrity after SNF canister loading operation. R&D work is now in progress on a forced gas drying system that can be used to remove residual water in canisters. In this work, preliminary design is performed to manufacture the forced gas drying system. This process includes a case study of dry methods for canister moisture removal, relative codes and standards, confirmation of adequate dryness, needs analysis at plant sites, and characteristics of SNF stored in pools. Through this preliminary design work, we obtained a conceptual flow diagram and preliminary P&ID of the forced gas drying system. The results of this study can be used to determine details of the design to manufacture the forced gas drying system.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.75-76
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• 2018