• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.149-154
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• 2009

The microstructure, the corrosion behavior and the oxide properties were examined for Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr (HANA-4) alloys which were subjected to two different final annealing temperatures: $470^{\circ}C$ and $570^{\circ}C$. HANA-4 was shown to have $\ss$-enriched phase with a bcc crystal structure and Zr(Nb,Fe,Cr)$_2$ with a hcp crystal structure with $\ss$-enriched phase being more frequently observed compared with Zr(Nb,Fe,Cr)$_2$. The corrosion rate of HANA-4 was increased with an increase of the final annealing temperature in the PWR-simulating loop, $360^{\circ}C$ pure water and $400^{\circ}C$ steam conditions, which was correlated well with a reduction in the size of the columnar grains in the oxide/metal interface region. The oxide growth rate of HANA-4 was considerably affected by the alloy microstructure determined by the final annealing temperature.

• Journal of the Korean Society for Precision Engineering
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• v.32 no.9
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• pp.815-824
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• 2015

In this study, we conducted the approximate multi-objective optimization of gap sizes of pressurized-water reactor (PWR) annular fuels. To determine the contacting tendency of the inner-outer gaps between the annular fuel pellets and cladding, thermoelastic-plastic-creep (TEPC)analysis of PWR annular fuels was performed, using in-house FE code. For the efficient heat transfer at certain levels of stress, we investigated the tensile, compressive hoop stress and temperature, and optimized the gap sizes using the non-dominant sorting genetic algorithm (NSGA-II). For this, response surface models of objective and constraint functions were generated, using central composite (CCD) and D-optimal design. The accuracy of approximate models was evaluated through $R^2$ value. The obtained optimal solutions by NSGA-II were verified through the TEPC analysis, and we compared the obtained optimum solutions and generated errors from the CCD and D-optimal design. We observed that optimum solutions differ, according to design of experiments (DOE) method.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.294-299
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• 1994

Margin benefits are quantatively assessed when a Digital Monitoring System(DMS) is assumed to be installed to an operating Westinghouse analog type plant. Applied plant and cycle is YongGwang Unit 1 Cycle 6. The referenced digital monitoring system is the COLSS (Core Operating Limit Supervisory System) of ABB-CE. Considered fuel design limits are DNBR and LDCA Fq. 2003-D Power distributions within the present CAOC (Constant Axial Offset Control) limits are calculated for the analysis. The most limiting DNB prevention event of CEA Withdrawal is analyzed with the ROPM (Required OverPower Margin) concept of ABB-CE. The result show that the DMS can bring around 7% more margins for both DNB and LOCA Fq standpoints of view. The DMS can also monitor the PCI (Pellet-Cladding Interaction) limits.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.153-162
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• 1980

The investigation of the fuel cladding temperature behavior and heat transfer mechanism during the reflooding phase of a LOCA plays an important role in performance evaluation of ECCS and safety analysis of water reactors. Reflooding experiments were performed with horizontal and vertical flow channels to investigate the effect of coolant flow channel orientation on rewetting process. Emphasis was mainly placed on the CANDU reactor which has horizontal pressure tubes in core, and the results were compared with those of vertical channel. Also to investigate the rewetting process visually, the experiments by using a rod in annulus and a quartz tube heated outside were performed. It can be concluded that the rewetting velocity in horizontal flow channel is clearly affected by flow stratification, however, the average rewetting velocity is similar to those in vertical flow channel for same conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.55-61
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• 2013

The recovery of Zr from Zircaloy-4 (Zry-4) cladding hulls using a chlorination method was demonstrated for complete conversion of Zr into $ZrCl_4$. A chlorination reaction was performed by reacting Zry-4 hulls for 8 h under a 70 cc/min $Cl_2$ + 70 cc/min Ar flow at $380^{\circ}C$. The initial weight of the reactant (51.7 g) decreased to 0.49 g after 8 h of operation, which is only 0.95wt% of the initial weight. The weight of the total reaction products was 121.7 g with a high Zr purity of 99.80wt%. Fe and Sn were identified as major (0.18wt%) and minor (0.02wt%) impurities of the reaction products, respectively. It was also shown that Zr exhibited a high recovery ratio of 96.95wt% with a relatively small experimental loss of 2.34wt%. Observation of the reaction residues revealed that the chlorination reaction was dominant along the longitudinal direction, and surface oxide layers remained as reaction residues. The high purity and recovery ratio of Zr proposed the feasibility of the chlorination technique as an effective hull waste treatment method.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.89-98
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• 1989

The expanding copper mandrel test performed at three strain rates (3.2$\times$10E -5/s, 2.0$\times$10E-6/s and 1.2$\times$10E-7/s) over 553-873 K temperature range by varying the heating rates (8-1$0^{\circ}C$/s, 1-2$^{\circ}C$/s and 0.5$^{\circ}C$/s) in air and in vacuum (5$\times$10E-5 torr). The yield stress peak, the strain rate sensitivity minimum and the activation volume peaks could be explained in terms of the dynamic strain aging. The activation energy for dynamic strain aging obtained from the yield stress peak temperature and strain rate was 196 KJ/mol and this value was in good agreement with the activation energy for oxygen diffusion in $\alpha$-zirconium and Zircaloy-2 (207-220 KJ/mol). Therefore, oxygen atoms are responsible for the dynamic strain aging which appeared between 573 K and 673 K. The yield stress increase due to the oxidation was obtained by comparing the yield stress in air with that in vacuum and represented by the percentage increase of yield stress ( $\sigma$$^{a}$ $_{y}$ - $\sigma$$^{v}$ $_{y}$ / $\sigma$$^{v}$ $_{y}$ ). The slower the strain rate, the greater the percentage increase occurs. In order to estimate the yield stress of PWR fuel cladding material under the service environment, the yield stress in water was obtained by comparing the oxidation rate in air that in water assuming the relationship between the oxygen pick-up amount and the yield stress increase.

