ng Chull Jo

Numerical Simulation of Subcooled Flow Boiling Heat Transfer in Helical Tubes

This paper addresses the numerical simulation of two-phase flow heat transfer in the helically coiled tubes of an integral type pressurized water reactor steam generator under normal operation using a computational fluid dynamics code. The shell-side flow field where a single-phase fluid flows in the downward direction is also calculated in conjunction with the tube-side two-phase flow characteristics. For the calculation of tube-side two-phase flow, the inhomogeneous two-fluid model is used. Both the Rensselaer Polytechnic Institute wall boiling model and the bulk boiling model are implemented for the numerical simulations of boiling-induced two-phase flow in a vertical straight pipe and channel, and the computed results are compared with the available measured data. 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. Both the internal and external turbulent flows are simulated using the standard k-e model. From the results of the present numerical simulation, it is shown that the bulk boiling model can be applied to the simulation of two-phase flow in the helically coiled steam generator tubes. In addition, the present simulation method is considered to be physically plausible in the light of discussions on the computed results.

Fluidelastic instability analysis of operating nuclear steam generator U-tubes

This paper presents a systematic assessment methodology of the potential for steam generator tube failures caused by fluidelastic instability in operating nuclear power reactors and provides the results of assessment for the U-tube steam generator (UTSG) model being employed at Kori units 2, 3 and 4, and Yonggwang units 1 and 2 in Korea. The assessment process involves evaluation of anti-vibration bar insertion conditions for the UTSG, three-dimensional thermal-hydraulic analysis of the steam generator, determination of flow distributions along the length of a specific U-tube, calculation of natural frequencies and mode shapes of the tube, and fluidelastic tube instability analysis. The thermal-hydraulic analysis for providing the detailed three-dimensional two-phase flow field in the secondary side of the steam generator model was accomplished using the ATHOS3 steam generator thermal-hydraulic analysis code. The UTVA code designed for calculating both the free vibration responses and fluidelastic stability ratio of a specific U-tube was used to assess the potential for fluidelastic instability of the steam generator U-tubes at various conditions of anti-vibration bar (AVB) inactive modes. In addition, the effects of tube plugging on the forced response of either plugged or intact tubes were discussed.

Multidimensional phase change problems by the dual-reciprocity boundary-element method

This article addresses the numerical analysis of nonlinear transient heat transfer with melting or solidification. An effective and simple procedure is presented for the simulation of the motion of the boundary and the transient temperature field during the phase change process. To accomplish this purpose, an iterative implicit solution algorithm has been developed by employing the dual-reciprocity boundary-element method. The dual-reciprocity boundary-element approach provided in this article is much simpler than the usual boundary-element method in applying a reciprocity principle and an available technique for dealing with the domain integral of the boundary element formulation simultaneously. In this article, attention is focused on two-dimensional melting (ablation) solidification problems for simplicity. The accuracy and effectiveness of the present analysis method have been illustrated through comparisons of the calculation results of some examples of one-phase ablation solidification problems with...

Modal Analysis of Conical Shell Filled with Fluid

As a basic study on the fluid-structure interaction of the shell structure, a theoretical formulation has been suggested on the free vibration of a thin-walled conical frustum shell filled with an ideal fluid, where the shell is assumed to be fixed at both ends. The motion of fluid coupled with the shell is determined by means of the velocity potential flow theory. In order to calculate the normalized natural frequencies that represent the fluid effect on a fluid-coupled system, finite element analyses for a fluid-filled conical frustum shell are carried out. Also, the effect of apex angle on the frequencies is investigated.

Numerical Analysis of Unsteady Flow Field in the RWT for the Prediction of the Potential for Air Ingression Into the ECC Supply Lines During the SBLOCA at the KSNPs

This paper addresses the three-dimensional analysis of unsteady flow in the RWT (Refueling Water Tank) for the prediction of the potential for air ingression into the ECC (Emergency Core Cooling) pump during the SBLOCA (Small Break Loss Of Coolant Accident) at KSNPs (Korean Standard Nuclear Power plants). Upon the receipt of RAS (Recirculation Actuation Signal) by the occurrence of SBLOCA, the RWT outlet valve is designed to be isolated manually. At the nuclear power plants without the provision of automatic isolation operation of the valve on the downstream of the RWT line, the refueling water begins to discharge from the RWT, which may result in forming and developing the vortex flow in the RWT, under the condition of the minimum pressure of containment and minimum water level of containment recirculation sump during the phase of RAS. Due to the vortex flow, when the water level is below the critical height, a dip starts to develop, causing air ingression before the refueling water drains fully. Hence it can be surmised that there is a possibility of ECC pump failure due to air ingression into the ECC supply line even before the RWT is fully drained. Therefore, in this work, when the RAS is actuated followed by the SBLOCA occurrence, a quantitative evaluation for the maximum limiting allowable time for the manual closing of RWT outlet valve is carried out to eliminate the possibility of air ingression into the ECC pump from the RWT. To do this, the unsteady flow field in the RWT including the drain pit with the connected discharge piping in the process of SBLOCA is analyzed using a CFD (Computational Fluid Dynamics) code. In addition, the transient flow behavior accompanying air entrainment resulting from the dip formation due to vortex flow at the upper part of RWT is examined and the applicable limiting time of the isolation valve closing for preventing air ingression is assessed.Copyright

Free Vibration Analysis of Perforated Plate Submerged in Fluid

An analytical method to estimate the coupled frequencies of the circular plate submerged in fluid is developed using the finite Fourier-Bessel series expansion and Rayleigh-Ritz method. To verify the validity of the analytical method developed, finite element method is used and the frequency comparisons between them are found to be in good agreement. For the perforated plate submerged in fluid, it is almost impossible to develop a finite element model due to the necessity of the fine meshing of the plate and the fluid at the same time. This necessitates the use of solid plate with equivalent material properties. Unfortunately the effective elastic constants suggested by the ASME code are found to be not valid for the modal analysis. Therefore in this study the equivalent material properties of perforated plate are suggested by performing several finite element analyses with respect to the ligament efficiencies.

