• Korean Journal of Materials Research
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• v.21 no.11
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• pp.596-603
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• 2011

The very high temperature gas reactor (VHTR) is one of the next generation nuclear reactors for its safety, long-term stability, and proliferation-resistance. The high operating temperature of over 800$^{\circ}C$ enables various applications with high energy efficiency. Heat is transferred from the primary helium loop to the secondary helium loop through the intermediate heat exchanger (IHX). The IHX material requires creep resistance, oxidation resistance, and corrosion resistance in a helium environment at high operating temperatures. A Ni-based superalloy such as Alloy 617 is considered as a primary candidate material for the intermediate heat exchanger. In this study, the microstructures of Alloy 617 crept in pure helium and air environments at 950$^{\circ}C$ were observed. The rupture time in helium was shorter than that in air under small applied stresses. As the exposure time increased, the thickness of outer oxide layer of the specimens clearly increased but delaminated after a long creep time. The depth of the carbide-depleted zone was rather high in the specimens under high applied stress. The reason was elucidated by the comparison between the ruptured region and grip region of the samples. It is considered that decarburization caused by minor gas impurities in a helium environment caused the reduction in creep rupture time.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.2
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• pp.103-111
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• 2014

In this study, the probabilistic internal pressure fragility analysis was performed by using the non-linear finite element analysis method. The target structure is one of the containment buildings of typical domestic pressurized water reactors(PWRs). The 3-dimensional finite element model of the containment building was developed with considering the large equipment hatches. To consider uncertainties in the material properties and structural capacities, we performed the sensitivity analysis of the ultimate pressure capacity with respect to the variation of four important uncertain parameters. The results of the sensitivity analysis were used to the selection of the probabilistic variables and the determination of their probabilistic parameters. To reflect the present condition of the tendon pre-stressing force, the data of the pre-stressing force acquired from the in-service inspections of tendon forces were used for the determination of the median value. Two failure modes(leak, rupture) were considered and their limit states were defined to assess the internal pressure fragility of target containment building. The internal pressure fragilities for each failure mode were evaluated in terms of median internal pressure capacity, high confidence low probability of failure(HCLPF) capacity, and fragility curves with respect to the confidence levels. The HCLPF capacity was 115.9 psig for leak failure mode, and 125.0 psig for rupture failure mode.

• Journal of Korean Foot and Ankle Society
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• v.6 no.1
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• pp.106-110
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• 2002

The rupture of the Achilles tendon is rather uncommon, but its incidence has been increasing. Main causes are usually due to direct injury or sudden indirect high energy trauma such as sports activity without predisposing disease. Spontaneous rupture of the Achilles tendon are sporadically reported especially from person who took steroid or with similiar predisposing disease. We experienced a patient with bilateral ruptures of the Achilles tendon that had occurred almost spontaneously, without any steroid related medication or underlying diseases.

• Bulletin of the Korean Chemical Society
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• v.34 no.11
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• pp.3223-3226
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• 2013

Spherical phospholipid-bilayers, vesicles, were prepared using the layer-by-layer double emulsion technique, which allows the bilayer to be formed asymmetrically. On the outer layer of the vesicles, the phospholipase D (PLD) reacted to convert phosphatidylcholine (PC) to phosphatidic acid (PA). The reaction induced the curvature change of the vesicles, which eventually led to rupture. The response time from the time of PLD injection to the time of rupture was measured against different vesicle curvatures and the outer layer phase, using the fluorescence intensity change of a pH-sensitive dye encapsulated within the vesicles. The effect of the vesicle curvature on the response was observed to be more significantly dramatic at the solid phase, compared to the liquid phase. Furthermore, in the solid phase, the response time was faster for 80 and 155 nm vesicles and, slower for 605 nm vesicles than similarly sized vesicles in the liquid phase vesicles. This difference in the response time was thought to result from the configuration determined by the phase difference and the PLD behavior.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.537-556
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• 2005

A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.161-170
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• 2022

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm^-1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm^-1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

• 한국금형공학회:학술대회논문집
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• 2008.06a
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• pp.39-44
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• 2008

Magnesium alloys have given high attention to the industry of light-weigh as automobile and electronics with aluminium, titanium and composite alloys due to their high strength, low specific density and good damping characteristics. But the magnesium contained structures under high temperature have the problems related to creep deformation and rupture life, which is a reason of developing the new material against creep deformation to use them safely. The purpose of this study is to predict the creep deformation mechanism and rupture time of AZ31 magnesium alloy. For this, creep tests of AZ31 magnesium alloy were done under constant creep load and temperature with the equipment including automatic temperature controller with acquisition computer. The apparent activation energy Qc and the applied stress exponent n, rupture life have been determined during creep of AZ31 Mg alloy over the temperature range of $150^{\circ}C$ to $300^{\circ}C$. In order to investigate the creep behavior. Constant load creep tests were carried out in the equipment including automatic temperature controller, whose data are sent to computer. At around the temperature of $150^{\circ}C{\sim}300^{\circ}C$ the creep behavior obeyed a simple power-law relating steady state creep rate to applied stress and the activation energy for the creep deformation was nearly equal and a little low, respectively, to that of the self diffusion of Mg alloy.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2008.11a
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• pp.555-556
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• 2008

• Journal of the Korean Society of Safety
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• v.20 no.2 s.70
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.