• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.248-254
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• 1993

The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38$^{\circ}C$ for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277$^{\circ}C$ and 226$^{\circ}C$, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.

• Corrosion Science and Technology
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• v.6 no.4
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• pp.170-176
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• 2007

Iodine-induced stress corrosion cracking (ISCC) properties and the associated ISCC process of Zircaloy-4 and an Nb-containing advanced nuclear fuel cladding were evaluated. An internal pressurization test with a pre-cracked specimen was performed with a stress-relieved (SR) or recrystallized (RX) microstructure at $350^{\circ}C$, in an iodine environment. The results showed that the $K_{ISCC}$ of the SR and RX Zircaloy-4 claddings were 3.3 and 4.8MPa\;m^{0.5}, respectively. And the crack propagation rate of the RX Zircaloy-4 was 10 times lower than that of the SR one. The chemical effect of iodine on the crack propagation rate was very high, which was increased $10^4$ times by iodine addition. Main factor affecting on the micro-crack nucleation was a pitting formation and its agglomeration along the grain boundary. However, this pitting formation on the grain-boundary was suppressed in the case of an Nb addition, which resulted in an increase of the ISCC resistance when compared to Zircaloy-4. Crack initiation and propagation mechanisms of fuel claddings were proposed by a grain boundary pitting model and a pitting assisted slip cleavage model and they showed reasonable results.

• Corrosion Science and Technology
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• v.13 no.2
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• pp.56-61
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• 2014

The purpose of this work is to investigate the effect of dissolved hydrogen concentration on crud deposition onto the fuel cladding surface in the simulated primary environments of a pressurized water reactor. Crud deposition tests were conducted in the dissolved hydrogen concentration range of 5~70 cc/kg at $325^{\circ}C$ for 14 days. Needle-shaped NiO deposits were formed in the hydrogen range of 5~25 cc/kg, while polygonal nickel ferrite deposits were observed at a hydrogen concentration above 35 cc/kg. However, the dissolved hydrogen content seems to have little effect on the amount of crud deposits.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1110-1119
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• 2020

The Floating Absorber for Safety at Transient (FAST) is a safety device used in the innovative Sodium-cooled Fast Reactor (iSFR). The FAST insert negative reactivity under transient or accident conditions. However, behavior of the FAST is still unclear under transient conditions. Therefore, the existing Floating Absorber for Safety at Transient Analysis Code (FASTAC) is improved to analyze the FAST movement by considering the reactivity and temperature distribution within the reactor core. The current FAST system is simulated under a single control rod withdrawal accident condition. In this investigation, the reactor thermal power does not return to its initial thermal power even if the FAST inserts negative reactivity. Only a 9 K of coolant temperature margin, in the hottest fuel assembly at EOL, can lead to unnecessary insertion of the negative reactivity. On the other hand, the FASTs cannot contribute to controlling the reactivity when normalized radial power is less than 0.889 at BOL and 0.972 at EOL. These simulation results suggest that the current FAST design needs to be optimized depending on its installed location. Meanwhile, the FAST system keeps the fuel, cladding and coolant temperatures below their limit temperatures with given conditions.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.564-569
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• 2003

During fuel irradiation tests, all parts of cylindrical structure with multiple holes act as heat sources due to fussion heal and ${\gamma}-flux$. The high temperature is especially generated over $2500^{\circ}C$ in the center of pellet. Due to the high temperature, many problems occur, such as melting of pellet and declining of heat transfer between cladding and coolant. [n this study, it is attempted 10 minimize the temperature of pellet using optimization method about geometric variables. For thermal and optimization analysis or structure. the finite element method code. ANSYS 5.7 is used. In this procedure. subproblem approximation method is used to the optimization methods. Through the optimum design process, the temperature of sealed basket type is reduced from $2537^{\circ}C$ to $2181^{\circ}C$ and the temperature of open basket type is reduced from $2560^{\circ}C$ to $2106^{\circ}C$.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.54-64
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• 1988

This paper assesses the models in the SCDAP code using the results of the FLHT-2 test. Calculations show that the SCDAP correctly predicts Ire temperatures, oxidation front movement, overall hydrogen generation and peak generation rate, internal fuel rod pressure, and cladding rupture due to ballooning. A comparison of the calculated results with measured data shows that two phase level is underpredicted, and that radiation heat transfer and auto-catalytic reaction temperature of zircaloy are overpredicted. These models are recommended to be modified. The analysis also shows that the simulation of the gap in a fuel rod improves the code prediction on core damage progression.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.381-387
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• 2003

Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull -from PWR spent fuel of 32, 000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4 -7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radio active waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull.

• Nuclear Engineering and Technology
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• v.23 no.4
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• pp.363-373
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• 1991

The current rod internal pressure criterion permits fuel rods to operate with internal pressures in excess of system pressure only if internal overpressure does not cause the diametral gap enlargement. In this study, the generic allowable internal gas pressure not violating this criterion is estimated as a function of rod power. The results show that the generic allowable internal gas pressure decreases linearly with the increase of rod power. Application of the generic allowable internal gas pressure for the rod internal pressure design criterion will result in the simplication of the current design procedure for checking the diametral gap enlargement caused by internal overpressure because according to the current design procedure the cladding creepout rate should be compared with the fuel swelling rate at each axial node at each time step whenever internal pressure exceeds the system pressure.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2299-2305
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• 2020

Zr alloy specimens were coated with Cr-Al alloy to enhance their resistance to oxidation. The coated samples were oxidized at 1200 ℃ in a steam environment for 300 s and showed extremely low oxidation when compared to uncoated Zr alloy specimens. The microstructure and elemental distribution of the oxides formed on the surface of Cr-Al alloys have been investigated by transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS). A very thin protective layer of Cr₂O₃ formed on the outer surface of the Cr-Al alloy, and a thin Al₂O₃ layer was also observed in the Cr-Al alloy matrix, near the surface. Our results suggest that these two oxide layers near the surface confers excellent oxidation resistance to the Cr-Al alloy. Even after exposure to a high temperature of 1200 ℃, inter-diffusion between the Cr-Al alloy and the Zr alloy occurred in very few regions near the interface. Analysis of the inter-diffusion layer by high-resolution transmission electron microscopy (HRTEM) and energy dispersive X-ray spectroscopy (EDS) measurement confirmed its identity as Cr₂Zr.