• Nuclear Engineering and Technology
• /
• 제33권4호
• /
• pp.397-408
• /
• 2001

A marine reactor shows very different thermal-hydraulic characteristics compared to a land- based reactor. Especially, study on the variation of flow field due to ship motions such as inclination, heaving and rolling is essential since the flow variation has great influence on the reactor cooling capability. In this study, the natural circulation characteristics of integral type marine reactor with modular steam generators were analyzed using computational fluid dynamics code, CFX-4, for inclined conditions. The numerical analyses are performed using the results of natural circulation experiments for integral reactor which are already conducted at Seoul National University. From the results, it was found that the flow rate in the ascending steam generator cassettes increases due to buoyancy effect. Due to this flow variation, temperature difference occurs at the outlets of the each steam generator cassettes. which is mitigated through downcomer by thermal mixing. Also, around the upper pressure header the flow from descending hot leg goes up to the ascending steam generator cassettes due to large natural circulation driving force in ascending steam generator cassettes. From this result, the increase of How rate in the ascending steam generator cassettes could be understood qualitatively.

• 한국전산유체공학회:학술대회논문집
• /
• 한국전산유체공학회 2004년도 춘계 학술대회논문집
• /
• pp.49-55
• /
• 2004

This paper addresses a numerical simulation of the flow and heat transfer in a simplified model of helically coiled tube steam generator using a general purpose computational fluid dynamic analysis computer code. The steam generator model is comprised of a cylindrical shell and helically coiled tubes. A cold feed water entered the tubes is heated up, evaporates. and finally become a superheated steam with a large amount of heat transferred continuously from the hot compressed water at higher pressure flowing counter-currently through the shell side. For the calculation of tube side two-phase flow field formed by boiling, inhomogeneous two-fluid model is used. Both the internal and external turbulent flows are simulated using the standard k-e model. The conjugate heat transfer analysis method is employed to calculate the conduction in the tube wall with finite thickness and the convections in the internal and external fluids simultaneously so as to match the fluid-wall-fluid interface conditions properly. The numerical calculations are peformed for helically coiled tubes of steam generator at an integral type pressurized water reactor under normal operation. The effects of tube-side inlet flow velocity are discussed in details. The results of present numerical simulation are considered to be physically plausible based on the data and knowledge from previous experimental and numerical studies where available.

• 한국전산유체공학회:학술대회논문집
• /
• 한국전산유체공학회 2004년도 추계 학술대회논문집
• /
• pp.127-130
• /
• 2004

In this study, three-dimensional numerical calculations are peformed to simulate the flow and heat transfer in helically coiled tube steam generator employing a commercial CFD (Computational Fluid Dynamics) code. The problem considered herein includes the boiling phase change flow of tube side fluid and the single-phase counter-current flow of shell side hot fluid transferring heat to the tube side flow thru the tube wall. Detailed investigations are performed for both shell-side and tube-side flow fields in terms of density and volume fractions of each phase of fluids as well as for the tube wall heat transfer field in terms of heat transfer coefficients.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2007년도 제38회 하계학술대회
• /
• pp.1605-1606
• /
• 2007

An electrical generator in power plant is driven and maintained its speed at rated by steam turbine. By the way, after synchronization in parallel with the power system, as the steam flow into turbine can not be reduced fast even though the electrical load is lost, the turbine gets into dangerous situation due to the increase of its speed. At this time, the duty of the turbine governor is to limit the speed to its overspeed trip set point by stopping the steam flow as soon as possible, the test of which is called load rejection test. It is introduced in this paper for a field simulation test of generator load rejection to be implemented on the turbine governor in a 680MW nuclear power plant before its startup.

• Nuclear Engineering and Technology
• /
• 제35권5호
• /
• pp.442-453
• /
• 2003

This study investigates the safety assessment of the potential for fretting-wear damages on steam generator (SG) U-tubes caused by foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions are obtained from three-dimensional SG flow calculation using the ATHOS3 code. Modal analyses are performed for the finite element modelings of U-tubes to get the natural frequency, corresponding mode shape and participation factor. The wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted. Also, discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube.

• Nuclear Engineering and Technology
• /
• 제53권3호
• /
• pp.1029-1040
• /
• 2021

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

• 한국소음진동공학회논문집
• /
• 제13권5호
• /
• pp.400-408
• /
• 2003

This paper investigates the vibration characteristics of steam generator (SG) U-tubes with defect. The operating SG shell-side flow field conditions for determining the fluidelastic instability parameters such as added mass are obtained from three-dimensional SG flow calculation. Modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, addressed is the effect of the internal pressure on the vibration characteristics of the tube.

• 대한기계학회논문집
• /
• 제17권1호
• /
• pp.200-214
• /
• 1993

본 연구에서는 Westinghouse Model 51 증기발생기의 U-bend 영역에서 2차측 유체의 횡단유동으로 유발될 수 있는 튜브군의 유체탄성불안정을 예측하기 위한 해석 을 수행하고 그 대표적인 결과들을 제시하였다. 그리고 U-bend 영역에서 AVB에 의한 튜브의 지지상태와 형태 및 최상부 TSP에서 Denting 또는 이물질 고착으로 인하여 변 경된 튜브의 고정지지조건 등이 튜브의 유체탄성불안정 응답에 미치는 영향을 조사하 였다. 유체탄성불안정 해석과정에서 필수적으로 선행되어야 하는 2차측 3차원 2상 유동장 계산은 증기발생기 열수력 해석용인 ATHOS3 코드로써 수행되었으며, U-튜브의 고유진동수와 모우드 형상은 공학해석용 유한요소 프로그램인 ANSYS코드로써 계산되었 다.

• Nuclear Engineering and Technology
• /
• 제51권4호
• /
• pp.1008-1016
• /
• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• Nuclear Engineering and Technology
• /
• 제53권9호
• /
• pp.3122-3125
• /
• 2021

Electromagnetic vortex flowmeter is a new type of instrument for detecting leakage of steam generator, and the signal processing method based on the envelope to calculate energy ratio can effectively detect bubbles in sodium flow. The signal processing method is not affected by changes in the amplitude of the sensor output signal, which is caused by changes in magnetic field strength and other factors. However, the detection sensitivity of the electromagnetic vortex flowmeter is reduced. To this end, a signal processing method based on inverse Fourier transform to calculate energy ratio is proposed. According to the difference between the frequency band of the bubble noise signal and the flow signal, only the amplitude in the frequency band of the flow signal is retained in the frequency domain, and then the flow signal is obtained by the inverse Fourier transform method, thereby calculating the energy ratio. Using this method to process the experimental data, the results show that it can detect 0.1 g/s leak rate of water in the steam generator, and its performance is significantly better than that of the signal processing method based on the envelope to calculate energy ratio.