• Journal of the Korean Ceramic Society
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• v.50 no.2
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• pp.122-126
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• 2013

The surface roughness of the inner and outer surfaces of a tube is an important requirement for nuclear fuel cladding. When an inner SiC clad tube, which is considered as an advanced Pressurized Water Cooled Reactor (PWR) clad with a three-layered structure, is fabricated by Chemical Vapor Deposition (CVD), the surface roughness of the substrate, graphite, is an important process parameter. The surface character of the graphite substrate could directly affect the roughness of the inner surface of SiC deposits, which is in contact with a substrate. To evaluate the effects of the surface roughness changes of a substrate, SiC deposits were fabricated using different types of graphite substrates prepared by the following four polishing paths and heat-treatment for purification: (1) polishing with #220 abrasive paper (PP) without heat treatment (HT), (2) polishing with #220 PP with HT, (3) #2400 PP without HT, (4) polishing with #2400 PP with HT. The average surface roughnesses (Ra) of each deposited SiC layer are 4.273, 6.599, 3.069, and $6.401{\mu}m$, respectively. In the low pressure SiC CVD process with a graphite substrate, the removal of graphite particles on the graphite surface during the purification and the temperature increasing process for CVD seemed to affect the surface roughness of SiC deposits. For the lower surface roughness of the as-deposited interlayer of SiC on the graphite substrate, the fine controlled processing with the completed removal of rough scratches and cleaning at each polishing and heat treating step was important.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.119-128
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• 2008

Thermal assessment of a new CANDU spent fuel disposal system, which improves the retrievability of the spent fuel and enhances the densification factor compared with the Korean Reference disposal System, is carried out in this study. The canisters for CANDU spent fuels are stored for long term and cooled by natural convection in the proposed disposal system for the retrievability. The steady state thermal analyses for proposed CANDU disposal system are carried out with the ANSYS 10.0 CFX code. The thermal analyses are performed through two steps. At the first step, the sensitivity of the disposal tunnel spacing is analysed. The differences of maximum temperatures by several tunnel spacings are calculated at three points in the disposal tunnel. The result shows that the differences of the temperature at the three points are almost negligible because 99% of the decay heat is removed by natural convection. At the second procedure, 60m tunnel spacing with a ventilation system instead of natural convection is considered. The result is applied to the calculation of the canister surface temperature in disposal tunnel as boundary conditions. Consequently, the average and the maximum surface temperature of disposal canisters are $79.9^{\circ}C$ and $119^{\circ}C$, respectively. The inner maximum temperature of a basket in the disposal canister is calculated as $140.9^{\circ}C$. The maximum temperature of the basket meets the thermal requirement for the CANDU spent fuel cladding.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.11-19
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• 2015

The production of nuclear fuel cladding tube is expected to increase with the nuclear power plant expansion. Zirconium(Zr) scrap that is generated during manufacturing is also expected to increase. Zr electrorefining experiment was carried out in the fluoride salt of LiF-KF-ZrF₄ using multiple electrode for scale up and improving throughput Zr electrorefiner develop-ment. The Zr reduction peak observed at-0.8 V(vs.Ni). Polarization behavior showed that the amount of applied current increases because of decreasing cell resistance as the number of cathode increases. Experimental results showed the highest recovery rate about 98% at lowest current density of 25.64 mA/cm² using 6 electrodes. XRD and TG analysis result show that pure Zr was recovered 99.92% and ICP analysis shows that lower impurity content than conventional impurity content of the Anode(97.8%). Electrorefining consumes energy about 7.15 kWh/kg less than 39.7% compared to the Kroll process using 6 electrode width of 20 mm and height of 65 mm. Because of increasing cell efficiency and recovery rate, using multiple cathode is determined as an efficient technique for scale up electrorefining Zr scrap.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.125-133
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• 2017

KAERI is conducting research on spent cladding hulls and additive metals to generate a solidification host matrix for the noble metal fission product waste in anode sludge from the electro-refining process to minimize the volume of waste that needs to be disposed of. In this study, alloy compositions Zr-17Cr, Zr-22Cr, and Zr-27Cr were prepared with or without eight noble metals representing fuel waste using induction melting. The microstructures of the resulting alloys were characterized and electrochemical corrosion tests were conducted to evaluate their corrosion characteristics. All the compositions had better corrosion characteristics than other Zr-based alloys that were evaluated for comparison. Analysis of the leach solution after the corrosion test of the Zr-22Cr-8NM specimen indicated that the noble metals were not leached during corrosion under 500 mV imposed voltage, which simulates a highly oxidizing disposal environment. The results of this study confirm that Zr-Cr based compositions will likely serve as chemically stable waste forms.