• Nuclear Engineering and Technology
• /
• v.56 no.5
• /
• pp.1887-1894
• /
• 2024

Fatigue strength of the structural materials of lead-cooled fast reactors (LFRs) and accelerator-driven systems (ADS) may be degraded in liquid metal (Lead or lead-bismuth eutectic (LBE)) environments. The fatigue crack growth (FCG) data of structural materials in liquid LBE are necessary for damage tolerance design, safety assessment and life management of key equipment. A novel monitoring system for fatigue crack length was designed on the compliance method and the monitor technology of crack opening displacement (COD) of CT specimens by the linear variable differential transformers (LVDT) system. It can be used to predict the crack length by monitoring the COD of CT specimens in harsh high-temperature liquid LBE using a LVDT system. The prediction accuracy of this system was verified by FCG experiments in room temperature air and liquid LBE at 150, 250 and 350 ℃. The first results obtained in the FCG test for T91 steel in liquid LBE at 350 ℃ are presented.

• Nuclear Engineering and Technology
• /
• v.52 no.12
• /
• pp.2880-2886
• /
• 2020

A small leak occurring on the surface of a fuel rod due to damage exposes UO₂ to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO₂ oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO₂. However, the thermodynamic calculation of the UO₂ oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO₂ pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO₂ oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO₂ pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1652-1657
• /
• 2021

Nuclear materials can be utilized not only for peaceful uses, but also for military purposes; hence, the international community has devoted itself to the control, management and safeguarding of nuclear materials. Nuclear materials are of varying degrees of usability for development of nuclear weapons. Thus, several methods for assessing the attractiveness of nuclear materials for nuclear weapons purposes have been proposed. When these methods are applied to unirradiated depleted, natural, and low-enriched uranium (DU, NU, and LEU), they are certainly classified as non-attractive nuclear materials. However, when nuclear material attractiveness is to be evaluated for potential radiological dispersal device (RDD) uses, it is required to develop a different method for the different aspects and factors. In the present study, we derived a novel method for evaluating nuclear material attractiveness for use in RDD development. To this end, the specific activity and dose coefficient were identified as the two sub-factors, and, in consideration of those, the mass causing detrimental health effects was determined to be the main factor impacting on nuclear materials attractiveness. Based on this factor, the attractiveness of unirradiated DU, NU, and LEU for RDD use was qualitatively compared with that of ¹³⁷Cs.

• Nuclear Engineering and Technology
• /
• v.50 no.7
• /
• pp.1131-1137
• /
• 2018

The effect of a serrated grain boundary on stress corrosion cracking (SCC) of Alloy 600 was investigated in terms of improvement of SCC resistance. Serrated grain boundaries and straight grain boundaries were obtained by controlled heat treatment. SCC cracks preferentially initiated and grew at grain boundaries normal to the tensile loading axis. Resolved tensile stress normal to the grain boundary was lower in serrated grain boundaries compared to straight grain boundaries. The specimen with serrated grain boundaries showed higher SCC resistance than that with straight grain boundaries due to a lower resolved tensile stress normal to the grain boundary.

• Advances in environmental research
• /
• v.6 no.4
• /
• pp.301-315
• /
• 2017

An Amberlite XAD-4 resin impregnated with di(2-ethylhexyl)phosphoric acid was prepared and its adsorption-desorption behaviors with Sr(II) ions under various conditions was examined. The resin was characterized by fourier transform infrared and thermal analysis techniques. The effects contact time, temperature, pH, interfering ions and eluants were studied. Results showed that adsorption of Sr (II) well fitted with pseudo-second-order kinetic model. The equilibrium adsorption data of Sr (II) on the impregnated resin were analyzed by Jossens, Weber-van Vliet, Redlich-Peterson and Fritz-Schlunder models to find out desirable equilibrium condition. Among them, the Fritz-Schlunder model best fitted to the experimental data. The maximum sorption capacity of impregnated resin amounted to 0.45 mg/ g at pH 8.0 and $20^{\circ}C$.

• Proceedings of the Korean Powder Metallurgy Institute Conference
• /
• 2006.04a
• /
• pp.21-21
• /
• 2006

• 한국초전도학회:학술대회논문집
• /
• v.13
• /
• pp.73-73
• /
• 2003

• 한국초전도학회:학술대회논문집
• /
• v.13
• /
• pp.74-74
• /
• 2003

• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.05a
• /
• pp.294-294
• /
• 2001

• 한국초전도학회:학술대회논문집
• /
• v.14
• /
• pp.96-96
• /
• 2004