• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.265-276
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• 2013

Pressure tubes made of Zr-2.5 wt% Nb alloy are important components consisting reactor coolant pressure boundary of a pressurized heavy water reactor, in which unanticipated through-wall cracks and rupture may occur due to a delayed hydride cracking (DHC). The Canadian Standards Association has provided deterministic and probabilistic structural integrity evaluation procedures to protect pressure tubes against DHC. However, intuitive understanding and subsequent assessment of flaw behaviors are still insufficient due to complex degradation mechanisms and diverse influential parameters of DHC compared with those of stress corrosion cracking and fatigue crack growth phenomena. In the present study, a deterministic flaw assessment program was developed and applied for systematic integrity assessment of the pressure tubes. Based on the examination results dealing with effects of flaw shapes, pressure tube dimensional changes, hydrogen concentrations of pressure tubes and plant operation scenarios, a simple and rough method for effective cooldown operation was proposed to minimize DHC risks. The developed deterministic assessment program for pressure tubes can be used to derive further technical bases for probabilistic damage frequency assessment.

• Journal of the Korean Institute of Gas
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• v.20 no.5
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• pp.112-117
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• 2016

The standard(KS D 3631) should be obey for using fuel gas pipe in Korea and a carbon steel pipe is coated with synthetic resine for protecting the corrosion. The coating is just performed for anti-corrosion, and actual strength of pipe is given by the carbon steel pipe when the fuel gas is transported. When the flaws are occurred in the steel pipe, the estimation method which is used for residual strength is necessary. But the suitable method is not exist for applying the fuel gas pipe. Eventually, the residual strength is estimated by overseas regulation such as ASME B31G or DNV RP F101. But the method based on the relative regulations are not sufficiently valid and it can not estimate the 85% over even. Therefore, the test was performed for validation with pipe specimen. The specimen was made with artificial flaw. The test results showed the certain differences according to flaw depth and DNV RP F101 is suitable to estimate the residual strength in Korea. The results in this paper contain in case of the severe flaw depth and suitable mode for Korea, so the result can be expected as valuable and widely used in various field.

• Smart Structures and Systems
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• v.19 no.4
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• pp.371-382
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• 2017

Concrete filled steel tubes are extensively applied in engineering structures due to their resistance to high tensile and compressive load and convenience in construction. But one major flaw, their vulnerability to environmental attack, can severely reduce the strength and life of these structures. Degradation due to corrosion of steel confining the concrete is one of the major durability problems faced by civil engineers to maintain these structures. The problem accelerates as inner surface of steel tube is in contact with concrete which serves as electrolyte. If it remains unnoticed, it further accelerates and can be catastrophic. This paper discusses a non-destructive degradation monitoring technique for early detection corrosion in steel tubes in CFST members. Due to corrosion, damage in the form of debonding and pitting occurs in steel sections. Guided ultrasonic waves have been used as a feasible and attractive solution for the detection and monitoring of corrosion damages in CFST sections. Guided waves have been utilized to monitor the effect of notch and debond defects in concrete filled steel tubes simulating pitting and delamination of steel tubes from surrounding concrete caused by corrosion. Pulse transmission has been used to monitor the healthy and simulated damaged specimens. A methodology is developed and successfully applied for the monitoring of concrete filled steel tubular sections undergoing accelerated chloride corrosion. The ultrasonic signals efficiently narrate the state of steel tube undergoing corrosion.

• Journal of the Korean Society for Nondestructive Testing
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• v.23 no.1
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• pp.38-44
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• 2003

Performance demonstration with real flawed specimens has been strongly required for nondestructive evaluation of safety class components in nuclear power plant. Mechanical or thermal fatigue crack and intergranular stress corrosion cracking could be occured in the in-service nuclear power plant and mechanical fatigue crack was selected to study in this paper. Specimen was designed to produce mechanical fatigue flaw under tensile stress. The number of cycles and the level of stress were controlled to obtain the desired flaw roughness. After the accurate physical measurement of the flaw size and location, fracture surface was seal-welded in place to ensure the designed location and site. The remaining weld groove was then filled by using gas-tungsten are welding(GTAW) and flux-cored arc welding(FCAW). Results of radio graphic and ultrasonic testing showed that fatigue cracks were consistent with the designed size and location in the final specimens.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.16-22
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• 2010

