Convective heat transfer for an annulus, having its inner cylinder wrapped with a helical wire, has been numerically studied. Constant heat flux boundary condition is applied at the inner wall of the annulus whereas the outer wall is insulated. The computer code with k-omega SST turbulence model has been validated with the results of bare annulus. Several numerical results such as variations in velocity, pressure, vorticity, turbulent kinetic energy, temperature, friction factor and Nusselt number are presented for various annulus diameter ratios (D / d = 1.5, 1.8 and 2.0), pitch ratios (P / d = 9.09, 18.18 and 30.30), wire diameter ratios (Dw / d = 0.15, 0.25 and 0.30) and Reynolds number (Re = 20000 to 180000). The zones of possible hot-spots in the cusp region close to the wrap-wire are discerned. © 2008 Elsevier Masson SAS. All rights reserved.

B. V.S.S.S. Prasad

Department of Mechanical Engineering

T. Sreenivasulu

augmentation

CYLINDRICAL ANNULI

Flow and heat transfer

HELICALLY WRAPPED WIRE

Hotspots

Heat exchangers

Heat transfer

Kinetic energy

Reynolds number

Turbulence

Turbulence models

Wire

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IIT Madras

International Journal of Thermal Sciences

Hydraulic characteristics of a 7 pin helical spacer fast reactor fuel bundle have been investigated with focus on flow non-uniformity in the central and peripheral sub-channels. It is seen that, there is large variation in the flow distribution among the sub-channels with significant flow bypass in the peripheral sub-channels. To improve the flow uniformity, the option of providing inserts in the peripheral sub-channels has been investigated. In this direction, three types of inserts, viz., semi-circular, triangular and circular inserts, have been studied. The improvement in the flow distribution among the central and peripheral sub-channels and the increase in friction factor as a result of inserts have been quantified. The investigation is based on 3 dimensional computational fluid dynamic simulations as well as experiments in water models. In the experiments the flow pattern at the exit of fuel pin bundle has been measured using Particle Imaging Velocimetry. The pressure drop has been measured using differential pressure transmitter. The measured flow pattern and pressure drop in the bundle are compared with predicted results. The detailed parametric studies indicate that, among the three inserts studied the triangular insert provides lowest friction factor and largest axial velocity fraction in the central sub-channels. The circular inserts offer the maximum friction factor. © 2019

Progress in Nuclear Energy

Computational fluid dynamic investigations on effects of inserts in fast reactor fuel bundle hydraulics

Fuel subassemblies of a fast reactor consist of fuel pin bundle with helically wound spacer wires, arranged in a triangular pitch within a hexagonal wrapper. The fuel pins are located within the subassembly. Further the subassembly comprises of a diffuser where the cross section changes from cylindrical to hexagonal, mixing plenum before the exit of pin bundle and a specially designed blockage adapter. Accurate assessment of the pressure drop in the fuel subassembly is essential to ensure adequate core cooling and design of sodium pump. Experimental determination of pressure drop characteristics in the subassembly by simulating the hydraulic condition in the subassemblies of the reactor core is considered essential as a better choice as correlations reported in the literature cannot be directly used for all the complex regions present in the subassembly. This is due to the fact that flows in the interconnecting sections are highly under developed. Further, the flow regime in a fuel subassembly varies from laminar (during shutdown heat removal under natural convection) to completely turbulent under full power condition. To understand the hydraulic characteristics of the 500 MWe Proto type Fast Breeder Reactor (PFBR) fuel subassembly, an experimental facility has been commissioned. Experiments on full scale subassembly with dummy fuel pins have been performed using water as simulant. Experiments have been conducted covering a wide range of Reynolds number encompassing laminar, transition and turbulent regimes. In the rod bundle, no abrupt changes in friction factor were observed when the flow changes from laminar to turbulent. The experimental results have been transposed using Euler number similarity, to determine the pressure drop for sodium flow in the reactor. Possible uncertainties in the experimental data have been quantified. Useful pressure drop correlations have been proposed for various section of the subassembly of a fast breeder reactor. Finally the results for fuel pin bundle region are compared with the data reported in the literature and a satisfactory agreement has been observed. © 2016 Elsevier Ltd

Annals of Nuclear Energy

Hydraulic characteristics of a fast reactor fuel subassembly: An experimental investigation

