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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.
B. Cazalis, J. Desquines, C. Poussard, M. Petit, Y. Monerie, C. Bernaudat, P. Yvon, X. Averty
Nuclear Technology | Volume 157 | Number 3 | March 2007 | Pages 215-229
Technical Paper | Reactivity-Initiated Accident (RIA) | doi.org/10.13182/NT07-A3814
Articles are hosted by Taylor and Francis Online.
An assessment of the mechanical properties of the highly irradiated fuel claddings under high strain rate has been carried out in the framework of the PROMETRA program undertaken by the French Institut de Radioprotection et de Sûreté Nucléaire in collaboration with Electricité de France and Commissariat à l'Energie Atomique (CEA). Three types of tests, including burst tests, hoop and axial tensile tests, have been performed at CEA-Saclay hot laboratories to determine the cladding tensile properties to use in the SCANAIR code. The prototypicality of each test with regard to the reactivity-initiated accident loading conditions can be addressed and analyzed in terms of strain or stress ratio. The high-strain-rate ductile mechanical properties of irradiated ZIRLO and M5 alloys derived from the PROMETRA program and their comparison to the stress-relieved irradiated Zircaloy-4 are reported. Then, the clad brittle behavior, in particular for highly corroded or spalled Zircaloy-4 cladding, is investigated.
