• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.3
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• pp.7-12
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• 2012

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.27-35
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• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.