• 대한기계학회논문집B
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• 제41권10호
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• pp.649-657
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• 2017

고압의 파이프 파단 시 파이프 내에 있던 유체가 고속으로 대기로 분출될 때 압축성유동을 동반하는 초음속제트가 발생한다. 이러한 초음속제트는 일반적으로 복잡한 비정상거동을 보여줄 수 있다. 본 연구는 이러한 고압파이프에서 분출되는 초음속제트에 의해 생성되는 압축성유동을 고찰하기 위하여 전산유체역학 해석이 수행되었다. 분출기체의 종류 및 파이프직경 변화에 따른 비정상유동 특성을 해석하기 위해 SST $k-{\omega}$ 난류모델이 채택되었다. 전산해석 시 기본 경계조건은 파이프직경 10 cm, 제트 압력비 5, 기체온도 300 K로 가정하였다. 그 해석결과로 초음속제트로 인해 생성되는 충격파의 거동이 관찰되었고, 간접적인 영향으로 폭풍파도 발생됨을 알 수 있었다. 기체의 분자량이 가장 작은 $H_2$의 압력파 특성은 안전영역까지의 거리가 가장 짧았으며, 분자량이 비슷한 $N_2$, 공기 및 $O_2$는 큰 차이가 없었다. 또한 파이프직경이 커져 제트에 의한 영향범위도 더욱 증대됨을 알 수 있었다.

• 설비공학논문집
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• 제24권10호
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• pp.745-749
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• 2012

Heating pipes burst was occurred in the apartment complex that was applied floor radiant heating system. There were two opinions for the cause of the bursted heating pipes that was the flaw during construction and defects in the product and also there were conflicting among them. Officials analyzed it in order to investigate the cause of the rupture. Tensile test results showed different tensile strength between the lower part of heating pipe and the upper part of heating pipes. The lower tensile strength is maintained while the top was not secured. The reason why rupture heating pipes is that flow velocity isn't secured and then the air get stagnant. Stagnant air makes hardening. It is caused rupturing. The proper flow rate was confirmed 0.166 m/sec after experiment. It isn't make stagnant air inside heating pipes.

• Journal of Advanced Marine Engineering and Technology
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• 제24권3호
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• pp.79-87
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• 2000

This paper describes experimental investigations of helium-air exchange flow through openings with vertical partitions. Such exchange flows may occur following rupture accident of stand pipe in high temperature gas cooled reactor. Exchange flow rates are investigated experimentally by using partitioned opening and opening with extended partition to assess fluids interference of the exchange flow at the stand pipe rupture accident. A tests vessel with the two types of opening on top of test cylinder is used in the experiments. An estimation method of mass increment is developed and applied to measure the exchange flow rate. A technique of flow visualization by Mach-Zehnder interferometer is provided to recognize the exchange flows. Amplitude and progress of interference fringes of the flows are observed and used as a support in comparison with the exchange flow rates. Flow passages of upward flow of the helium and downward flow of the air for both two types of the opening are separated by inserted partition within the opening, but in the case of partitioned opening, unseparated flow is formed at the opening entrance and the two flows interface. The exchange flow rate for the partitioned opening is not greater than that of the opening with extended partition because of the fluids interference at the entrance of opening. Finally, the fluids interference at the opening entrance is found to be one of important factors on the helium-air exchange flow rate.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1939-1950
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• 2019

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

• 한국안전학회지
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• 제14권2호
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• pp.32-41
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• 1999

If the Leak-Before-Break (LBB) concept is applicable to the nuclear piping design, it is not necessary to consider the dynamic effect due to pipe rupture. Therefore, the construction cost can be significantly reduced by eliminating unnecessary pipe whip restraints and jet impingement devices. The objective of this paper is to develop the Piping Evaluation Diagram (PED) for efficient application of LBB concept to piping system at an initial piping design stage. For this purpose, the 3-D finite element analyses were performed to evaluate the crack stability. And the stress-strain curve based on the pipe material tests were used to calculate the detectable leakage crack length. Finally, the present PED which was composed as a function of NOP load and allowable SSE load, was developed for an application of LBB concept to the safety injection and shutdown cooling line in Korean Next Generation Reactor (KNGR).

• Corrosion Science and Technology
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• 제12권4호
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• pp.179-184
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• 2013

Pipe wall thinning caused by erosion and corrosion can adversely affect the operation of aged nuclear power plants. Some injured workers owing to pipe rupture has been reported and power reduction caused by unexpected pipe damage has been occurred consistently. Therefore, it is important to develop erosion-corrosion damage prediction model and investigate its mechanisms. Especially, liquid droplet impingement erosion(LDIE) is regarded as the main issue of pipe wall thinning management. To investigate LDIE mechanism with corrosion environment, we developed erosion-corrosion damage simulation apparatus and its capability has been verified through the preliminary damage experiment of 6061-Al alloy. The apparatus design has been based on ASTM standard test method, G73-10, that use high-speed rotator and enable to simulate water hammering and droplet impingement. The preliminary test results showed mass loss of 3.2% in conditions of peripheral speed of 110m/s, droplet size of 1mm-diameter, and accumulated time of 3 hours. In this study, the apparatus design revealed feasibility of LDIE damage simulation and provided possibility of accelerated erosion-corrosion damage test by controlling water chemistry.

