• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.44-50
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• 2011

Turbulent mixing of hot and cold coolants is one of the possible causes of high cycle thermal fatigue in piping systems of nuclear power plants. A typical situation for such mixing appears in turbulent flow through a T-junction. Since the high cycle thermal fatigue caused by thermal striping was not considered in the piping fatigue design in several nuclear power plants, it is very important to evaluate the effect of thermal striping on the integrity of mixing tees. In the present work, before conducting detailed evaluation, three thermal striping evaluation methodology suggested by EPRI, JSME and NESC are analyzed. Then, a by-pass pipe connected to the shutdown cooling system heat exchanger is investigated by using these evaluation methodology. Consequently, the resulting thermal stresses and the fatigue life of the mixing tee are reviewed and compared to each other. Futhermore, the limitation of each methodology are also presented in this paper.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.

• Structural Engineering and Mechanics
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• v.72 no.1
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• pp.31-41
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• 2019

Structural integrity assessment of piping components is of paramount important for remaining life prediction, residual strength evaluation and for in-service inspection planning. For accurate prediction of these, a reliable fracture parameter is essential. One of the fracture parameters is stress intensity factor (SIF), which is generally preferred for high strength materials, can be evaluated by using linear elastic fracture mechanics principles. To employ available analytical and numerical procedures for fracture analysis of piping components, it takes considerable amount of time and effort. In view of this, an alternative approach to analytical and finite element analysis, a model based on relevance vector machine (RVM) is developed to predict SIF of part through crack of a piping component under fatigue loading. RVM is based on probabilistic approach and regression and it is established based on Bayesian formulation of a linear model with an appropriate prior that results in a sparse representation. Model for SIF prediction is developed by using MATLAB software wherein 70% of the data has been used for the development of RVM model and rest of the data is used for validation. The predicted SIF is found to be in good agreement with the corresponding analytical solution, and can be used for damage tolerant analysis of structural components.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.57-62
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• 2011

As NPPs' operating times increase, the integrity of nuclear components is continually degraded due to aging effects of systems, structures and components. In addition, for the case of continued operation beyond design life, additional aging effects occurred during the extended operating period lead to more degradation of the integrity of nuclear components. Therefore, it is very important to mange and evaluate the aging to secure the safety of NPPs. Wolsong unit 1 is approaching to its design life of 30 years in 2012. The license renewal documents for continued operation of Wolsong unit 1 Is under reviewing now. In this paper, regulatory technologies for continued operation of Wolsong unit 1 developed by KINS will be introduced. That technologies include the safety review guidelines, regulatory guides for aging management program and regulatory program for audit calculation.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.1
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• pp.17-26
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• 2011

CANDU reactor core is composed a few hundreds pressure tubes, which support and locate the nuclear fuels in the reactor. Each pressure tube provides pressure boundary and flow path of primary heat transport system in the core region. In order to guarantee the structural integrity of pressure tube flaws which can be found by in-service inspection, crack growth and fracture initiation assessment have to be performed. Stress intensity factors are important and basic information for structural integrity assessment of planar and laminar flaws (e. g. crack). This paper reviews and confirms the stress intensity factor of axial crack, proposed in CSA N285.8-05, which is an fitness-for-service evaluation code for pressure tubes in CANDU nuclear reactors. The stress intensity factors in CSA N285.8-05 were compared with stress intensity factors calculated by three methods (finite element results, API 579-1/ASME FFS-1 2007 Fitness-For-Service and ASME Boiler and Pressure Vessel Code Section XI). The effects of Poisson's ratio and anisotropic elastic modulus on stress intensity factors were also discussed.

• Journal of Welding and Joining
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• v.31 no.6
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• pp.1-7
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• 2013

In nuclear power plants, lined carbon steel pipes or PCCPs (pre-stressed concrete cylinder pipes) have been widely used for sea water transport systems. However, de-bonding of linings and oxidation of PCCP could make problems in aged NPPs (nuclear power plants). Recently at several NPPs in the United States, the PCCPs or lined carbon steel pipes of the sea water or raw water system have been replaced with HDPE (high density polyethylene) pipes, which have outstanding resistance to oxidation and seismic loading. ASME B&PV Code committee developed Code Case N-755, which describes rules for the construction of buried Safety Class 3 polyethylene pressure piping systems. Although US NRC permitted HDPE materials for Class 3 buried piping, their permission was limited to only 10-year operation because of several concerns including the quality of fusion zone of HDPE. In this study, various requirements for fusion qualification test of HDPE and some regulatory issues raised during HDPE application review in foreign NPPs are introduced.

• 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.

• Journal of the Korean GEO-environmental Society
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• v.11 no.11
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• pp.19-26
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• 2010

An influence factors for soundness evaluation of river levee include consisting embankment in case piping, permeability coefficient of ground, height of embankment, the width of crest, material characteristics of embankment and foundation ground, shape of embankment slope, an influence for penetration of rainfall or river water in case slope stability. In this study, it was operated a feasibility investigation of existing design result, stability evaluation for permeability coefficient use and permeability coefficient change of foundation ground to investigate an influence in line with permeability coefficient change for result of river levee penetration analysis. The evaluation results of influence factors, the permeability coefficient was used in design and it was evaluated influence in safety factor of piping. After the evaluation of influence factors, the permeability coefficient used in the design appears with the fact that differs in a design report about same soil.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.13-19
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• 2008

In order to predict a remaining life of a plant, it is necessary to select the components that are critical to the plant life. The remaining life of those components shall be evaluated by considering the aging effect of materials used as well as numerous factors. However, when evaluating reliability of nuclear structural components, some problems are quite formidable because of lack of information such as operating history, material property change and uncertainty in damage models. Accordingly, if structural integrity and safety are evaluated by the deterministic fracture mechanics approach, it is expected that the results obtained are too conservative to perform a rational evaluation of plant life. The probabilistic fracture mechanics approaches are regarded as appropriate methods to rationally evaluate the plant life since they can consider various uncertainties such as sizes and shapes of cracks and degradation of material strength due to the aging effects. The objective of this study is to evaluate the structural integrity for a reactor pressure vessel under the small break loss of coolant accident by applying the deterministic and probabilistic fracture mechanics. The deterministic fracture mechanics analysis was performed using the three dimensional finite element model. The probabilistic integrity analysis was based on the Monte Carlo simulation. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT.

• Journal of the Korea institute for structural maintenance and inspection
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• v.26 no.4
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• pp.30-39
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• 2022

Maintaining structural integrity of major apparatuses in a nuclear power plant, including piping system, is recognized as a critical safety issue. The integrity of piping system is also a critical matter related to the safety of a nuclear power plant. The actual failure mode of a piping system due to a seismic load is the leakage due to a fatigue crack, and the structural damage mechanism is the low-cycle fatigue due to large relative displacement that may cause plastic deformation. In this study, in-plane cyclic loading tests were conducted under various constant amplitudes using specimens composed of steel straight pipes and a steel pipe tee in the piping system of a nuclear power plant. The loading amplitude was increased to consider the relative displacement generated in the piping system under seismic loads, and the test was conducted until leakage, which is the limit state of the steel pipe tee, occurred due to fatigue cracks. The limit state of the steel pipe tee was expressed using a damage model based on the damage index that used the force-displacement relationship. As a result, it was confirmed that the limit state of the steel pipe tee can be quantitatively expressed using the damage index.