A comprehensive study of rod bundle thermal-hydraulics in a nuclear reactor, enables us with safe and reliable margins for its operation. Although computational fluid dynamics (CFD) is an ideal choice, the number of degrees of freedom and the detail increases exponentially with increase in bundle size. Performing a large number of simulations, over the entire range of operating conditions and other parametric combinations makes CFD prohibitively expensive. Hence, it is preferable to pursue simple and effective reduced order models. To this end, a wide range of sub-channel analysis codes have been devised in the literature. One such model is the popular deformation and intermixing analysis in nuclear assemblies (DIANA) algorithm. In the present study, this algorithm is numerically implemented and validated for different rod bundle configurations. The effect of eccentricity and blockage and their combined effect on the thermal hydraulics of a 19 rod bundle is systematically analyzed. The thermal-hydraulics simulations are performed in the eccentricity (e) range of 0.0 ≤ e ≤ 0.7. It was observed that, eccentricity leads to flow maldistribution in some sub-channels, which in turn effects the coolant temperature distribution at the outlet. Using the present approach, a number of what-if type scenarios such as, blockage, bundle deformation and their combinations are also numerically investigated. It was noticed that, the presence of blockage in a sub-channel results in a reduction in the local coolant mass flux and hence the local clad temperature shoots up. In the narrowest sub-channel, for a typical case of eccentricity e = 0.5 and blockage ratio (b) = 0.4, maximum decrease in mass flux was observed to be 85% and increase in the coolant and clad temperatures was found to be 150% and 304% respectively. To emphasize the utility of the present study, the simulations are extended for a 217 sub-assembly with multiple blockages. © 2016 Elsevier B.V. All rights reserved.

Nuclear Engineering and Design

Sub-channel analysis of rod bundle thermal hydraulics: Effect of eccentricity and blockage

Simultaneous development of flow and temperature fields in the entrance region of fast breeder reactor (FBR) fuel pin bundles with helical spacer wires has been investigated by three-dimensional computational simulations. The Reynolds number, pitch of helical spacer wire and number of pins in the bundle are systematically varied. It is found that the magnitude of mean cross-stream velocity in the fully developed region is inversely proportional to the helical pitch length and it is nearly independent of the number of pins. But, there is a strong correlation between the locations of spacer wire and the peak cross-stream velocity. Flow attains full development at an axial length of 70 times hydraulic diameter in all the cases and this length is found to be unaffected by the helical pitch length. Friction factor is seen to fluctuate periodically over a mean value and the fluctuation over each helical pitch corresponds to a specific position of helical wire. The mean value of the friction factor in the entrance region reduces below the mean value in the fully developed region contrary to that seen in ducted flows. The mean fully developed friction factor is inversely proportional to the helical pitch. But, it is independent of the number of pins in the bundle. The Nusselt number passes through multiple minima before attaining fully developed periodic fluctuations and its development is slower than that of friction factor. For larger number of pins thermal development length is larger. Traditionally, the correlations reported for fully developed flow are considered for core design. But, the present study indicates that this approach is not conservative. Further, the entrance region effects and the oscillations in the fully developed region have to be properly accounted in the core design. Nusselt number exhibits a strong dependence on helical pitch similar to that of friction factor. A correlation for Nusselt number is proposed as a function of helical pitch and other influencing parameters. The hotspot factor in the pin bundle is found to vary from 3 to 4. Further, it increases, when either Reynolds number is low or the number of pins is more. © 2013 Elsevier B.V.

Simultaneous development of flow and temperature fields in wire-wrapped fuel pin bundles of sodium cooled fast reactor

Purpose: The purpose of this paper is to investigate the effect of narrow gap on the fluid flow and heat transfer through an eccentric annular region is numerically. Flow through an eccentric annular geometry is a model problem of practical interest. Design/methodology/approach: The approach involves standard finite volume-based SIMPLE scheme. The numerical simulations cover the practically relevant Reynolds number range of 104-106. Findings: In the narrow gap region, temperature shoot up was observed due to flow maldistribution with an attendant reduction in the heat removal from the wall surfaces. CFD analysis is presented with the aid of, streamlines, isotherms, axial velocity contours, etc. The engineering parameters of interest such as, Nusselt number, wall shear stress, etc., is presented to study the effect of eccentricity and radius ratio. Research limitations/implications: The present investigation is a simplified model for the rod bundle heat transfer studies. However, the detailed study of sectorial mass flux distribution is a useful precursor to the thermal hydraulics of rod bundles. Practical implications: For nuclear reactor fuel rods, the effect of eccentricity is going to be detrimental and might lead to the condition of critical heat flux. A thorough sub-channel analysis is very useful. Social implications: Nuclear safety standards require answers to a wide a range of what-if type hypothetical scenarios to enable preparedness. This study is a highly simplified model and a first step in that direction. Originality/value: The narrow gap region has been systematically investigated for the first time. A detailed sectorial analysis reveals that, flow maldistribution and the attendant temperature shoot up in the narrow gap region is detrimental to the safe operation. © Emerald Group Publishing Limited.

Fulltext

International Journal of Numerical Methods for Heat and Fluid Flow

Numerical simulation of flow through an eccentric annulus with heat transfer

International Journal of Heat and Mass Transfer

A performance evaluation plot of enhanced heat transfer techniques oriented for energy-saving

Heat transfer enhancement in circular tubes using helical swirl generator insert at the entrance

International Communications in Heat and Mass Transfer

Heat transfer characteristics in a tube fitted with helical screw-tape with/without core-rod inserts

Heat Transfer Engineering

Convective Heat Transfer Coefficients in Concentric Annuli

AIAA Journal

Two-equation eddy-viscosity turbulence models for engineering applications

Flow and heat transfer characteristics in an annulus wrapped with a helical wire

Journal	International Journal of Thermal Sciences
ISSN	12900729
Open Access	No