Advanced reactor technologies have generated interest for their potential to reduce fossil fuel emissions, improve energy efficiency and cut down on nuclear waste. We have designed, constructed , and operated a specialized test facility to measure hydraulic parameters and validate computational tools used in reactor design and testing. The fast reactor assembly design used is significant, not only because of the complex inner knowledge it can provide about advanced reactors, but also because the Texas A&M experiment is using the largest transparent test fuel assembly of its kind to date.
Measurements of axial and azimuthal pressure drops, and velocity fields are collected and analyzed used the state-of-the-art instrumentation and advanced techniques.
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New experiments are being conducted to investigate the flow behavior at different locations in the bundle. New advance techniques of flow visualization are being explored. The new activity is sponsored by the Nuclear Energy University Program (NEUP), under the project DE-NE0008652
Journal Publications
1.Nguyen et al. (2017), “PIV Measurements of turbulent flows in a 61-pin Wire-Wrapped Hexagonal Fuel Bundle”, International Journal of Heat and Fluid Flow 65, 2017, pp 47-59.
2.Vaghetto et al. (2017), “Pressure Measurements in a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle”, J. Fluids Eng 140 (3).
3.Brockmeyer et al. (2017), “Numerical simulations for determination of minimum representative bundle size in wire wrapped tube bundles”, Nuclear Engineering and Design 322, 577-590
4.Nguyen et al. (2017), “Stereoscopic PIV measurements of near-wall flow in a tightly packed rod bundle with wire spacers”, Experimental Thermal and Fluid Science, 92, 2018.
5.Goth et al (2018), “PTV/PIV measurements of turbulent flows in interior subchannels of a 61-pin wire-wrapped hexagonal fuel bundle”, International Journal of Heat and Fluid Flow 71, 295-304.
6.Goth et al. (2018), “Comparison of experimental and simulation results on interior subchannels of a 61-pin wire-wrapped hexagonal fuel bundle”, Nuclear Engineering and Design 338, 130-136.
