• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.8
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• pp.1259-1269
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• 1997

The procedure for the determination of the residual life of the pressurizer surge line nozzle in the nuclear plant is developed. The design fatigue life for the 1800 $ft^3$ pressurizer surge line nozzle in cast head design is compared with that of Westinghouse stress report, and the percentage difference between two results is less than 9%. The design fatigue life evaluation of the 1000 $ft^3$ pressurizer surge line nozzle in fabricated head design is carried out, and the consuming rate and residual life are estimated using the operating data.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.94-97
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• 2008

In the welding process, weldments usually include repair weld during the manufacturing process. Repair welds is supposed to cause strong tensile residual stress. Moreover weldments, usually made by Alloy 82/182, is susceptible to PWSCC. Therefore, mitigation of welding residual stress in weldments is important for reliable operating. PWOL is one of the methods for mitigation and verified for over twenty years. In this paper, residual stress distribution of repaired weldments and the effect of PWOL on mitigation is examined for surge nozzle.

• Journal of Welding and Joining
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• v.29 no.2
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• pp.34-39
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• 2011

Nickel-based austenitic alloys such as Alloy 82 and 182 had been employed as the weld metals in nuclear power plants (NPPs) due to their high corrosion resistance as well as good mechanical properties. However, since the 2000s, the occurrence of primary water stress corrosion cracking has been reported in conjunction with these alloys in domestic and oversea NPPs. In the present work, we assumed an imaginary crack at the inner surface of a surge nozzle weld that had previously experienced the outside repair welding, and constructed its finite element model. Finite element analysis was performed with respect to the heat transfer, and then to the residual stress for obtaining the total applied stress distributions. These stress distributions were finally converted to the stress intensity factors for estimating crack growth rate. From the comparison of crack growth rate curves for the cases of no repair welding and outside repair welding, it was found that the outside repair welding did not exhibit negative effect on the crack growth for the surge nozzle under consideration in this work; in both cases, the cracks stopped growing before they became the through-wall cracks.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.22 no.6
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• pp.557-564
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• 2009

Welding residual stress is occurred after welding process. Tensile residual stress is one factor of PWSCC. Repair welding usually happened during the manufacturing welding process. Repair welds cause strong tensile residual stress. In PWR, Repair weldments made by Alloy 82/182 is susceptible to PWSCC caused by tensile stress, material and environment. Therefore, mitigation of welding residual stress in weldments is important for reliable operating. PWOL is one of the methods for mitigation and verified for over twenty years. In this paper, residual stress distribution of repaired weldments and the effect of PWOL on mitigation is examined for surge nozzle.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.69-76
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• 2018

The paper presents preliminary investigation results for the effect of the baseline correction in the acceleration excitation method on finite element seismic analysis results (such as accumulated equivalent plastic strain, equivalent plastic strain considering cyclic plasticity, von Mises effective stress, etc) of nuclear safety Class I components. For investigation, finite element elastic-plastic time-history seismic analysis is performed for a surge line including a pressurizer lower head, a pressurizer surge nozzle, a surge piping, and a hot leg surge nozzle using the Chaboche hardening model. Analysis is performed for various seismic loading methods such as acceleration excitation methods with and without the baseline correction, and a displacement excitation method. Comparing finite element analysis results, the effect of the baseline correction is investigated. As a result of the investigation, it is identified that finite element analysis results using the three methods do not show significant difference.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.1
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• pp.19-25
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• 2022

Nuclear Power Plants (NPP) should be designed to have sufficient safety margins and to ensure seismic safety against earthquake that may occur during the plant life time. After the 9.12 Gyeongju earthquake accident, the structural integrity of nuclear power plants due to the beyond design basis earthquake is one of key safety issues. Accordingly, it is necessary to conduct structural integrity evaluations for domestic NPPs under beyond design basis earthquake. In this study, the Level 3 LBB (Leak Before Break) evaluation was performed by considering the beyond design basis earthquake for the surge line of a OPR1000 plant of which design basis earthquake was set to be 0.2g. The beyond design basis earthquake corresponding to peak ground acceleration 0.4g at the maximum stress point of the surge line was considered. It was confirmed that the moment behaviors of the hot leg and pressurized surge nozzle were lower than the maximum allowable loading in moment-rotation curve. It was also confirmed that the LBB margin could be secured by comparing the LBB margin through the Level 2 method. It was judged that the margin was secured by reducing the load generated through the compliance of the pipe.

