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UNM-ISNPS makes a strong presence at national meeting
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Ferris Engineering Center
Room 1120

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Institute for Space & Nuclear Power Studies
1 University of New Mexico
Albuquerque, NM 87131

Phone: 505.277.5442
Fax: 505.277.5433

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BREAKING NEWS - Master's Thesis Defense

TBoyce W. Travis and ISNPS committee.

On Monday, 26 November 2012, Boyce W. Travis (middle front), a Research Assistant with the Institute for Space and Nuclear Power Studies, successfully defended his Master's Thesis in Nuclear Engineering. Thesis committee included Research Assistant Professor Jean-Michel Tournier (front right), Adjunct Associate Professor Sal Rodriguez (front left), and Committee Chair and Thesis Advisor, Regents` Professor Mohamed S. El-Genk.

An Effective Methodology for Thermal- Hydraulics Analysis of a VHTR Core and Fuel Elements


The Very High Temperature Reactor (VHTR), a Generation-IV design, is graphite moderated and helium cooled and operates at an exit temperature of up to 1273 K, for generating electricity at a plant thermal efficiency upwards of 48% and the co-generation of process heat for hydrogen production and other industrial uses. This research developed an effective thermal-hydraulics analyses methodology that markedly reduces the numerical meshing requirements and computational time. It couples the helium's 1-D convective flow and heat transfer in the channels to 3-D heat conduction in graphite and fuel compacts of VHTR fuel elements. Besides the helium local bulk temperature, the heat transfer coefficient is calculated using a Nusselt number correlation, developed and validated in this work. In addition to omitting the numerical meshing in the coolant channels, the simplified analysis methodology effectively decreases the total computation time by a factor of ~ 33 - 40 with little effect on the calculated temperatures (< 5 K), compared to a full 3-D thermal-hydraulics analysis. The developed convective heat transfer correlation accounts for the effect of entrance mixing in the coolant channels, where z/D < 25. The correlation compares favorably, to within 12%, with Taylor's (based on high temperature hydrogen heat transfer) and to within 2% of the calculated results for full 3-D analyses of a VHTR single channel module and multiple channels in the fuel elements. The simplified methodology is used to investigate the effects of helium bypass flow in interstitial gaps between fuel elements and of the helium bleed flow in control rod channels on calculated temperatures in the VHTR fuel elements. Thermal-hydraulics analysis of a one-element high and of a full height VHTR 1/6 core are also conducted. Results show that the interstitial bypass flow increases the temperatures near the center of the core fuel elements by 10-15 K, while reducing temperatures along the edges of the elements by ~30 K. Without helium bypass flow, hotspots may occur at the location of burnable poison rods in the fuel elements, depending on the assumed volumetric heat generation rate in the rods. The helium bleed flow through the control rod channels reduces temperatures near them by 2-5 K, and only slightly increases the temperatures within the rest of the core fuel elements. In the VHTR 1/6 core thermal-hydraulics analysis, the helium bypass flow decreases the heat transfer from the core fuel elements to the adjacent radial graphite reflector blocks. Results demonstrate the effectiveness of the developed methodology and its potential use in future thermal-hydraulics design and in the safety analyses of VHTRs.