Large-Scale Pulsation Detection by Means of Temperature Measurements

2008 ◽  
Vol 130 (11) ◽  
Author(s):  
N. Silin ◽  
V. Masson ◽  
A. Rauschert
Keyword(s):  
Nuclear Fuel ◽  
Large Scale ◽  
Axial Flow ◽  
Temperature Measurements ◽  
Time Resolved ◽  
Structure Information ◽  
Rod Bundle ◽  
Nuclear Fuel Assembly ◽  
Large Scale Flow ◽  
Fuel Elements

In the present work we explore the potential of time-resolved temperature measurements to obtain information on large-scale pulsations in a rod bundle geometry with axial flow. Large-scale flow pulsation is the phenomenon that dominates the turbulent mixing between the subchannels of rod bundles, which explains why it is of great importance for the design or assessment of nuclear fuel elements. The objective of the present work is to determine the characteristics of large-scale pulsations that can be used for the verification or validation of computational fluid dynamics code results. The method proposed is to generate a temperature gradient across the location of flow pulsations and to measure the time-varying temperature field downstream. Pulsation characteristic times, lengths, and traveling speed have been obtained. This study has been performed in a rod bundle similar to a nuclear fuel assembly and the results obtained are in good agreement with previous works on similar geometries. The technique can be applied to obtain additional large-scale structure information in test sections designed for thermal measurements, in situations where convection is dominated by these structures.

Turbulent spots in a channel: large-scale flow and self-sustainability

2013 ◽  
Vol 731 ◽  
Author(s):  
Grégoire Lemoult ◽  
Jean-Luc Aider ◽  
José Eduardo Wesfreid
Keyword(s):  
Large Scale ◽  
Large Time ◽  
Rectangular Channel ◽  
Small Scale ◽  
Turbulent Spot ◽  
Time Resolved ◽  
Image Velocimetry ◽  
Streamwise Streaks ◽  
Large Scale Flow ◽  
Scale Motion

AbstractUsing a large-time-resolved particle image velocimetry field of view, a developing turbulent spot is followed in space and time in a rectangular channel flow for more than 100 advective time units. We show that the flow can be decomposed into a large-scale motion consisting of an asymmetric quadrupole centred on the spot and a small-scale part consisting of streamwise streaks. From the temporal evolution of the energy of the streamwise and spanwise velocity perturbations, it is suggested that a self-sustaining process can occur in a turbulent spot above a given Reynolds number.

Preliminary Investigation on Convective Heat Transfer of Rod Bundle Flow With Spacer Grid for Nuclear Fuel Assembly Application

2015 ◽  
Author(s):  
Chi Young Lee ◽  
Chang Hwan Shin ◽  
Wang Kee In ◽  
Dong Seok Oh ◽  
Tae Hyun Chun
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Convective Heat Transfer ◽  
Convective Heat ◽  
Heat Transfer Performance ◽  
Axial Flow ◽  
Transfer Performance ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Nuclear Fuel Assembly

The convective heat transfer of rod bundle flow with spacer grid was investigated preliminarily for nuclear reactor core application. As the test fluid, the water was used. To simulate the nuclear fuel assembly, 4×4 rod bundle with P/D (=pitch between rods/rod diameter) of ∼1.35 was prepared together with a spacer grid with twist-mixing vane. A single heated section with five thermocouples embedded in the surface along the circumferential direction was installed around the center subchannel. The measurements of wall temperatures were carried out upstream and downstream of spacer grid. For the rod bundle flow at the inlet of spacer grid (i.e., upstream of spacer grid), the wall temperatures at the gap and subchannel centers exhibited the higher and lower, respectively, which was because in the subchannel center, the axial flow velocity became higher, as compared with the gap center. On the other hand, downstream of spacer grid, the rod wall toward the tip of twist-mixing vane showed the lowest temperature in the measurements along the circumferential direction of rod wall. Near the twist-mixing vane, the averaged wall temperature was observed to be remarkably low, which implies that the twist-mixing vane is an effective tool to enhance the convective heat transfer performance. However, along the axial flow direction behind the spacer grid, the averaged wall temperatures became to increase, and the enhancement of convective heat transfer performance by mixing vane faded away.

Sub-channels thermal-hydraulic analysis of rod bundle nuclear fuel elements under rolling motions

2021 ◽  
pp. 108874
Author(s):  
M. Behzadi ◽  
A. Zolfaghari ◽  
M.R. Abbassi ◽  
A. Norouzi ◽  
M.M. Mirzaeegoudarzi
Keyword(s):  
Nuclear Fuel ◽  
Hydraulic Analysis ◽  
Rod Bundle ◽  
Fuel Elements ◽  
Thermal Hydraulic Analysis

Spectral Response to Harmonic Excitation of Rods in a Confined Nuclear Fuel Mini-Bundle

2005 ◽  
Author(s):  
Radu O. Pomiˆrleanu
Keyword(s):  
Nuclear Fuel ◽  
Spectral Response ◽  
Variable Mass ◽  
Axial Flow ◽  
Harmonic Excitation ◽  
Viscous Effects ◽  
Rod Bundle ◽  
Fuel Bundle ◽  
Computational Fluid Dynamics Cfd ◽  
Euler Bernoulli Beam

In this paper, the effect of the confinement on the vibration of rods within nuclear fuel bundle is analyzed and discussed. Differential added damping of the rods depending on their position within the bundle adds to the classical development of modal and spectral analysis. This is accomplished by coupling the classical modeling of the variable-mass, Euler-Bernoulli beam on elastic supports with the two-dimensional computational fluid dynamics (CFD) estimation of motion-dependent effects, with the viscous effects integrally modeled. A 5×5 reduced-scale (shortened) rod bundle subject to room-temperature single-phase nominally axial flow within a square confinement is used to illustrate the response to a harmonically varying excitation. The harmonic analysis showed that there are significant differences in the response of the rods, depending on their position within the bundle.

A Mathematical Model for Transient Subchannel Analysis of Rod-Bundle Nuclear Fuel Elements

1973 ◽  
Vol 95 (2) ◽  
pp. 211-217 ◽  
Author(s):  
D. S. Rowe
Keyword(s):  
Mathematical Model ◽  
Nuclear Fuel ◽  
Flow Rate ◽  
Transient Flow ◽  
Momentum Equation ◽  
Flow Area ◽  
Inlet Flow ◽  
Rod Bundle ◽  
Inlet Flow Rate ◽  
Fuel Elements

This paper presents a mathematical method for analyzing transient flow and enthalpy transport in rod-bundle nuclear fuel elements during both boiling and nonboiling conditions. A mathematical model is formulated by dividing the bundle flow area into flow subchannels that are assumed to contain one-dimensional flow and are coupled to each other by turbulent and diversion crossflow mixing. The mathematical model neglects sonic velocity propagation and neglects temporal and spatial acceleration in the transverse momentum equation. A semiexplicit finite-difference scheme is used to perform a boundary-value solution where the boundary conditions are the inlet enthalpy, inlet flow rate, and exit pressure. Calculations are presented to show the effect of rapid changes in heat flux, inlet enthalpy, and inlet flow rate on the subchannel flow and enthalpy distribution in rod bundles.

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