Crossflow-induced vibrations of heat exchanger tube banks

This paper presents a mathematical model for crossflow-induced vibration of tube banks. Motion-dependent fluid forces and various types of flow noises are incorporated in the model. An analytical solution for the fluid inertia force, hydrodynamic damping force, and fluid elastic force is given for tube banks arranged in an arbitrary pattern. Based on the model, a better understanding of the vibrations of heat exchanger tube banks subjected to various flow excitations can be developed.     

Modelling of fluidelastic vibrations of heat exchanger tubes with loose supports

The existence of gaps at tube supports necessitates time domain modelling of fluid forces to predict flow-induced vibrations and associated wear in heat exchangers and steam generators. This paper presents a new time-domain model for fluidelastic instability forces of tubes with loose-supports. In this model, the fluidelastic force, which is dependent on flow velocity and array geometry, is superimposed on the turbulence forcing function. The model was used to calculate the critical flow velocity, tube response, and tube/support interaction parameters, such as impact force and work rate. The critical velocity for linear cases was accurately predicted. The critical flow velocity for the loose support case was found to be sensitive to both the gap size and the turbulence level.     

Local heat transfer and hot-spot factors in wire-wrap tube bundle

Heat transfer coefficients and hot-spot factors have been determined from measured local temperatures and calculated local mass flux in seven adjacent tubes and associated subchannels of a 61 wire-wrap tube bundle characteristic of the blanket of a GCFR (Gas Cooled Fast Reactor). The bundle consisted of 2.11 cm OD stainless steel tubes on a triangular array with a pitch/diameter ratio of P/D = 1.05. The helical wire of 0.1067 cm in diameter was coiled on the tube with a respective initial orientation of 0–120–240°C and 30.48 cm helical pitch. The experiment used water at atmospheric pressure and temperature as coolant. The resulting dimensionless correlation for heat transfer is applicable to gases and all non-metal fluids in one phase flow when the fluid properties at subchannel bulk temperature are used. This correlation is based on local subchannel mass flux and is applicable to all wire-wrap configurations. Local subchannel mass fluxes were determined with a computer program COBRA IV and used to correlate the average Nusselt number for each subchannel in terms of local Reynolds number and fluid Prandtl number. The differences of up to 19% between that correlation and the one presented in earlier work are discussed in the text. The hot-spot factors on the convective heat transfer coefficient for tubes and subchannels are given as a function of Reynolds number based on a bundle average mass flux and a local subchannel hydraulic diameter. These factors are specific to the bundle configuration and are also dependent on the wire-wrap configuration.     

Modeling and analysis of thermal damping in heat exchanger tube bundles

Most structures and equipment used in nuclear power plant and process plant, such as reactor internals, fuel rods, steam generator tubes bundles, and process heat exchanger tube bundles, are subjected to flow-induced vibrations (FIV). Costly plant shutdowns have been the source of motivation for continuing studies on cross-flow-induced vibration in these structures. Damping has been the target of various research attempts related to FIV in tube bundles. A recent research attempt has shown the usefulness of a phenomenon termed as ‘thermal damping’. The current paper focuses on the modeling and analysis of thermal damping in tube bundles subjected to cross-flow. It is expected that the present attempt will help in establishing improved design guidelines with respect to damping in tube bundles.     

Investigation of pre- and post-dryout heat transfer of steam-water two-phase flow in a rod bundle

Pre- and post-dryout heat transfer experiments were performed for steam-water two-phase flow in a 5 × 5 rod bundle under conditions of total mass fluxes from 80 to 320 kg/m²s, inlet qualities from 0.1 to 0.8, heat fluxes from 3 to 26 W/cm² and a pressure of 3 MPa. Heater rod surface temperatures or heat transfer coefficients predicted by several correlations were compared with experimental data with emphasis on the applicability of the correlations to the present experimental conditions which were pertinent to thermal-hydraulic conditions during a LOCA in a nuclear reactor. The Chen and Biorge et al. correlations underestimated heat transfer coefficients in the pre-dryout region. The Varone-Rohsenow prediction which accounted for the thermal nonequilibrium effect, calculated heater rod surface temperatures relatively well in the post-dryout region over the whole region of the present experimental conditions. The Dittus-Boelter and Groeneveld correlations predicted heater rod surface temperatures relatively well in the post-dryout region under high total mass flux conditions, but underestimated considerably under low total mass flux conditions.     

Analytical and experimental studies of tube/support interaction in multi-span heat exchanger tubes

Fretting damage of tubes in heat exchangers can be very costly and should be avoided. Information on vibratory responses and dynamic interaction between tubes and supports are prerequisites for understanding the relationship between frettingwear and tube vibration.A finite-element computation technique has been developed to predict the vibratory response and tube/support interaction of multi-span tubes.Experimental verficication of the computer prediction using a multi-span single tube apparatus has been performed in air and in water with clearance supports and with no-clearance supports. The natural frequencies, support impact forces and mid-span displacements of the experimental and analytical results were compared. The results are in fairly good agreement. An example is used to illustrate the application of the computation technique to a hypothetical heat exchanger tube. The life of the hypothetical heat exchanger tube is estimated based on the predicted support impact forces and existing wear data.     