• Transactions of the Korean Society of Machine Tool Engineers
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• v.14 no.6
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• pp.88-94
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• 2005

A swirl ratio of a charge in the cylinder could be calculated by measuring both the rotary speed of paddle and the intake air flow rate in the swirl measurement apparatus fur several positions of valve lift. The automation of the swirl ratio measurement for a cylinder head is achieved by controlling both the valve lift of cylinder head and a suction pressure of the surge tank, instead of controlling them manually. PID control of the surge tank pressure and positioning a valve lift of the cylinder head are also achieved by using two step motors, respectively. Rotating speed of a paddle are measured using an optical sensor and a counter. Flow rate are measured from ISA 1932 flow nozzle by reading a differential pressure gauge position using IEEE-1394 camera. Time to measure the swirl ratio for a port in the cylinder head is drastically reduced from an hour to 3 minutes by automation control of the apparatus.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2004.03a
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• pp.513-520
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• 2004

A steady-state/transient performance simulation model was newly developed for the propulsion system of the CRW (Canard Rotor Wing) type UAV (Unmanned Aerial Vehicle) during flight mode transition. The CRW type UAV has a new concept RPV (Remotely Piloted Vehicle) which can fly at two flight modes such as the take-off/landing and low speed forward flight mode using the rotary wing driven by engine bypass exhaust gas and the high speed forward flight mode using the stopped wing and main engine thrust. The propulsion system of the CRW type UAV consists of the main engine system and the duct system. The flight vehicle may generally select a proper type and specific engine with acceptable thrust level to meet the flight mission in the propulsion system design phase. In this study, a turbojet engine with one spool was selected by decision of the vehicle system designer, and the duct system is composed of main duct, rotor duct, master valve, rotor tip-jet nozzles, and variable area main nozzle. In order to establish the safe flight mode transition region of the propulsion system, steady-state and transient performance simulation should be needed. Using this simulation model, the optimal fuel flow schedules were obtained to keep the proper surge margin and the turbine inlet temperature limitation through steady-state and transient performance estimation. Furthermore, these analysis results will be used to the control optimization of the propulsion system, later. In the transient performance model, ICV (Inter-Component Volume) model was used. The performance analysis using the developed models was performed at various flight conditions and fuel flow schedules, and these results could set the safe flight mode transition region to satisfy the turbine inlet temperature overshoot limitation as well as the compressor surge margin. Because the engine performance simulation results without the duct system were well agreed with the engine manufacturer's data and the analysis results using a commercial program, it was confirmed that the validity of the proposed performance model was verified. However, the propulsion system performance model including the duct system will be compared with experimental measuring data, later.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.62-69
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• 2013

A probabilistic assessment code, PRO-LOCA ver. 3.7 which was developed in an international co-operative research program, PARTRIDGE was evaluated by conducting sensitivity analysis. The effect of some variables such as simulation methods (adaptive sampling, iteration numbers, weld residual stress model), crack features(Poisson's arrival rate, maximum numbers of cracks, initial flaw size, fabrication flaws), operating and loading conditions(temperature, primary bending stress, earthquake strength and frequency), and inspection model(inspection intervals, detectable leak rate) on the failure probabilities of a surge line nozzle was investigated. The results of sensitivity analysis shows the remaining problems of the PRO-LOCA code such as the instability of adaptive sampling and unexpected trend of failure probabilities at an early stage.

• Journal of Mechanical Science and Technology
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• v.19 no.5
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.