Cast austenitic stainless steel(CASS) is used in the primary cooling piping system of nuclear power plant for it's relative low cost, corrosion resistance and easy of welding. However, the coarse-grain structure of cast austenitic stainless steel can strongly affect the inspectability of ultrasonic testing. The major problems encountered during inspection are beam skewing, high attenuation and high background noise of CASS component. So far, the best inspection performance involving CASS components have been achieved using low frequency TRL(Transmitter/Receiver side-by-side L wave) angle beam probe. But TRL technique could not detect shallow defect and it contains an uncertainty for sizing capability. Currently, most of researchers are studying to overcome these challenge issue. In this study, low-frequency phased array TRL technique used to detect and sizing the flaws in CF8A cast austenitic stainless steel.As conclusion, we could detect and size not only axial flaw but also circumferential flaw using low frequency phased array technique.

• Proceedings of the Korean Society of Machine Tool Engineers Conference
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• 2005.05a
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• pp.216-227
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• 2005

The head penetrations for control rod drive mechanism and instrumentation systems are installed at the reactor pressure vessel head of PWRs. Primary coolant water and the operating conditions of PWR plants can cause cracking of these nickel-based alloy through a process called primary water stress corrosion cracking (PWSCC). Inspection of the head penetrations to ensure the integrity of the head penetrations has been interested since reactor coolant leakages were found at U. S. reactors in 2000 and 2001. The complex geometry of the head penetrations and the very low echo amplitude from the fine, multiple flaws due to the nature of the see made it difficult to detect and size the flaws using conventional pulse-echo UT methods. Time-of-flight-diffraction technique, which utilizes the time difference between the flaw tips while pulse-echo does the flaw response amplitude from the flaw, has been selected for this inspection for it's best performance of the detection and sizing of the head penetration see flaws. This study defines the limits of the detectable and accurately sizable minimum flaw size which can be detected by the General TOFD and the Delta TOFD techniques for circumferentially and axially oriented flaws respectively. These results assures the reliability of the inspection techniques to detect and accurately size for various kind of flaws, and will also be utilized for the future development and qualifications of the TOFD techniques to enhance the detecting sensitivity and sizing accuracy of the flaws of the reactor head penetrations in nuclear power plants.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2010.05a
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• pp.779-780
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• 2010

• Proceedings of the KWS Conference
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• 2004.05a
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• pp.308-310
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• 2004

1980년대 초 미국의 비등형 경수로(Boiling Water Reactor : BWR) 원자력발전소 배관계통의 입계 응력 부식 균열(Inter-Granular Stress Corrosion Crack) 검사 결과 및 미국 EPRI(Electric Power Research Institute)에서 실시한 round robin test 결과에서 기존 초음파 검사 방법의 실효성에 많은 문제점이 제기 되었다. (중략)

• Korean Journal of Materials Research
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• v.25 no.10
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• pp.516-522
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• 2015

Aging aircraft structures are inevitably exposed to environment for a long time facing many potential problems, including corrosion and wide spread fatigue damage, which in turn cause the degradation of flight safety. In this study, the environmental surface damages on aging aircraft structures induced during service were quantitatively analyzed. Additionally, S-N fatigue tests were performed with center hole specimens extracted from aging aircraft structures. From the results of quantitative analyses of the surface damages and fatigue tests, it is concluded that corrosion pits initiated during service reduce the fatigue life significantly. Finally, using the fracture mechanics and the EIFS (equivalent initial flaw size) concepts, the remaining fatigue life was predicted based on actual fatigue test results.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.2
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• pp.275-282
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• 2001

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement. Number of subclad cracks may be found during an in-service-inspection due to the presence of cladding. It is specified, in ASME Sec. XI, that a subclad crack is characterized as a surface crack when the thickness of the clad is less than 40% of the crack depth. This condition is provided to keep the crack integrity evaluation conservative. In order to refine the fracture assessment procedures for such subclad cracks under a pressurized thermal shock condition, three dimensional finite element analyses are applied for various subclad cracks existing under cladding. A total of 36 crack geometries are analyzed, and the results are compared with those for surface cracks. The resulting stress intensity factors for subclad cracks are 6 to 44% less than those for surface cracks. It is proven that the flaw characterization condition as specified in ASME Sec. XI can be overly conservative for some subclad cracks.