In a nuclear fuel cask, the heat generating spent fuel rods are packed in a housing and the resulting bundle is placed inside a cask of thick outer shell made of materials like lead or concrete. The cask presents a wide variation in geometrical dimensions ranging from the diameter of the rods to the diameter of the cask. To make the problem tractable, first the heat generating rod bundle alone is considered for analysis and the effective thermal conductance of the bundle is correlated in terms of the relevant parameters. In the second part, the bundle is represented as a solid of equivalent thermal conductance and the attention is focused on the modelling of the cask. The first part, dealing with the effective thermal conductance is solved using Fluent software, considering coupled conduction, natural convection and surface radiation in the heat generating rod bundle encased in a hexagonal sheath. Helium, argon, air and nitrogen are considered as working media inside the bundle. A correlation is obtained for the critical Rayleigh number which signifies the onset of natural convection. A correlation is also developed for the effective thermal conductance of the bundle, considering all the modes of transport, in terms of the maximum temperature in the rod bundle, pitch-to-diameter ratio, bundle dimension (or number of rods), heat generation rate and the sheath temperature. The correlation covers pitch-to-diameter ratios in the range 1.1-2, number of rods ranging from 19 to 217 and the heat generation rates encountered in practical applications. The second part deals with the heat transfer modeling of the cask with the bundle represented as a solid of effective (or equivalent) thermal conductance. The mathematical model describes two-dimensional conjugate natural convection and its interaction with surface radiation in the cask. Both Boussinesq and non-Boussinesq formulations have been considered for convection. Numerical solutions are obtained on a staggered mesh with a pressure correction method using a custom-made Fortran code. The surface radiation is coupled to the conduction and convection at the solid-fluid interfaces. Steady-state results are obtained using time-marching. Results for various quantities of interest, namely, the flow and temperature distributions, Nusselt numbers, and interface temperatures, are presented. The Grashof number based on the volumetric heat generation and gap width is varied from 105 to 5 ×109. The emissivities of the interfaces are varied from 0.2-0.8 for the radiative calculations. The solid-to-fluid thermal conductivity ratio for the inner cylinder is varied in the range 5-20 in the parametric studies. Simulations are also performed with thermal conductivity calculated in an iterative manner from bundle parameters. The dimensionless outer wall conductivity ratio is chosen to correspond to cask walls made of lead or concrete. The dimensionless thickness (with respect to gap width) of the outer shell is in the range of 0.0825-1, while the inner cylinder dimensionless radius is 0.2. Air is the working medium in the cask for which the Prandtl number is 0.71. Correlations are obtained for the average temperatures and Nusselt numbers at the inner interface in terms of the parameters. The radiation heat transfer is found to contribute significantly to the heat dissipation.

Vertical up-flow of supercritical fluid in the subchannel of a heated rod bundle was numerically simulated using the Computational Fluid Dynamics (CFD) codes ANSYS CFX and ANSYS FLUENT. A total of seven cases from three different sets of experiments were simulated. Three-dimensional steady-state predictions of fluid velocity, pressure, and temperature were made using five versions of two-equation RANS turbulence models with accompanying wall treatments. In addition, the temperature distribution in a solid region comprising a heater and sheathing was also computed in some cases. The k-epsilon turbulence model, implemented using CFX and scalable wall functions, provided the numerical results that have the smallest overall deviation from experimental results for three of the seven cases, and predicts the experimental data of the remaining four cases reasonably well, unlike other turbulence models that severely over-predict the experimental data for wall surface temperature. / February 2016

The objective of this thesis research is to improve our understanding of the flow obstacle effect on heat transfer at supercritical and high subcritical pressures by experimentally studying the effect of different obstacles on heat transfer in two vertical upward-flow test sections: a 3-rod bundle and an 8 mm ID tube. The heat transfer measurements cover the region of interest of the Canadian Super-critical Water Cooled Reactor (SCWR). A thorough analysis of the obstacle effect on supercritical heat transfer (SCHT) was performed. In the 3-rod bundle, two types of obstacles were employed: wire wraps and low-impact grid spacers. Wire wraps were found to be more effective than grid spacers to enhance the SCHT. In the tubular test section, obstacles appeared to suppress the heat transfer deterioration (HTD) or decrease its severity; obstacles also generally enhanced the SCHT both in the liquid-like and the gas- like region. The experiment in the tubular test section revealed that, at certain flow conditions (low mass flux, low inlet subcooling), flow obstacles can have an adverse impact on the SCHT. A criterion to predict the onset of this adverse effect was developed. At high subcritical pressures, obstacles increased the CHF and reduced the maximum post-CHF temperature. A comparison of the experimental data with prediction methods for the SCHT, single phase heat transfer, CHF and post-dryout heat transfer was performed. Lastly, a new correlation to predict the enhancement in SCHT due to obstacles was developed for heat transfer in the liquid-like and gas-like regions.

<p> Wire-wrapped rod bundles are often used in nuclear reactors operating in a fast neutron spectrum, as designers seek to minimize neutron scattering by packing the fuel pins into a hexagonal lattice. Bundles with many rods have extensively been studied as representative of large fuel assemblies, however far fewer experiments have investigated bundles with 7 rods (7-pin bundles). The large difference in subchannel number between these bundles leads to 7-pin bundles having different pressure drop characteristics. The Versatile Test Reactor (VTR) sodium cartridge loop proposes to use a 7-pin bundle as its experimental core region, highlighting the need for additional data and models. The current work seeks to establish a better understanding of the pressure drop in 7-pin wire-wrapped rod bundles through scaled experiments and a novel pressure drop model. A scaling analysis is first performed to demonstrate the applicability of water experiments to the VTR sodium cartridge loop, before an experimental test facility is designed and constructed. Experiments are then performed at a range of Reynolds numbers to determine the pressure drop. Current models are able to predict the data well, but are complex and can be difficult to use. A comparatively simpler model is developed, based on exact laminar solutions of a simplified rod bundle, which also offers a theoretical lower bound for the pressure drop in wire-wrapped bundles. The proposed model compares well with the existing experimental database, able to predict bundle friction factor with an average absolute percent difference of 10.8%. This accuracy is also similar to existing correlations, while relying on fewer empirical coefficients. The theoretical lower bound is also used to identify several datasets in literature that may feature data that is systemically lower than the true pressure drop, which agrees with previous observations in literature. </p>