• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.78-82
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• 2010

The feeder pipe of Main Primary Heat Transfer System in Wolsong Nuclear Power Plant was inspected by the Ultrasonic Phase Array technique in 2010. It is the first time to apply this method to the construction at Nuclear Power Plant in Korea. The time required for UT technique is less than RT method. The UT method doesn't need to evacuate personnel who works nearby inspecting area and doesn't need to wait developing of film. For these reasons, the UT method is the fastest method among the volumetric inspections. As a result of the examination, it became clear that main defect of the feeder pipe is the Lack of fusion in the welded area. Moreover, the rate of defect was reduced gradually as improvement of welder's skill. If welding machine has problem, the defect has tended to same pattern(occurred same position in the welding area) but these defects were founded without specific rules. For these reasons, the creation of defect is dependent on the skill of worker not on the automatic welding machine. This evaluation of defect signal and collecting data would be useful to further examination in ISI.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.2
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• pp.14-20
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• 2011

The end of design lifetime for Wolsong unit 1 will be reached on 20th November in 2012. So the license renewal documents for the continuous operation of Wolsong unit 1 is under reviewing now. Major components of primary system such as pressure tubes, feeder pipes including delayed neutron monitoring system tubing are being replaced and many components of secondary system are also being repaired. In this paper, the assessment on the wear degradation of delayed neutron monitoring system tubing(on the other hand, DN tube was called) was performed for the ageing management of the same component. The wear defects of this component was one of causes that resulted in heavy water leakage accidents. Therefore design specifications of Wolsong uint 1 and heavy water leakage accidents of pressurized heavy water reactors were reviewed and causes of wear defect for DN tubes were analyzed. Wear propagation equations based on the heavy water leakage history were made and the proper repairing time was possible to be expected if the continued operation was considered. Finally design change items of DN tubes that were conducted for the long term operation of Wolsong unit 1 are introduced.

• Fire Science and Engineering
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• v.16 no.1
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• pp.24-29
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• 2002

The purpose of this study is developing the computer program for hydraulic design sprinkler systems have performed the means for the general use of network analysis method. The computer program is based on the theoretical concepts of the related Hazen-Williams equations, a modified Bernoulli equations, and the Hardy Cross method of pipe network analysis. Looped piping calculations are solved by using either the Hardy Cross method or the other iteration methods. While the other methods are solved using simultaneous equations, the Hardy Cross method is concerned with one loop at a time using reiterative process. Due to its simplicity the Hardy Cross method will be the primary method described in this thesis. The purpose of this study is to develope hydraulic calculation program by using algorithm for network analysis method. The development of computer program for the hydraulic design of sprinkler systems will perform the means in the performance-based sprinkler system design.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.48-55
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• 2013

In this paper, W-DCAP-HTR(Web-based Design Compatibility Assessment Program for High Temperature Reactor) which will be used to check the design criteria for high temperature reactor is newly proposed. To do this, the assessment procedure of the ASME Sec.III Div.5 such as time-dependent primary stress limit, accumulated inelastic strain, and creep-fatigue damage evaluation were investigated. Furthermore, the trend of candidate materials for high temperature reactor was also reviewed. Then, all assessment procedures for high temperature reactor have been computerized to enhance the efficiency and to reduce the possibility of human error during calculating procedure by hand calculation. It can be directly conducted by adopting the actual thermal and structural analysis results. The validation of W-DCAP-HTR has been demonstrated by benchmark analysis.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.115-125
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• 2016

The prototype SFR has currently been under design by KAERI. The size of its major components is much larger than that of APR1400 and high temperature materials are applied for it. The increased size of components and those specific materials effect on material procurement, manufacturing process and fabrication facilities. The manufacturing methods are studied for Reactor Vessel/Guard Vessel, Control Rod Drive Mechanism, Heat Exchanger, Primary Pump, Reactor Vessel Internals, Steam Generator and In-Vessel Transfer Machine. The proper manufacturing methods are suggested for each component including side forging technology for ultra large forgings of Reactor Vessel to minimize the weld seams on which In-service Inspection should be conducted.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.78-84
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• 2016

The FHX (Forced-draft sodium-to-air Heat Exchanger) employed in the ADHRS (active decay heat removal system) is a shell-and-tube type counter-current flow heat exchanger with M-shape finned-tube arrangement. Liquid sodium flows inside the heat transfer tubes and atmospheric air flows over the finned tubes. The unit is placed in the upper region of the reactor building and has function of dumping the system heat load into the final heat sink, i.e., the atmosphere. Heat is transmitted from the primary cold sodium pool into the ADHRS sodium loop via DHX (decay heat exchanger), and a direct heat exchange occurs between the tube-side sodium and the shell-side air through the FHX tube wall. This paper describes the DHRS and the structural design of the FHX.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.30-36
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• 2020

In SMART, the flow mixing header assembly (FMHA) is used to mix the coolant flowing into the reactor core to maintain a uniform temperature. The FMHA is designed to have enough stiffness so the resonance with reactor internal structures does not occurs during the pipe break and the seismic accidents. Since the gap between the FMHA and the core support barrel assembly is very narrow compared with the diameter of FMHA, the hydrodynamic mass effect acting on the FMHA is not negligible. Therefore the hydrodynamic mass characteristics on the FMHA are investigated to consider the fluid and structure interaction effects. The result of modal analysis for the dry and underwater conditions, the natural frequency of primary vibration mode for the horizontal direction is reduced from 136.67 Hz to 43.76 Hz. Also the result of frequency response spectrum seismic analysis for the dry and underwater conditions, the maximum equivalent stress are increased from 13.89 MPa to 40.23 MPa. Therefore, reactor internal structures located in underwater condition shall consider carefully the hydrodynamic mass effects even though they have sufficient stiffness required for performing its functions under the dry condition.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.534-537
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• 2008

In recent years, the dissimilar metal, Alloy 82/182 welds used to connect stainless steel piping and low alloy steel or carbon steel components in nuclear reactor piping system have experienced cracking due to primary water stress corrosion(PWSCC). It is well known that one reason of the cracking is the residual stress by the weld. But, it is difficult to estimate exactly weld residual stress due to many parameters of welding. In this paper, the analysis of 3 FEM models made by ABAQUS Code is performed to estimate exactly the weld residual stress on the dissimilar metal weld. 3 FEM models are Butt model, Repair model and Overlay model and are the plane.strain 2D model. The thermal analysis and the stress analysis are performed on each model and the residual stresses on each model were calculated and compared respectively. Also, the specimen of Butt model was made and the residual stresses were measured by X-Ray method and Hole Drilling Technique. These results were compared with the FEM result of Butt model.

• 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.