• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.9-13
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• 2009

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS-56.8(1994) in Korea. Two methods, the make-up flow rate and the pressure decay, are used for LLRT. Though ANSI/ANS-56.8 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the test period for type "C" LLRT is differently applied to each NPPs. Therefore, this study presents a unified test criteria for data stabilization and test duration through experiments to improve the test reliability for type "C".

• Journal of the Korean Society for Nondestructive Testing
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• v.9 no.2
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• pp.42-49
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• 1989

It has not been performed to inspect the underclad cracking in Korea nuclear plant since there is no Code Requirements for inspection. However, underclad cracks in nuclear pressure vessels were reported firstly in 1970. The objection of this study is to be established the ultrasonic inspection techniques for underclad cracking. The ultrasonic inspection of bimetalic stainless steel weld is very difficult by high attenuation and multiple scattering at weld surface and weld/base metal interface. The various inspection methods using $70^{\circ}$ refracted longitudinal wave, 50/70 tandem transducer, $45^{\circ}\;and\;60^{\circ}$ single shear wave are compared. Experiments on limited specimens applied same condition to nuclear pressure vessels shows that $70^{\circ}$ refracted longitudinal wave method is the best one for the detection of underclad cracks. 50/70 tandem transducer using SPOT(Satellite Pulse Observation Technique) is more effective for underclad crack sizing than other sizing methods.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.79-86
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• 2015

The purpose of this study is to evaluate structural integrity of a weldment which is partially screwed and then welded. Two finite element models are constructed and solved: operating temperature and internal pressure are considered in the first simple model, and welding process and normal operating condition including heat-up process are simulated in the second model. Structural integrity assessment criteria are satisfied with both finite element models, therefore the repair weldment finely sustains structural integrity of this assembly and prevents leakage. Stresses are dramatically increased when weld residual stress is considered, but it should be considered as a secondary stress.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.42-47
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• 2012

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS 56.8-1994 in Korea. Two methods, the make-up flow rate and the pressure decay, are used for local leakage rate testing. Though ANSI/ANS 56.8-1994 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the testing time is differently applied to each NPPs. Therefore, this study presents a standardized test procedure for data stabilization and testing time through experiments to improve the test reliability.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.1
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• pp.27-34
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• 2011

The objective of this study is to classify the geometry of wall-thinning defect that causes a circumferential crack in the pipe elbows subjected to internal pressure. For this objective, first of all a criterion to determine the occurrence of circumferential cracking at wall-thinned area was developed based on finite element simulation for burst tests of pipe elbow specimens that showed axial and circumferential cracking at wall-thinned area. In addition, parametric finite element analysis including various wall-thinning geometries, locations, and pipe geometries was conducted and the wall-thinning geometries that initiate circumferential crack were determined by applying the criterion to the results of parametric analysis. It showed that the circumferential crack occurs at wall-thinning defect, which has a deep, wide, and short geometry. Also, it is indicated that the pipe elbows with larger radius to thickness ratio are more susceptible to circumferential cracking at wall-thinned area.

• 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.2
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• pp.31-37
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• 2017

The heat transfer function or thermal performance is the most important function of the steam generator component in nuclear power plants. The declining of thermal performance, fouling does not affect the electric power of the nuclear power plant within a certain fouling level, but it affects the output when goes beyond the governor valve wide open of the turbine. The VWO steam pressure can be predicted through the thermal performance evaluation of steam generators in the nuclear power plant. In consideration of the fouling characteristics of the steam generator, methods of the thermal performance evaluation and fouling cases are reviewed, and also the critical VWO value is estimated through the actual thermal performance evaluation. It is necessary to apply the VWO theory based on the thermal performance of the steam generators.

• 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.

• Journal of the Korean Society for Precision Engineering
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• v.29 no.9
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• pp.938-945
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• 2012

To design and estimate material failures of Type III pressure vessels, which have excellent stability and performance, various modeling techniques have been introduced. This paper provided a hybrid modeling technique composed of ply-based modeling for a cylinder part and laminate-base modeling technique for a dome part for enhancing modeling efficiency. The ply-based modeling technique provided accurate ply stresses directly for predicting material failure, on the other hand, additional manipulations in stress calculations, which may cause some errors, were needed for the case of the laminate-based modeling technique. The ply stresses in fiber, transverse and in-plane shear directions were compared with the corresponding material strengths to predict material failure.

• International Journal of Ocean System Engineering
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• v.3 no.3
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• pp.136-141
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• 2013

The present research was aimed at comparing performance of metallic and polymer composite shells of a typical underwater vessel of length and inner diameter of 1650 mm and 350 mm respectively, based on the critical buckling pressure for operating depth of 1000 m using ANSYS. High strength steel, aluminium alloy, titanium alloy, glass / epoxy and carbon / epoxy materials were examined. The results indicated weight savings of 46 % in carbon/epoxy and 31 % in glass / epoxy when compared with high strength steel, based on the thickness of the shell for sustaining 10 MPa buckling pressure.