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Conference Spotlight
2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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The journey of the U.S. fuel cycle
Craig Piercycpiercy@ans.org
While most big journeys begin with a clear objective, they rarely start with an exact knowledge of the route. When commissioning the Lewis and Clark expedition in 1803, President Thomas Jefferson didn’t provide specific “turn right at the big mountain” directions to the Corps of Discovery. He gave goal-oriented instructions: explore the Missouri River, find its source, search for a transcontinental water route to the Pacific, and build scientific and cultural knowledge along the way.
Jefferson left it up to Lewis and Clark to turn his broad, geopolitically motivated guidance into gritty reality.
Similarly, U.S. nuclear policy has begun a journey toward closing the U.S. nuclear fuel cycle. There is a clear signal of support for recycling from the Trump administration, along with growing bipartisan excitement in Congress. Yet the precise path remains unclear.
Dae-Hyun Hwang, Kyong-Won Seo, Chung-Chan Lee
Nuclear Technology | Volume 158 | Number 2 | May 2007 | Pages 219-228
Technical Paper | Nuclear Reactor Thermal Hydraulics | doi.org/10.13182/NT07-A3837
Articles are hosted by Taylor and Francis Online.
Critical heat flux (CHF) in rod bundles is a parameter of great importance for the thermal-hydraulic design and safety analysis of advanced light water reactors. An experimental investigation has been conducted for the 19-rod hexagonal test bundles with a tightly spaced nonsquare arrangement of heater rods. The parametric effects on the CHF were examined for the heated length, the unheated rod, and the nonuniform axial power shape. As a result, a pertinent CHF correlation has been developed on the basis of the bundle cross-sectional averaged conditions. The available CHF database for rod bundles with square and nonsquare rod pitches was employed for the assessment of representative CHF correlations that were applicable to the round tubes and rod bundles. The database covered a wide range of operating conditions and test bundle geometries that are applicable to advanced light water reactors. The prediction accuracy of the CHF correlations was evaluated on the basis of the local thermal-hydraulic conditions calculated by a subchannel analysis code.
