Investigation of natural convection in large pools

Krepper, Eckhard; Hicken, E.; Jaegers, H.

doi:10.1016/s0142-727x(02)00183-2

Cited by 38 publications

(13 citation statements)

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“…As an example, blockage of flow channel is one of the postulated accidents that have been incorporated into the one-dimensional (1D) simulation, assuming a valve at the inlet of the blocked channel, without any consideration of any spatial variations in the geometry that could lead to blockage of flow channels. In the past few decades CFD simulations have been introduced into the subject [e.g., [8][9][10][11], first simulating the most heat-stressed subchannels in two dimensions (2D), and later a full three-dimensional (3D) simulation of a sector of a fuel bundle is introduced. Interesting phenomena were observed, including the probable existence of secondary flows, swirl, and so on that were not possible to observe using one-dimensional simulations (Toth and Aszodi [12], Chang and Tavoularis [13,14], Tzanos [15], and many others).…”

Section: Introductionmentioning

confidence: 99%

Three-Dimensional, Numerical Investigation of Flow and Heat Transfer in Rectangular Channels Subject to Partial Blockage

Salama

El-Amin

Sun

2014

Heat Transfer Engineering

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Numerical simulation of flow and heat transfer in two adjacent channels is conducted with one of the channels partially blocked. This system simulates typical channels of a material testing reactor. The blockage is assumed due to the buckling of one of the channel plates inward along its width. The blockage ratio considered in this work is defined as the ratio between the cross-sectional area of the blocked and the unblocked channel. In this work, we consider a blockage ratio of approximately 40%. However, the blockage is different along the width of the channel, ranging from 0% at the end of the channel to 90% in the middle. The channel walls are sandwiching volumetric heat sources that vary spatially as chopped cosine functions. Interesting patterns are highlighted and investigated. The reduction in the flow area of one channel results in the flow redistributing among the two channels according to the changes in their hydraulic conductivities. The results of the numerical simulations show that the maximum wall temperature in the blocked channel is well below the boiling temperature at the operating pressure. INTRODUCTIONThere is no doubt that materials used in the design and operations of nuclear reactors are subjected to a harsh environment that is very unusual. In addition to structural stresses resulting from the interaction of the different components of reactor systems and the thermal stresses associated with the generation of heat within the core, which may be common in several engineering applications, there are also induced stresses that are unique to nuclear reactor cores and may not be found in any other application. For example, the exposure of these material to continuous bombardment with neutrons and fission products (irradiation) results in new phenomena such as swelling of fuel elements and the wear of containment materials. It is therefore important that experimental studies on irradiation effects in fuel and reactor structural materials be conducted to accurately predict their Color versions of one or more of the figures in the article can be found online at www.tandfonline.com/uhte. mechanical behavior and the possible changes in their nuclear chemistry. Unfortunately, it is not possible to mimic such an environment in a single isolated experiment. In other words, the only way to study such effects is within nuclear reactors themselves, which motivated the design of special types of nuclear reactors, called material testing reactors (MTR). These are designed to address in considerable detail the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimize safe operations of existing and future power reactors. MTR typically operate at low temperatures (coolant below 100 • C), the operating conditions are severe in a sense. That is, while power reactor fuel operates at power density of about 5 kW/cm 3 , a research reactor fuel may be at 17 kW/cm 3 in the fuel region. Furthermore, in MTR, bu...

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Section: Introductionmentioning

confidence: 99%

Three-Dimensional, Numerical Investigation of Flow and Heat Transfer in Rectangular Channels Subject to Partial Blockage

Salama

El-Amin

Sun

2014

Heat Transfer Engineering

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“…In order to validate the present CFD model, to accurately predict the natural convection and heat transfer for a pilot scale (10 L) and large scale GDWP ($10,000 m 3 ), experimental data from Gandhi et al (2011) and Krepper et al (2002) was used. The details of their experimental set up are given in Table 3.…”

Section: Model Validationmentioning

confidence: 99%

“…3 are in close agreement with experimental measurements and clearly indicate the accumulation of hot fluid at the top. Krepper et al (2002) investigated the heat transfer capability of an emergency condenser consisting of 8 tube bundle assembly and submerged in a horizontal cylindrical pool (18 m 3 ). The time varying temperature profiles at two axial locations shown in Fig.…”

Section: Model Validationmentioning

confidence: 99%

“…Comparison of CFD model and experimental data for the temperature distribution for the geometry ofKrepper et al (2002); Experimental value ( ) at y = 1.Comparison of CFD model and experimental data for the temperature distribution for the geometry ofGandhi et al (2011); (1) x/W = 0.01, experiment (h), OpenFOAM-1.6 (-); (2) x/W = 0.1, experiment (D), OpenFOAM-2.2 (--). by standard k-e model as shown inFig.…”

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confidence: 99%

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3D CFD simulations to study the effect of inclination of condenser tube on natural convection and thermal stratification in a passive decay heat removal system

Minocha

Joshi

Nayak

et al. 2016

Nuclear Engineering and Design

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“…The code allows in-depth analysis of local temperature and flow fields around fuel geometries as well as flow in reactor coolant system [1][2] [3][4] [5]. While the predictive capabilities of such codes provide 3D view of thermal flow changes, the certainty of such prediction relies on reactor data that essentially provide code validation.…”

Section: Introductionmentioning

confidence: 99%