The elbow centroid for hockey stick heat exchanger designs with hexagonal tube bundles

In nuclear component design it is the conservative (and usual) practice to assume that the center of gravity of 90° elbows lies on the centerline of the elbow. However, there is some logic for a more exact design with explicit safety factors, as opposed to safe assumptions. An exact design improves the accuracy in determining the overall vessel center of gravity, in assessing effects of seismically induced moments, in determining mechanical loads at vessel supports, and in calculating discontinuity stresses at elbow junctions.     

On the dynamic spatial response of a heat exchanger tube with intermittent baffle contacts

Flow-induced vibration in heat exchanger tubes can result in fretting wear at the baffle supports and subsequent tube failure. As one step in correlating the random flow excitation to the rate of fretting wear, this paper presents a dynamic finite element technique for predicting the motions and baffle contact forces of a single heat exchanger tube. Using a modal superposition approach, the modal equations of motion are generated and numerically integrated. The predicted results are compared with experimental data for both planar and spatial vibration of harmonically-excited cantilevered beams with a clearance support at the free end.     

BACCHUS; A numerical approach to two-phase flow in a rod bundle

We present in this paper the computer code BACCHUS, to analyze the thermal-hydraulics in a rod bundle in single or two-phase flow regime. The model is 2-D and uses the porous body approach. The two-phase model is an extension of the classical homogeneous model, and includes a differential non-equilibrium equation. Results are shown for the extension of the boiling region in a 19-pin bundle.     

A review of flow-induced vibrations in reactors and reactor components

This paper presents a review of the state of the art of two classes of vibration problems encountered in reactors and reactor peripherals: namely, vibration of cylindrical structures induced by cross flow and by axial flow. A historical perspective is given first, in which the milestone contributions that have advanced the state of the art are highlighted. Then recent developments in the last decade, with emphasis on those in the last three years, are discussed, concluding with an assessment of the state of understanding of the fluidelastic mechanisms involved, on the one hand, and of predictive tools available to the designer, on the other.     

Characteristics of gas-liquid two-phase flow in a capillary tube

Gas-liquid two-phase phenomena in capillary tubes were investigated with special attention on the flow patterns, the time varying void fraction and pressure loss. The directions of flow were vertical upward, horizontal and vertical downward. Pipe inner diameters used were 1 mm, 2.4 mm and 4.9 mm. As a result it is made clear that due to the strong effects of surface tension the flow pattern is not severely affected by the direction of flow, the smaller the pipe inner diameter, the easier the formation of liquid slug and the pressure loss in a unit length takes much larger than the estimated value by the Chisholm equation.     

Thermal hydraulic study on a high-temperature gas–gas heat exchanger with helically coiled tube bundles

An experimental study was carried out to investigate flow-induced vibration, heat transfer and pressure drop of helically coiled tubes of an intermediate heat exchanger (IHX) for the HTTR, using a full-size partial model and air as the fluid. The test model has 54 helically coiled tubes separated into three layer bundles, surrounding the center pipe. The vibration of the tube bundles was mainly at the center pipe, and the individual vibrations of the tube bundles were not significant under the operation conditions of the IHX. The heat transfer of the tube outside, due to forced convection, was obtained as a function of Re^0.51Pr^0.3, and the friction factor, depending on the tube arrangement, was proportional to Re^−0.14.     

非能动余热排出热交换器流动和传热数值模拟

非能动余热排除系统(Passive Residual Heat Removal system,PRHR)是非能动核电厂的重要安全设施,在全厂断电事故下,大部分的堆芯衰变热是通过PRHR热交换器传递至内置换料水箱(In-containment Refueling Water Storage Tank,IRWST)。但PRHR热交换器属于大型非稳态换热器,其传热机理十分复杂。基于PRHR系统的重要性和复杂性,有必要研究PRHR系统的流动和传热特性。利用计算流体动力学(Computational Fluid Dynamics,CFD)软件针对非能动堆芯冷却系统试验装置中的PRHR系统进行建模计算,分析了PRHR热交换器及IRWST的流动和传热特性,发现IRWST内部沿垂直高度上呈现明显的温度分层现象,温度沿水平方向的分布趋于均匀;IRWST内部的流动主要是沿着C型传热管竖直段向上流动,流速逐渐增大,但在两相阶段,水箱上部区域流动明显增强;C型传热管上部水平段和竖直段上部区域的换热系数要明显高于其它区域,且在上部水平段与竖直段连接弯管处换热系数最大,在两相阶段,上部区域的换热系数明显增大。     

Modeling of condensation heat transfer for a PRHRS heat exchanger in a SMART-P plant

