• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.400-401
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• 2005

• 한국세라믹학회지
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• 제51권5호
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.423-430
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• 2014

An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.313-317
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• 2018

Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2887-2900
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• 2020

Cladding failure in a fuel rod during operation in a BWR is analyzed using a FALCON code-based model. Comparative calculation with a neighbouring, intact rod is presented, as well. A considerable 'hot spot' effect in cladding temperature is predicted with the SLICE-DO model due to a thermal barrier caused by the localized crud deposition. Particularly significant overheating is expected to occur on the affected side of the cladding of the failed rod, in agreement with signs of significant localized crud deposition as revealed by Post Irradiation Examination. Different possibilities (criteria) are checked, and Pellet-Cladding Mechanical Interaction (PCMI) is shown to be one of the plausible potential threats. It is shown that PCMI could lead to discernible concentrated inelastic deformation in the overheated part of the cladding. None of the specific mechanisms considered can be experimentally or analytically identified as an only cause of the rod failure. However, according to the current calculation, a possibility of cladding failure by PCMI cannot be excluded if the crud thickness exceeded 300 ㎛.

• 시스템엔지니어링학술지
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• 제13권1호
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• pp.33-39
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• 2017

This paper describes the system engineering approach for the heat transfer analysis of plus7 fuel rod for APR1400 using, a commercial software, ANSYS. The fuel rod is composed of fuel pellets, fill gas, end caps, plenum spring and cladding. The heat is transferred from the pellet outward by conduction through the pellet, fill gas and cladding and further by convection from the cladding surface to the coolant in the flow channel. The goal of this paper is to demonstrate the temperature and heat flux change from the fuel centerline to the cladding surface when having maximum fuel centerline temperature at 100% power. This phenomenon is modelled using the ANSYS FEM code and analyzed for steady state temperature distribution across the fuel pellet and clad and the results were compared to the standard values given in APR1400 SSAR. Specifically the applicability of commercial software in the evaluation of nuclear fuel temperature distribution has been accounted. It is note that special codes have been used for fuel rod mechanical analysis which calculates interrelated effects of temperature, pressure, cladding elastic and plastic behavior, fission gas release, and fuel densification and swelling under the time-varying irradiation conditions. To satisfactorily meet this objective we apply system engineering methodologies to formulate the process and allow for verification and validation of the results acquired. The close proximity of the results obtained validated the accuracy of the FEM analysis of the 2D axisymmetric model and 3D model. This result demonstrated the validity of commercial software instead of proprietary in-house code that is more costly to develop and maintain.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.1-15
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• 2016

For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell $UO_2$ and high-density composite pellet concepts are being developed as ATF pellets. A microcell $UO_2$ pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts-surface-modified Zr-based alloy and SiC composite material-are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.

• 방사성폐기물학회지
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• 제21권3호
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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.

• 한국재료학회지
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• 제14권7호
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• pp.462-469
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• 2004

The effect of Cu addition on the corrosion characteristics of Zr alloys that developed for nuclear fuel cladding in KAERI (Korea Atomic Energy Research Institute) was evaluated. The alloys having different element of Nb, Sn, Fe, Cr and Cu were manufactured and the corrosion tests of the alloys were performed in static autoclave at $360^{\circ}C$, distilled water condition. The alloys were also examined for their microstructures using the optical microscope and the TEM equipped with EDS and the oxide property was characterized by using X-ray diffraction. From the result of corrosion test more than 450 days, the corrosion rate of the Zr-based alloys was changed with alloying element such as Nb, Sn, Fe, Cr and especially affected by Cu addition. The corrosion resistance was increased with increasing the Cu content and the tetragonal $ZrO_2$ layer was more stabilized on the Cu-containing alloys.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2475-2490
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• 2022

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO₂-Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.