• 한국가스학회지
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• 제24권4호
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• pp.32-38
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• 2020

화학반응기에 발생한 화재·폭발 사고사례를 기반으로 화학반응기에 설치되어 있는 예방조치의 문제점을 분석하였다. 화학반응기는 다품종의 화학제품을 생산하며, 반응폭주시 급격히 상승하는 압력을 해소하기 위해 파열판을 설치하고 파열판의 기능을 유지하기 위해 배출물질을 대기로 배출하도록 허용하고 있어 화재·폭발사고가 발생하였다. 이를 개선하기 위한 방안으로 안전건전성수준(SIL3)을 기반으로 한 안전계장시스템(SIS)을 화학반응기의 예방조치로 적용하였다. 화학반응기의 원재료를 적하하는 배관에 긴급차단밸브를 직렬로 2개 설치하여 반응폭주시 긴급차단밸브 2개 중 1개만 작동하여도 원재료 공급을 차단할 수 있도록 하고, 반응응제제 공급배관에는 자동 ON/OFF 밸브를 병렬로 설치하여 반응폭주시 1개의 밸브만 열려도 반응억제제가 투입될 수 있게 하였다.

• Corrosion Science and Technology
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• 제15권4호
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• pp.182-188
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• 2016

Significant piping wall thinning caused by Flow-Accelerated Corrosion (FAC) and Erosion-Corrosion (EC) continues to occur, even after the Mihama Power Station unit 3 secondary pipe rupture in 2004, in which workers were seriously injured or died. Nuclear power plants in many countries have experienced FAC and EC-related cases in steam cycle piping systems. Korea has also experienced piping wall thinning cases including thinning in the downstream straight pipe of a check valve in a feedwater pump line, the downstream elbow of a control valve in a feedwater flow control line, and failure of the straight pipe downstream of an orifice in an auxiliary steam return line. Cause analyses were performed by reviewing thickness data using Ultrasonic Techniques (UT) and, Scanning Electron Microscope (SEM) images for the failed pipe, and numerical simulation results for FAC and EC cases in Korea Nuclear Power Plants. It was concluded that the main cause of wall thinning for the downstream pipe of a check valve is FAC caused by water vortex flow due to the internal flow shape of a check valve, the main cause of wall thinning for the downstream elbow of a control valve is FAC caused by a thickness difference with the upstream pipe, and the main cause of wall thinning for the downstream pipe of an orifice is FAC and EC caused by liquid droplets and vortex flow. In order to investigate more cases, additional analyses were performed with the review of a lot of thickness data for inspected pipes. The results showed that pipe wall thinning was also affected by the operating condition of upstream equipment. Management of FAC and EC based on these cases will focus on the downstream piping of abnormal or unusual operated equipment.

• 한국안전학회지
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• 제20권3호
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• pp.34-41
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• 2005

For assessing residual lift of the steam pipe in fossil power plants, inspections and analysis are usually focused on the critical locations such as butt welds, elbows, Y-piece and T-piece of the steam pipes. In predicting the residual life of T-piece, determination of local stress near welds considering system load as well as internal pressure is not a simple problem. In this study, stress analysis of a T-piece pipe was conducted using a three-dimensional model which represents the T-piece of a domestic fossil power station. Elastic and elastic-creep analysis showed the maximum stress level and its location. Residual creep rupture life was also calculated using the stress analysis results. It was argued that the calculated life is reasonably same as the measured one. The stress analysis results also support life prediction methodology based on in-field replication technique.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.1-8
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• 2024

In this study, the influence of thermal aging on structural integrity is investigated for Gr. 91 steel. A commercial grade Gr. 91 steel is used for the virgin material, and service-exposed Gr. 91 steel is sampled from a steam pipe of a super critical plant. Time versus creep strain curves are obtained through creep tests with various stress levels at 600 ℃ for the virgin and service-exposed Gr. 91 steels, respectively. Based on the creep test results, the improved Omega model is characterized for describing the total creep strain curve for both Gr. 91 steels. The proposed parameters for creep deformation model are used for predicting the steady-state creep strain rate, creep rupture curve, and stress relaxation. Creep-fatigue damage is evaluated for the intermediate heat exchanger (IHX) in a large-scale sodium test facility of STELLA-2 by using creep deformation model with proposed creep parameters and creep rupture curve for both Gr. 91 steels. Based on the comparison results of creep fatigue damage for the virgin and service-exposed Gr. 91 steels, the thermal aging effect has been shown to be significant.