• 한국소음진동공학회논문집
• /
• 제15권8호
• /
• pp.988-995
• /
• 2005

Vibration has been severly increased at the branch pipe of main steam header since the commercial operation of nuclear power plant. Intense broad band disturbance flow at the discontinuous region such as elbow, valve, and header generates the acoustical pulsation which is propagated through the piping system. The pulsation becomes the source of low frequency vibration at piping system. If it coincide with natural frequency of the pipe system, excessive vibration is made. High level vibration due to the pressure pulsation related to high dynamic stress, and ultimately, to failure probability affects fatally the reliability and confidence of plant piping system. This paper discusses vibration effect for the branch pipe system due to acoustical pulsations by broad band disturbance flow at the large main steam header in 700 MW nuclear power plant. The exciting sources and response of the piping system are investigated by using on-site measurements and analytical approaches. It is identified that excessive vibration is caused by acoustical pulsations of 1.3 Hz, 4.4 Hz and 6.6 Hz transmitted from main steam balance header, which are coincided with fundamental natural frequencies of the piping structure. The energy absorbing restraints with additional stiffness and damping factor were installed to reduce excessive vibration.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2005년도 춘계학술대회논문집
• /
• pp.780-785
• /
• 2005

Vibration has been severly increased at the branch pipe of main steam header since the commercial operation of a nuclear power plant. Intense broad band disturbance flow at the discontinuous region such as elbow, valve or heather generates the acoustical pulsation which is propagated through the piping system. The pulsation becomes the source of low frequency vibration at piping system. If it coincide with natural frequency of the pipe system, excessive vibration is made. High level vibration due to the pressure pulsation related to high dynamic stress, and ultimately, to failure probability affects fatally the reliability and confidence of plant piping system. This paper discusses vibration effect for the branch pipe system due to acoustical pulsations by broad band disturbance flow at the large main steam header in 7nn nuclear power plant. The exciting sources and response or the piping system are investigated by using on site measurements and analytical approaches. It is identified that excessive vibration is caused by acoustical pulsations of 1.3Hz, 4.4Hz and 6.6Hz transferred from main steam header, which are coincided with fundamental natural frequencies of the piping structure. The energy absorbing restraints with additional stiffness were installed to reduce excessive vibration.

• 한국가시화정보학회지
• /
• 제14권2호
• /
• pp.18-24
• /
• 2016

This paper presents an experimental investigation to visualize cross-sectional two-phase flow structure and identify liquid-gas interface for condensation of steam at a low mass flux in a slightly inclined tube using the axial-viewing technique, which permits to look directly into flow during condensation of steam. In this technique, two-phase flow is viewed along the axis of a pipe by locating a high-speed video camera in front of a viewer that is fitted at the outlet of the pipe. A short section of the pipe is illuminated and is recorded through the viewer, which is kept free of liquid by mildly introducing air. Experiments were conducted in a pipe of 19.05 mm in inner diameter at atmospheric pressure. Cross-sectional two-phase flow structure is obtained at a steam mass flux of $2.62kg/m^2s$ as a function of steam quality in the range from 0.5 to 0.9. The results show that stratified-wavy flow is a unique flow pattern observed in the scope of the present study. Condensate film thickness, stratification angle and void fraction were measured from the obtained flow structure images. Finally, heat transfer coefficient was calculated using the measurement data and discussed in comparison with existing correlations.

• 한국압력기기공학회 논문집
• /
• 제11권1호
• /
• pp.74-78
• /
• 2015

Nuclear power plant steam generator that is one of the main component has several thousands of thin tubes. And the steam generator tube is subject to damage because of the severe operation conditions such as the high temperature and pressure. Therefore periodic inspections are conducted to ensure the integrity of steam generator component. Hanul unit 3 also has been inspected in accordance with in-service inspection program and is scheduled to be replaced for exceeding the plugging rate which was recommended by manufacturer. During the steam generator replacement activity, we found several clustered porosity on inner surface of main feed water pipe. Additionally crack-like indications were found at weld interface between base material and weld of main feed water pipe. This paper describes the field experience and visual testing results for inner surface of main feed water pipes. The destructive test result had shown that these indications were porosities which were caused by manufacturing process not by operation service.