Heat Transfer of a Spray Droplet in a PWR Pressurizer

Heat transfer rates to spray droplets under conditions corresponding to those of spray transients in a pressurizer of pressurized water reactor (PWR) have been predicted by a simple droplet model with internal thermal resistance and partial internal mixing. In those processes, the temperature distributions in the droplet have been obtained using the integral method, and the physical properties of the saturated steam-hydrogen gas mixture surrounding the droplets are estimated applying the concept of compressibility factor and using appropriate correlations. Results have been provided for the temporal variations of total heat flux with its convection and condensation heat transfer components, dimensionless droplet bulk temperature and droplet flight distance. The effects of ambient pressure, initial droplet size, concentration of hydrogen gas in the mixture, initial injection velocity, and spray angle on the heat transfer of spray droplets have been discussed.

Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurring at PWR Plants

At conventional pressurized water reactors (PWRs), cold water stored in the refueling water tank of emergency core cooling system is injected into the primary coolant system through a safety injection (SI) line, which is connected to each cold leg pipe between the main coolant pump and the reactor vessel during the SI operation, which begins on the receipt of a loss of coolant accident signal. In normal reactor power operation mode, the wall of SI line nozzle maintains at high temperature because it is the junction part connected to the cold leg pipe through which the hot main coolant flows. To prevent and relieve excessive transient thermal stress in the nozzle wall, which may be caused by the direct contact of cold water in the SI operation mode, a thermal sleeve in the shape of thin wall cylinder is set in the nozzle part of each SI line. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the junction of primary coolant main pipe-SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in detail by using both computational fluid dynamics code and structure analysis finite element code. As a result, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15 Hz to 18 Hz. These frequencies coincide with the lower mode natural frequencies of thermal sleeve, which has a pinned support condition on the outer surface with the circumferential prominence set into the circumferential groove on the inner surface of SI nozzle at the midheight of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yields alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.

Root Cause Analysis of SI Nozzle Thermal Sleeve Breakaway Failures Occurred at PWR Plants

Thermal sleeves in the shape of thin wall cylinder seated inside the nozzle part of each safety injection (SI) line at pressurized water reactors (PWRs) have such functions as prevention and relief of potential excessive transient thermal stress in the wall of SI line nozzle part which is initially heated up with hot water flowing in the primary coolant piping system when cold water is injected into the system through the SI nozzles during the SI operation. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. To find out the root cause of thermal sleeve breakaway failures, the flow situation in the in the junction of primary coolant main pipe and SI branch pipe and the vibration modal characteristics of the thermal sleeve are investigated in details by using both computational fluid dynamic (CFD) code and structure analysis finite element code. As the results, the transient response in fluid pressure exerting on the local part of thermal sleeve wall surface to the primary coolant flow through the pipe junction area during the normal reactor operation mode shows oscillatory characteristics with the frequencies ranging from 15 to 18, which coincide with the lower mode natural frequencies of thermal sleeve having a pinned support condition on the circumferential prominence on the outer surface of thermal sleeve which is put into the circumferential groove on the inner surface of SI nozzle at the mid-height of thermal sleeve. In addition, the variation of pressure on the thermal sleeve surface yield alternating forces and torques in the directions of two rectangular axes perpendicular to the longitudinal axis of cylindrical thermal sleeve, which causes both rolling and pitching motions of the thermal sleeve. Consequently, it is seen that this flow situation surrounding the thermal sleeve during the normal reactor operation can induce resonant vibrations accompanying the shaking motion of the thermal sleeve at the pinned support condition, which finally leads to the failures of thermal sleeve breakaway from the SI nozzle.© 2006 ASME

Flow and Modal Analysis for the Investigation of PWR Safety Injection Line-Installed Thermal Sleeve Separation Failure Mechanism

In conventional pressurized water reactors, a thermal sleeve (named simply ‘sleeve’ hereafter) is seated inside the nozzle part of each safety injection (SI) branch pipe to prevent and relieve potential excessive transient thermal stress in the nozzle wall when cold water is injected during the safety injection mode. Recently, mechanical failures that the sleeves are separated from the SI branch pipe and fall into the connected cold leg main pipe occurred in sequence at some typical PWR plants in Korea. This paper investigates the flow field in the pipe junction through a numerical simulation and vibration characteristics of thermal sleeves through a modal analysis to analyze the root cause of sleeve separation mechanism. By performing both flow simulation in the SI pipe junction and modal analysis of thermal sleeve, the fluid force and modal characteristics have been identified to be able to lead or contribute to separate thermal sleeves inside safety injection branch pipes in PWR plants from their original seating locations.Copyright