A theoretical model using a heat and mass transfer analogy and a simple model using Lee and Kim's [Lee, K.-Y., Kim, M.H., 2008a. Experimental and empirical study of steam condensation heat transfer with a noncondensable gas in a small-diameter vertical tube. Nucl. Eng. Des. 238, 207-216] correlation were developed to investigate steam condensation in the presence of a noncondensable gas inside a vertical tube submerged in pool water. Rohsenow's correlation was used to consider the secondary pool-boiling heat transfer. Both models were assessed with the experimental data of Oh and Revankar [Oh, S., Revankar, S.T., 2005a. Investigation of the noncondensable effect and the operational modes of the passive condenser system. Nucl. Technol. 152, 71-86; Oh, S., Revankar, S.T., 2005b. Effect of noncondensable gas in a vertical tube condenser. Nucl. Eng. Des. 235, 1699-1712; Oh, S., Revankar, S.T., 2005c. Complete condensation in a vertical tube passive condenser. Int. Commun. Heat Mass Trans. 32, 593-602; Oh, S., Revankar, S.T., 2005d. Analysis of the complete condensation in a vertical tube passive condenser. Int. Commun. Heat Mass Trans. 32, 716-727; Oh, S., Revankar, S.T., 2006. Experimental and theoretical investigation of film condensation with noncondensable gas. Int. J. Heat Mass Trans. 49, 2523-2534; Oh, S., Gao, H., Revankar, S.T., 2007. Investigation of a passive condenser system of an advanced boiling water reactor. Nucl. Technol. 158, 208-218] for low pressure and Kim [Kim, S.J., 2000. Turbulent film condensation of high pressure steam in a vertical tube of passive secondary condensation system. Ph.D. dissertation, Korea Advanced Institute of Science and Technology] for high pressure, which were obtained from in-tube steam condensation with air in the pool water. These models predicted the data of Oh and Revankar well, but they slightly underestimated the data of Kim. The design of the Passive Residual Heat Removal System (PRHRS) condensation heat exchanger was evaluated with the theoretical model at real operating conditions (e.g., secondary pool-boiling, high system pressure). The PRHRS condensation heat exchanger designed was estimated to remove sufficiently the remaining heat in a reactor during a major accident.     

Assessment of flow induced vibration in a sodium-sodium heat exchanger

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam. It is a liquid metal sodium cooled pool type fast reactor with all primary components located inside a sodium pool. The heat produced due to fission in the core is transported by primary sodium to the secondary sodium in a sodium to sodium Intermediate Heat Exchanger (IHX), which in turn is transferred to water in the steam generator. PFBR IHX is a shell and tube type heat exchanger with primary sodium on shell side and secondary sodium in the tube side. Since IHX is one of the critical components placed inside the radioactive primary sodium, trouble-free operation of the IHX is very much essential for power plant availability. To validate the design and the adequacy of the support system provided for the IHX, flow induced vibration (FIV) experiments were carried out in a water test loop on a 60° sector model. This paper discusses the flow induced vibration measurements carried out in 60° sector model of IHX, the modeling criteria, the results and conclusion.     

A new LMR steam generator free from SWR with a double tube bundle configuration

This paper describes the activities made at KAERI to develop an advanced liquid metal reactor (LMR) steam generator which is free from a sodium water reaction (SWR) to resolve the concern of the SWR possibility and improve the economic features of the LMR. The steam generator design houses two tube bundles that are functionally different and its tube bundle region is radially or vertically divided into two. The SG is equipped with hot and cold fluid tube bundles, a medium fluid and a pump. It prevents the occurrence of the sodium water reaction while sodium is still used as the coolant for the primary heat transport system. The feasibility of using the SG with a double tube bundle for an actual use in a LMR plant is evaluated by setting up the skeleton of the NSSS for various possible configurations of the SG tube bundles.Analysis was made for various types of the new steam generator with a double tube bundle. Since the heat transfer in the SG is made among three different fluids, it has unique heat transfer characteristics. The analysis showed the possibility of the occurrence of an undesirable reversed heat transfer not only in the integrated single-region bundle type but also in the integrated double-region bundle type. The performance analysis revealed practical performance characteristics for the various bundle configurations. Also the feasibility study for the various NSSS configurations with the new SG is described.     

Sodium boiling experiments in a seven-pin bundle: Flow patterns and two-phase pressure drop

Two series of quasi-steady state sodium boiling experiments have been carried out in an electrically heated seven-pin bundle. The power levels (130–170 and 30–40 W/cm²) and other test conditions were selected to correspond to the core and radial breeder zones of a typical LMFBR. The test procedure involved the gradual reduction of mass flow rate through the bundle in a simulation of the consequences of a slowly growing blockage in the lower part of a reactor subassembly. By this means it was possible to study the development of quasi-steady state boiling up to the onset of permanent dryout. The results obtained provide information on flow regimes in the two-phase region, vapour qualities and flow rates at which cooling of the bundle can be effectively maintained, and the ultimate incidence of dryout. A relation for the two-phase pressure drop multiplier obtained from adiabatic pressure drop measurements in this geometry is given and compared with earlier results from single-channel geometry tests.