• 한국압력기기공학회 논문집
• /
• 제12권2호
• /
• pp.58-65
• /
• 2016

Since, in case of high energy piping, steam jets ejected from the rupture zone may cause damage to nearby structure, it is necessary to design it into consideration of nuclear power plant design. For the existing nuclear power plants, the ANSI / ANS 58.2 technical standard for high-energy pipe rupture was used. However, the US Nuclear Regulatory Commission (USNRC) and academia recently have pointed out the non-conservativeness of existing high energy pipe fracture evaluation methods. Therefore, it is necessary to develop a highly reliable evaluation methodology to evaluate the behavior of steam jet ejected during high energy pipe rupture and the effect of steam jet on peripheral devices and structures. In this study, we develop a method for analyzing the impact load of a jet by high energy pipe rupture, and plan to carry out an experiment to verify the evaluation methodology. In this paper, the basic data required for the design of the jet impact load experiment equipment under construction, 1) the load change according to the jet distance, 2) the load change according to the jet collision angle, 3) the load variation according to structure diameter, and 4) the load variation depending on the jet impact position, are numerically obtained using the developed steam jet analysis technique.

• 한국전산유체공학회지
• /
• 제20권3호
• /
• pp.87-93
• /
• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2003년도 추계학술대회
• /
• pp.1198-1205
• /
• 2003

Most domestic fossil power plant have exceeded 100,000 hours of operation with the severe operating condition. Among the critical components of fossil power plant, high temperature steam pipe system have had a many problems and damage from unstable displacement behavior because of frequent start up and shut down. In order to prevent the serious damage and failure of the critical pipe system in fossil power plant, 3-dimensional displacement measurement system were developed for the on-line monitoring system. 3-D Measurement system was developed with using the LVDT type sensor and rotary encoder type sensor, this system was installed and operated on the real power plant successfully. In the future time, network system of on-line diagnosis for critical pipe will be designed.

• Nuclear Engineering and Technology
• /
• 제32권2호
• /
• pp.142-156
• /
• 2000

An experimental study of steam condensation on a subcooled thick water layer (0.018 ~0.032 m) in a countercurrent stratified flow has been performed using a nearly horizontal circular pipe. A total of 103 average interfacial condensation heat transfer coefficients were obtained and parametric effects of steam and water flow rates and the degree of subcooling on condensation heat transfer were examined. The measured local temperature and velocity distributions in the thick water layer revealed that there was a thermal stratification due to the lack of full turbulent thermal mixing in the lower region of the water layer Two empirical Nusselt number correlations, one in terms of average steam and water Reynolds numbers, and the water Prandtl number, and the other in terms of the Jakob number in place of the Prandtl number, which agree with most of the data within $\pm$ 25%, were developed based on the bulk flow properties. Comparisons of the present data with existing correlations showed that the present data were significantly lower than the values predicted by existing correlations.

• International Journal of Safety
• /
• 제3권1호
• /
• pp.53-58
• /
• 2004

SMART-P is a promising advanced small and medium category nuclear power reactor. It is an integral type reactor with a sensible mixture of new innovative design features and proven technologies aimed at achieving a highly enhanced safety and improved economics. The enhancement of the safety and reliability is realized by incorporating inherent safety improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, and component modularization. Preliminary safety analyses on selected limiting accidents confirm that the inherent safety improving design characteristics and the safety system of SMART-P ensure the reactor's safety. SMART-P is an advanced integral pressurized water reactor. The purpose of this study is for the safety analysis of the steam generator module pipe break for the SMART-P. The integrity of the fuel rod is the major criteria of this analysis. As a result of this analysis, the safety of the RCS and the secondary system is guaranteed against the module pipe break of a steam generator of the SMART-P.

• 한국안전학회지
• /
• 제20권3호
• /
• pp.34-41
• /
• 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.