• Corrosion Science and Technology
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• v.14 no.1
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• pp.1-11
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion (FAC), cavitation, flashing and/or liquid drop impingements, are a main concern in carbon steel piping systems of nuclear power plant in terms of safety and operability. Thinned pipe management program (TPMP) had been developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning in the secondary side piping system. This program also consists of several technical elements such as prediction of wear rate for each component, prioritization of components for inspection, thickness measurement, calculation of actual wear and wear rate for each component. Decision making is associated with replacement or continuous service for thinned pipe components. Establishment of long-term strategy based on diagnosis of plant condition regarding overall wall thinning is also essential part of the program. Prediction models of wall thinning caused by FAC had been established for 24 operating nuclear plants. Long term strategies to manage the thinned pipe component were prepared and applied to each unit, which was reflecting plant specific design, operation, and inspection history, so that the structural integrity of piping system can be maintained. An alternative integrity assessment criterion and a computer program for thinned piping items were developed for the first time in the world, which was directly applicable to the secondary piping system of nuclear power plant. The thinned pipe management program is applied to all domestic nuclear power plants as a standard procedure form so that it contributes to preventing an accident caused by FAC.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.264-267
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• 2004

Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle piping systems in Nuclear Power Plants (NPP). Since the mid-1990s, secondary side piping systems in Korean NPPs have experienced wall thinning, leakages and ruptures caused by FAC. Korea Electric power Research Institute (KEPRI) and Korea Hydro & Nuclear Power Co., LTD. (KHNP) have conducted a study to develop the methodology for systematic pipe management and established the Korean Thinned Pipe Management Program (TPMP). To effectively maintain the integrity of piping system, FAC engineer should understand the criterions of the structural integrity evaluation and the safety margin assessment for the thinned pipe component. This paper describes the technical items of TPMP, and shows the example of the integrity evaluation and safety margin assessment for three thinned pipe component of a NPP.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.33-41
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• 2012

The component cooling water heat exchangers are critical components in a nuclear power plant. As the operation years of the heat exchanger go by, the maintenance costs required for continuous operation also increase. Most heat exchangers have carbon steel shells, tube support plates and flow baffles. The titanium tube is susceptible to flow induced vibration. The damage on carbon steel tube support rod and titanium tube around cooling water entrance area is inevitable. Therefore, analysis of internal flow around the component cooling water entrance and tube channel is a good opportunity to seek for failure prevention practice and maintenance method. The numerical study was carried out by FLUENT code to find out the causes of tube failure and its location.

• Corrosion Science and Technology
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• v.11 no.6
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• pp.218-224
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• 2012

Liquid droplet impingement erosion (LDIE) known to be generated in aircraft and turbine blades is recently appeared in nuclear piping. UT thickness measurements with both A-scan and B-scan UT inspection equipments were performed for a component estimated as susceptible to LDIE in feedwater heater vent system. The thickness data measured with B-Scan equipment were compared with those of A-Scan. Thermal hydraulic analysis based on ANSYS FLUENT code was performed to analyze the behavior of liquid droplets inside piping. The wall thinning rate and residual lifetime based on both existing Sanchez-Caldera equation and measuring data were also calculated to identify the applicability of the existing equation to the LDIE management of nuclear piping. Because Sanchez-Caldera equation do not consider the feature of magnetite formed inside piping, droplet size, colliding frequency, the development of new evaluation method urgently needs to manage the pipe wall thinning caused by LDIE.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.349-359
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• 2017

During the past 25 years, in the context of probabilistic safety assessment, efforts have been directed towards establishment of comprehensive pipe failure event databases as a foundation for exploratory research to better understand how to effectively organize a piping reliability analysis task. The focused pipe failure database development efforts have progressed well with the development of piping reliability analysis frameworks that utilize the full body of service experience data, fracture mechanics analysis insights, expert elicitation results that are rolled into an integrated and risk-informed approach to the estimation of piping reliability parameters with full recognition of the embedded uncertainties. The discussion in this paper builds on a major collection of operating experience data (more than 11,000 pipe failure records) and the associated lessons learned from data analysis and data applications spanning three decades. The piping reliability analysis lessons learned have been obtained from the derivation of pipe leak and rupture frequencies for corrosion resistant piping in a raw water environment, loss-of-coolant-accident frequencies given degradation mitigation, high-energy pipe break analysis, moderate-energy pipe break analysis, and numerous plant-specific applications of a statistical piping reliability model framework. Conclusions are presented regarding the feasibility of determining and incorporating aging effects into probabilistic safety assessment models.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.1
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• pp.25-32
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• 2012

Korea Hydro & Nuclear Power Co. (KHNP) has developed and implemented the maintenance effectiveness monitoring (MR) programs for the operating nuclear power plants. The MR program is developed by reflecting design characteristics of the operating nuclear power plants to monitor the plant performance for improving the safety and reliability. Recently, KHNP has built a new nuclear power plant, and developed the MR program to establish the advanced maintenance system by reflecting unique design characteristics based on the OPR1000 standard model. So, the MR program developed in this study has another characteristics in comparison with the OPR1000 standard model, and we will verify the suitability of the MR program through evaluating initial performance of the plant. The safety and reliability of the new plant will be improved by developing and implementing the MR program.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.25-34
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• 2020

As leakage in nuclear power plants could cause a variety of problems, it is very critical to monitor leakage from the safety point of view. Accordingly, a new type of leak detection system is currently being developed and flow characteristics of the sampling and transportation system are investigated by using numerical analysis as a part of the development process in this study. The results showed that the steam mass fraction varied according to the effect of the gap between the insulation and piping component, transportation velocity, and material properties of porous media during the sampling and transportation process. The results of this study should be useful for understanding flow characteristics of the sampling and transportation system and its design and application.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.1-6
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. A co-axial double-tube primary hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the VHTR. In this study, a preliminary design analysis for the primary HGD of the nuclear hydrogen system was carried out. These preliminary design activities include a determination of the size, a strength evaluation and an appropriate material selection. The determination of the size was undertaken based on various engineering concepts, such as a constant flow velocity model, a constant flow rate model, a constant hydraulic head model, and finally a heat balanced model.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.48-53
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• 2017

We diagnosed the defect using the data obtained from the nuclear power plant simulator. In this paper, we diagnosed faults in the nuclear power plant system for discovery instead of the traditional single-component or device unit. We created the six fault scenarios and used a fault simulator to obtain the fault data. It was extracted pattern from acquired failure data. Neural network model was trained and simple pattern matching algorithm was applied. We presented a simulation result and confirmed that the applied algorithm works correctly.