• 방사성폐기물학회지
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• 제19권2호
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• pp.205-214
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• 2021

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2901-2909
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• 2020

Thermal creep is a key property of zircaloy cladding. CZ developed by CGN is a new zircaloy used as PWR fuel cladding. This research is devoted to investigating the thermal creep behavior of CZ and build the thermal creep model of CZ. Twenty internal pressure creep tests were conducted, and the ranges of temperature and Tresca stress were 320-430 ℃ and 70-300 MPa, respectively. Real-time creep data were analyzed by separating primary creep and steady-state creep. Based on Soderberg model and creep test data, CZ thermal creep model is derived. As a whole, the mean value and the standard deviation of P/M of CZ saturated primary creep strain are very close to these from steady-state creep rate, however, the predictive effect of primary creep is less satisfactory. Four conditions, where there exists large deviation between predicted values and test data, are 320 ℃ and 300 MPa, 350 ℃ and 190 MPa, 380 ℃ and 160 MPa, 380 ℃ and 190 MPa, respectively. As primary creep was much smaller than steady-state creep in long-time operation, the thermal creep model built can be applied to predict the thermal creep behavior of CZ cladding.

• 한국공간구조학회논문집
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• 제20권1호
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• pp.49-59
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• 2020

This study investigates the wind pressure characteristics of elliptical plan retractable dome roof. Wind tunnel experiments were performed on spherical dome roofs with varying wall height-span ratios (0.1~0.5) and opening ratios (0%, 10%, 30% and 50%), similar to previous studies of cirular dome roofs. In previous study, wind pressure coefficients for open dome roofs have been proposed since there are no wind load criteria for open roofs. However, in the case of Eeliptical plan retractable dome roof, the wind pressure coefficient may be largely different due to the presence of the longitudinal direction and transverse direction. The analysis results leads to the exceeding of maximum and minimum wind pressure coefficients KBC2016 code.

• 대한기계학회논문집A
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• 제25권2호
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• pp.275-282
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• 2001

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement. Number of subclad cracks may be found during an in-service-inspection due to the presence of cladding. It is specified, in ASME Sec. XI, that a subclad crack is characterized as a surface crack when the thickness of the clad is less than 40% of the crack depth. This condition is provided to keep the crack integrity evaluation conservative. In order to refine the fracture assessment procedures for such subclad cracks under a pressurized thermal shock condition, three dimensional finite element analyses are applied for various subclad cracks existing under cladding. A total of 36 crack geometries are analyzed, and the results are compared with those for surface cracks. The resulting stress intensity factors for subclad cracks are 6 to 44% less than those for surface cracks. It is proven that the flaw characterization condition as specified in ASME Sec. XI can be overly conservative for some subclad cracks.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4499-4513
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• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4066-4076
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• 2023

Pressure resistance welding is usually used to seal the connection between the cladding tube and the end plug made of zirconium alloy. The seal welded joint has a direct effect on the service performance of the fuel rod cladding structure. In this paper, the pressure resistance welded joints of zirconium alloy tube-plug structure were obtained by thermal-mechanical simulation experiments. The microstructure and microhardness of the joints were both analyzed. The effect of processing parameters on the microstructure was studied in detail. The results showed that there was no β-Zr phase observed in the joint, and no obvious element segregation. There were different types of Widmanstätten structure in the thermo-mechanically affected zone (TMAZ) and heat affected zone (HAZ) of the cladding tube and the end plug joint because of the low cooling rate. Some part of the grains in the joint grew up due to overheating. Its size was about 2.8 times that of the base metal grains. Due to the high dislocation density and texture evolution, the microhardnesses of TMAZ and HAZ were both significantly higher than that of the base metal, and the microhardness of the TMAZ was the highest. With the increasing of welding temperature, the proportion of recrystallization in TMAZ decreased, which was caused by the increasing of strain rate and dislocation annihilation.

• Steel and Composite Structures
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• 제23권1호
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• pp.107-114
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• 2017

Design of stepwise foam claddings subjected to air-blast is performed based on random Voronoialgorithm. FE models are constructed using the random Voronoialgorithm, and numerical analysis is carried out to simulate deformation mode and energy absorption of the cladding by the ABAQUS/Explicit software. The FE model is validated by test result, and good agreement is achieved. The deformation patterns are presented to give an insight into the influences of distribution on deformation mechanisms. The energy absorbed by the stepwise foam cladding is examined, and the parameter effects, including layer number, gradient, and blast loading, are discussed. Results indicate that the energy absorption capacity increases with the number of layer, gradient degree, and blast pressure increasing.

• Corrosion Science and Technology
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• 제18권5호
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• pp.190-195
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• 2019

SA508 low-alloy steel for a reactor vessel was exposed to primary water in a pressurized water reactor (PWR) plant because the cladding layer of type 309 stainless steel for the RPV was removed, due to an accident in which the detachment of the thermal sleeve occurred. The major advantage of the electrochemical deposition (ECD) Ni plating technique is that the reactor pressure vessel can be repaired without significant thermal effects, and Ni has solid corrosion resistance that can withstand boric acid. The corrosion rate assessment of the damaged part was performed, and its trend was analyzed. Essential variables of the Ni plating for repair of the damaged part were derived. These conditions are applicable variables for the repair plating device, and have been carefully adjusted using the repair plating device. The process for establishing ASME technical standards called Code Case N-840 is described. The process of developing Ni-plating devices, and the electroplating procedure specification (EPS) are described.

• Nuclear Engineering and Technology
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• 제40권1호
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• pp.1-8
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• 2008

One of the major current challenges to nuclear energy lies in its competitiveness. To stay competitive the industry needs to reduce maintenance and fuel cycle costs, while enhancing safety features. Extended burnup is one of the methods applied to meet these objectives However, there are a number of potential fuel failure causes related to increased burnup, as follows: l) Corrosion of zirconium alloy cladding and the water chemistry parameters that enhance corrosion; 2) Dimensional changes of zirconium alloy components, 3) Stresses that challenge zirconium alloy ductility and the effect of hydrogen (H) pickup and redistribution as it affects ductility, 4) Fuel rod internal pressure, 5) Pellet-cladding interactions (PCI) and 6) pellet-cladding mechanical interactions (PCMI). This paper discusses current and potential failure mechanisms of these failure mechanisms.

• Journal of Welding and Joining
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• 제13권2호
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• pp.139-149
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• 1995

Many pressure vessels for the power plants are fabricated from low alloy ferritic steels. The inner sides of the pressure vessels are commonly weld_cladded with austenitic stainless steels to minimize problems of corrosive attack. The submerged-arc welding(SAW) process is now used in preference to other processes because of the possibilities open to automation to reduce the overaII welding times. The most reliable way to avoid underclad cracks(UCC) which are often detected at the overlap of the clad beads is to use nonsusceptible steels such as SA508 class 3. At present domestically developed forging steel of SA508 cl.S is now being cladded with single layer by using 90mm wide strip, which transfers higher heat input into the base metal compared to the conventional two layers strip cladding which has been in wide use with 30-60 mm wide strip. But the current indices for the influence of heat input on crack susceptibility are not accurate enough to express the subtle difference in crack susceptibility of the steel. Therefore, the purpose of this present study is: l) To determine UCC susceptibility on domestic forging steel, SA508 cl.S cladded with single layer by using submerged arc 90mm strip and, 2) To optimize heat input range by which the crack susceptibility could be eliminated.