• Corrosion Science and Technology
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• v.6 no.3
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• pp.112-119
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• 2007

This paper presents the importance of corrosion control and protection technology with a real case study of heater tube rupture damaged by High temperature H2S-H2 corrosion in the refinery. The heater was operated at the Hydrocracking unit and the operation temperature and pressure was $340^{\circ}C$ and $18kg/cm^{3}$ respectively. Top side of the convection tube was thinned by high temperature hydrogen sulfide and hydrogen gas as a uniform corrosion and finally ruptured under operation pressure. Damaged area (Convection tube zone) was blocked by protection wall, so it was impossible to inspect with conventional nondestructive examination. Instead the elbow area which is out of the protection wall was inspected regularly to evaluate the corrosion rate of convection tube indirectly. However the operation temperature and the phase of the process stream was different between inside the chamber and outside the chamber. As a result, it caused severe corrosion to the horizontal convection tube inside the chamber comparing to the elbow outside the chamber. Finally convection tube was corroded more rapidly than the elbow and ruptured after 13 years operation. Because of the rupture, the heater was totally burned and the operation was stopped for 3 months until it has been reconstructed. To prevent this kind of corrosion problem and accident, corrosion control should be strengthened and protection technology should be improved.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.37-42
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• 2000

The boiler tubes and steam Pipes operating both at high temperature and pressure for a long period of time in a power plant are degraded by creep because of internal pressure. So, the remaining life of a component is evaluated by the creep rupture strength. Although the conventional method to evaluate the creep damage is widely used, it has some disadvantages such as requires large size specimen and long employed to evaluate the correlation between fracture toughness and evaluation time. Recently, new method so called "small lunch test' is used to evaluate degradation of creep. In this study, a conventional creep test and a small punch test are conducted using 2.25Cr-1Mo steel which is mainly used for the boiler tubes and steam pipes in power plant. The creep life, approximately 1,500 hrs, is determined by conventional method under a severe condition then specimens for a small Punch test are obtained after certain time intervals such as 1/4, 1/2 and 3/4 of final rupture time, respectively.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.12-19
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• 2015

As a measure of improving the mechanical properties of a diffusion bonded joint of a ferritic/martensitic steel (FMS), the post-bonding heat treatment (PBHT) is applied. In the temperature range of normalizing condition ($950-1,050^{\circ}C$), diffusion bonding is employed with compressive stress (6 MPa). Due to the martensite structure distributed in the matrix, Vicker's hardness values of the as-bonded are much higher than those of the as-received. Through the PBHT for 1 h at $720^{\circ}C$, hardness values are recovered to as low as those of the as-received condition. Also, tensile properties of PBHT are similar to those of the as-received at up to the test temperature of $550^{\circ}C$, when the diffusion bonding is carried out over $1,000^{\circ}C$. Based on the creep-rupture testing performed at $650^{\circ}C$ in air environment, the joint efficiency of the PBHTed specimens is about 80% in, which is higher than that of the as-bonded specimens.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.351-361
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• 2015

External reactor vessel cooling (ERVC) for in-vessel retention (IVR) has been considered one of the most useful strategies to mitigate severe accidents. However, reliability of this common idea is weakened because many studies were focused on critical heat flux whereas there were diverse uncertainties in structural behaviors as well as thermal-hydraulic phenomena. In the present study, several key factors related to molten corium behaviors and thermal characteristics were examined under multi-layered corium formation conditions. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative reactor pressure vessel (RPV) to figure out the possibility of high temperature induced failures. From the sensitivity analyses, it was proven that the reactor cavity should be flooded up to the top of the metal layer at least for successful accomplishment of the IVR-ERVC strategy. The thermal flux due to corium formation and the relocation time were also identified as crucial parameters. Moreover, three-layered corium formation conditions led to higher maximum von Mises stress values and consequently shorter creep rupture times as well as higher damage factors of the RPV than those obtained from two-layered conditions.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.353-358
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• 2004

Tracing the study of the burst test of steam generator tube, few studies have been reported to effect of external pressure acting on secondary-side in service condition. In this study the burst tests of Inconel 690TT were conducted in order to evaluate burst strength characteristics under the effect of external pressure. We obtained the result that the burst strength of Inconel 690TT increased when external pressure increased while both total circumferential elongation and uniform burst elongation were not affected. Also, according to the increased of external pressure, the size of the burst opening became smaller and the tear was getting severe.

• Geomechanics and Engineering
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• v.3 no.4
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• pp.255-266
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• 2011

Knowledge of seismic earth pressure against rigid retaining wall is very important. Mononobe-Okabe method is commonly used, which considers pseudo-static approach. In this paper, the pseudo-dynamic method is used to compute the distribution of seismic earth pressure on a rigid cantilever retaining wall supporting dry cohesionless backfill. Planar rupture surface is considered in the analysis. Effect of various parameters like wall friction angle, soil friction angle, shear wave velocity, primary wave velocity, horizontal and vertical seismic accelerations on seismic earth pressure have been studied. Results are presented in terms of tabular and graphical non-dimensional form.

• 한국연소학회:학술대회논문집
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• 2013.06a
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• pp.95-96
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• 2013

Numerical simulations are conducted to investigate the mechanism of spontaneous ignition of hydrogen within a certain length of downstream pipe released by the failure of pressure boundaries of various geometric assumption. The results show that local ignition is developed in limited area such as boundary layer and the mixing of hydrogen and air is weak at the planar pressure boundary conditions, whereas the flame fronts at the contact region are developed at the pressure boundaries of the spherical shape.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.917-925
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• 2022

Steam generator (SG) tubes in a nuclear power plant can undergo rapid changes in pressure and temperature during an accident; thus, an accurate model to predict short-term creep damage is essential. The theta (𝜃) projection method has been widely used for modeling creep-strain behavior under constant stress. However, many creep test data are obtained under constant load, so creep rupture behavior under a constant load cannot be accurately simulated due to the different stress conditions. This paper proposes a novel methodology to obtain the creep curve under constant stress using a modified 𝜃 projection method that considers the increase in true stress during creep deformation in a constant-load creep test. The methodology is validated using finite element analysis, and the limitations of the methodology are also discussed. The paper also proposes a creep-strain model for alloy 690 as an SG material and a novel creep hardening rule we call the damage-fraction hardening rule. The creep hardening rule is applied to evaluate the creep rupture behavior of SG tubes. The results of this study show its great potential to evaluate the rupture behavior of an SG tube governed by creep deformation.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.