• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.225-236
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• 1994

The sudden discharge of the loop seal water, which is present upstream of the SEBIM POSRV, creates large momentum and inertia forces on the downstream of the discharge piping system. This study provides the procedures and results of analysis of the thermal-hydraulic transient in the SEBIM POSRV discharge piping during the valve opening. The analysis is peformed by RELAP5/MOD3. The appropriate modeling of the discharge piping system, SEBIM POSRV opening characteristics, and loop seal water discharge for the RELAP5/MOD3 analysis is suggested. Also performed is the sensitivity study for the selection of proper options for the junction and volume control. flags. The analysis results demonstrate the adequacy of the RELAP5/HOD3 for the thermal-hydraulic transient analysis of the loop seal water discharge of the SEBIM POSRV discharge piping system. From the sensitivity analysis results, it is shown that the smooth area change option with reasonable geometric pressure drop distribution, non-equilibrium option, and proper time step should be selected for loop seal water discharge analysis.

• 대한방사선방어학회:학술대회논문집
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• 2011.04a
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• pp.108-109
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• 2011

• Korean Chemical Engineering Research
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• v.55 no.2
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• pp.170-179
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• 2017

We investigated design requirements of feed water drain and hydrogen vent systems for the sodium-water reaction pressure relief system (SWRPRS) of the prototype generation IV sodium cooled fast reactor (PGSFR). We evaluated the areas of the gas vent pipe of the water dump tank and the length of the water drain pipe of the steam generator to rapid drain of the water steam inside the steam generator for the normal and refueling operations, respectively. We also calculated the diameter of the gas vent pipe of the sodium dump tank which met its design pressure.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.345.2-345
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• 1999

• Proceedings of the KWS Conference
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• 2004.05a
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• pp.308-310
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• 2004

1980년대 초 미국의 비등형 경수로(Boiling Water Reactor : BWR) 원자력발전소 배관계통의 입계 응력 부식 균열(Inter-Granular Stress Corrosion Crack) 검사 결과 및 미국 EPRI(Electric Power Research Institute)에서 실시한 round robin test 결과에서 기존 초음파 검사 방법의 실효성에 많은 문제점이 제기 되었다. (중략)

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.312-323
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• 1994

1986년 12월 미국의 Surry Unit 2 발전소에서 발생한 급수배관의 대형 파손사고가 침식-부식(erosion-corrosion) 현상에 의해 일어난 것으로 밝혀진 이후, 조사 결과 2차계통에 광범위하게 사용되는 탄소강, 저합금강 재질에서 이와 유사한 손상사례가 많이 나타나는 것으로 밝혀졌다. 이러한 침식-부식 손상은 물-증기로 이루어진 계의 단상(water) 또는 2상(water-wet steam) 조건에서 발생된다. 국내의 원자력 발전소 2차계통에서도 이러한 침식-부식 손상이 나타나고 있으며, 현재 손상원인 해석과 이에 대한 대책 수립이 시급히 요청되고 있다. 본 기고문은 국내 원전의 침식-부식 손상조사와 이의 대책수립을 위한 연구에 활용될 수 있는 침식-부식 손상의 개념, 현상학적 양상, 주요인자의 영향 및 해외 원전의 손상경험 사례 등을 종합하여 정리한 것이다.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.23 no.7 s.166
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• pp.1182-1195
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• 1999

The purpose of this study is to investigate experimentally the effects of both seismic loading and crack length on the fracture behavior of piping system with a circumferential crack in nuclear power plants. The experiments were performed using both large scale piping system facility and 4 points bending test machine under PWR operating conditions. The difference in the load carrying capacities between cracked piping and non-cracked piping was also investigated using the results from experiments and numerical calculations. The results obtained from the experiments and estimation are as follows : (1) The safety margin under seismic loading is larger than those under quasi static loading or simple cyclic loading. (2) There was no significant effect of crack length on tincture behavior of piping system with both a surface crack and a through-wall crack. (3) The load carrying capacity in cracked piping was reduced by factors of 7 to 46 compared to non-cracked piping.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.311-318
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• 1992

During mid-loop operation of Nuclear Power Plant, to prevent the Decay Heat Removal System (DHRS) from failure due to air entrainment of free surface vortex in the piping system, a set of simulating experiments was performed. Through these experiments, a relation between the non-dimensionalized numbers, such as H/d, Froude number, Reynolds number, was found. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from viewpoint of reactor safety, a modified inlet device which is reducer type is strongly recommended for the prevention of air entrainment into DHRS.

• Journal of the Korean Institute of Gas
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• v.15 no.1
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• pp.54-59
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• 2011

Generally, the flow hunting is observed in almost all of the orifice meters but the intensity of the flow hunting is different at each metering system. In order to investigate the relations between pipe diameter and the flow instability or the flow hunting in a real metering system, a one-dimensional pipeline network model was built and analyzed for the examination of flow characteristics and relations to the flow hunting, changing diameters of the meter and the pipes before and after the meter. Finally, the effects of pressuredifference variation and flow hunting following to the variations of the diameters of the meter and the pipes before and after the meter were investigated and the relations were examined as well.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.353-360
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• 1993

Many types of pipe whip restraints are installed to protect the structural components from the anticipated pipe whip phenomena of high energy lines in nuclear power plants. It is necessary to investigate these phenomena accurately in order to design the pipe whip restraints properly and/or to evaluate the acceptability of the pipe whip restraint design. Various research programs have been conducted in many countries to develop analytical methods and to verify the validity of the methods. In this study, various types of finite elements in ANSYS[1], the general purpose finite element computer program, was used to simulate the postulated pipe whips to obtain impact loads and the calculated results were compared with the specific experimental results from the sample pipe whip test for the U-shaped pipe whip restraints. Some calculational models, having the gap element or the spring element between the pipe whip restraint and the pipe line, give reasonably good transient responses of the restraint forces compared with the experimental results, and could be useful in evaluating the acceptability of the pipe whip restraint design.