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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.
Manwoong Kim, Hyun-Koon Kim, Hho-Jung Kim, Su Hyon Hwang, In Seob Hong, Chang Hyo Kim
Nuclear Technology | Volume 156 | Number 2 | November 2006 | