Goth, Nolan Edward (2018-12). Analysis of Experimental PIV/PTV Measurements on a Matched-Index-of-Refraction 61-Pin Wire-Wrapped Hexagonal Fuel Bundle. Doctoral Dissertation.
Thesis
In nuclear reactor thermal hydraulics modeling, detailed coolant velocity fields are essential to accurately anticipating flow induced vibrations and temperature distributions in the coolant and pins. Predictive computational tools are under development to produce these coolant velocity fields using computational fluid dynamics. Turbulence modeling validation is a necessary remaining step in the development of such tools. Prior liquid metal fast reactor experimental velocity field data is not suitable for computational fluid dynamics turbulence modeling validation. Therefore, a demand exists for experimental data to be collected using non-intrusive probes with sufficiently high spatiotemporal resolution and of a hydraulically similar fuel bundle. The primary objective of this research was to produce high spatiotemporal experimental velocity field data on a 61-pin wire-wrapped hexagonal fuel bundle such that commercial sodium fast reactor core design and computational fluid dynamics turbulence modeling validation may be performed. The primary phenomena of interest are the bulk swirl, local swirl, subchannel mixing, and bypass flow. Flow statistics such as ensemble-averaged velocity, root-mean-square fluctuating velocity, Reynolds stresses, Z-vorticity, and integral length scales were presented. The methodology to meet the primary objective was the utilization of a matched-indexof- refraction experimental flow facility with laser-based optical measurement techniques. A total of 9 different measurement locations at various bundle-averaged Reynolds numbers have been investigated to generate a total of 50 unique datasets. The scientific value of this work is twofold. It benefits the grand effort of turbulence modeling validation by providing new high spatiotemporal resolution experimental data for which benchmark activities may be performed. It also furthers the research and development of liquid metal fast breeder reactor core thermal hydraulics to progress the U.S Department of Energy's advanced reactor development agenda.
In nuclear reactor thermal hydraulics modeling, detailed coolant velocity fields are essential to accurately anticipating flow induced vibrations and temperature distributions in the coolant and pins. Predictive computational tools are under development to produce these coolant velocity fields using computational fluid dynamics. Turbulence modeling validation is a necessary remaining step in the development of such tools. Prior liquid metal fast reactor experimental velocity field data is not suitable for computational fluid dynamics turbulence modeling validation. Therefore, a demand exists for experimental data to be collected using non-intrusive probes with sufficiently high spatiotemporal resolution and of a hydraulically similar fuel bundle. The primary objective of this research was to produce high spatiotemporal experimental velocity field data on a 61-pin wire-wrapped hexagonal fuel bundle such that commercial sodium fast reactor core design and computational fluid dynamics turbulence modeling validation may be performed. The primary phenomena of interest are the bulk swirl, local swirl, subchannel mixing, and bypass flow. Flow statistics such as ensemble-averaged velocity, root-mean-square fluctuating velocity, Reynolds stresses, Z-vorticity, and integral length scales were presented. The methodology to meet the primary objective was the utilization of a matched-indexof- refraction experimental flow facility with laser-based optical measurement techniques. A total of 9 different measurement locations at various bundle-averaged Reynolds numbers have been investigated to generate a total of 50 unique datasets. The scientific value of this work is twofold. It benefits the grand effort of turbulence modeling validation by providing new high spatiotemporal resolution experimental data for which benchmark activities may be performed. It also furthers the research and development of liquid metal fast breeder reactor core thermal hydraulics to progress the U.S Department of Energy's advanced reactor development agenda.