• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.28-35
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• 2021

The flow accelerated corrosion (FAC) is one of significant aging and degradation mechanism and can affect structural integrity of CANDU feeder pipes. Pipe burst can occur under normal operation pressure (min. 10 MPa) if wall-thinning of the feeder pipe due to FAC is accumulated. Previous studies considered simple shapes of feeder pipe with local wall-thinning in order to conservatively assess structural integrity of wall-thinned feeder pipe. In this paper, a new FE model is developed, having an actual shape of the feeder pipe (double bent) as well as the actual wall-thinning shape and location based on the in-service inspection result. Then, the burst pressure assessment of the wall-thinned feeder pipe is performed using lower bound limit load analysis considering elastic-perfectly plastic material. In addition, an improved formulation to predict the burst pressure of the wall-thinned feeder pipe is presented and the safety margin is compared with an existing assessment method.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2982-2989
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• 2021

One of the consequences of the 475 ℃ embrittlement of duplex stainless steels is the reduction of the resistance to localized corrosion. Therefore, the detection of this type of embrittlement before the material exhibits significant loss in toughness, and corrosion resistance is important to ensure the structural integrity of critical components under corrosion threats. In this research, conductivity measurements are performed using the alternating current potential drop (ACPD) technique with using a portable four-point probe as a nondestructive evaluation (NDE) method for detecting the embrittlement in a 2507 (UNS S32750) super duplex stainless steel (SDSS) aged at 475 ℃ from as-received condition to 300 h. The electric conductivity results were compared against two electrochemical tests namely double loop electrochemical potentiokinetic reactivation (DL-EPR) and critical pitting temperature (CPT). Mechanical tests and the microstructure characterized using scanning electron microscopy (SEM) imaging are conducted to track the progress of embrittlement. It is shown that the electric conductivity correlates with the changes in impact energy, microhardness, and CPT corrosion tests result demonstrating the feasibility of the four-point probe as a possible field-deployable method for evaluating the 475 ℃ embrittlement of 2507 SDSS.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.1
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• pp.11-19
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• 2016

This paper focuses on the estimation of durability and service-life of reinforced concrete structures in Wolsong Low- and intermediate-level wastes Disposal Center (WLDC) in Korea. There are six disposal silos located in the saturated environment. The silo concrete is degraded due to reactions with groundwater and chemical attacks, and finally it will lose its properties as a transport barrier. The infiltration of sulfate and magnesium, leaching of potassium hydroxide, and chlorine induced corrosion are the most significant factors for degradation of reinforced concrete structure in underground environment. From the result of evaluation of the degradation time for each factor, the degradation rate of the reinforced concrete due to sulfate and magnesium is $1.308{\times}10^{-3}cm/yr$, and it is estimated to take 48,000 years for full degradation while potassium hydroxide is leached in depth of less than 1.5 cm at 1,000 years after the initiation of degradation. In case of chlorine induced corrosion, it takes 1,648 years to initiate corrosion in the main reinforced bar and 2,288 years to reach the lifetime limit of the structural integrity, and thus it is evaluated as the most significant factor.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.688-706
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• 2024

Austenitic stainless steels (ASS) are extensively employed in various sectors such as nuclear, power, petrochemical, oil and gas because of their excellent structural strength and resistance to corrosion. SS304 and SS316 are the predominant choices for piping, pressure vessels, heat exchangers, nuclear reactor core components and support structures, but they are susceptible to stress corrosion cracking (SCC) in chloride-rich environments. Over the course of several decades, extensive research efforts have been directed towards evaluating SCC using diverse methodologies and models, albeit some uncertainties persist regarding the precise progression of cracks. This review paper focuses on the application of Acoustic Emission Technique (AET) for assessing SCC damage mechanism by monitoring the dynamic acoustic emissions or inelastic stress waves generated during the initiation and propagation of cracks. AET serves as a valuable non-destructive technique (NDT) for in-service evaluation of the structural integrity within operational conditions and early detection of critical flaws. By leveraging the time domain and time-frequency domain techniques, various Acoustic Emission (AE) parameters can be characterized and correlated with the multi-stage crack damage phenomena. Further theories of the SCC mechanisms are elucidated, with a focus on both the dissolution-based and cleavage-based damage models. Through the comprehensive insights provided here, this review stands to contribute to an enhanced understanding of SCC damage in stainless steels and the potential AET application in nuclear industry.

• Corrosion Science and Technology
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• v.3 no.5
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• pp.216-221
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• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.281-286
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• 2008

Wall thinning defect due to flow accelerated corrosion is one of major aging phenomena in most power plant industries, and it results in reducing load carrying capacity of the piping systems. A failure testing system was set up for real scale elbows containing various simulated wall thinning defects, and monotonic in-plane bending tests were performed under internal pressure to find out the failure behavior of thinned elbows. Various finite element models were generated and analysed to figure out and simulate the behavior for other thinning shapes and loading conditions. This paper presents the decreasing trends of load carrying capacity according to the thinning dimensions which were revealed from the investigation of finite element analysis results. A mechanical integrity evaluation model for thinned elbows was proposed, also. This model can be used to calculate the TES plastic load of thinned elbows for general internal pressure, thinning location, and in-plane bending direction.

• 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.

• 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.

• Structural Monitoring and Maintenance
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• v.6 no.3
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• pp.255-268
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• 2019

Civil infrastructures, such as bridges and tunnels are most important assets and their failure during service will have significant economic and social impact in any country. Behavior of a bridge can be evaluated only through actual monitoring/measurements of bridge members under the loads of interest. Theoretical analysis alone is not a good predictor of the ability of a bridge. In some cases, theoretical analyses can give less effect than actual since theoretical analyses do not consider the actual condition of the bridge, support conditions, level of corrosion and damage in members and connections etc. Hence actual measurements of bridge response should be considered in making decisions on structural integrity, especially in cases of high value bridges (large spans and major crossings). This paper describes in detail the experimental investigations carried out on an open web type steel railway bridge. Strain gages and displacement transducers were installed at critical locations and responses were measured during passage of locomotives. Stresses were evaluated and extrapolated to maximum design loading. The responses measured from the bridge were within the permissible limits. The methodology adopted shall be used for assessing the structural integrity of the bridge for the design loads.

• Corrosion Science and Technology
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• v.20 no.4
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.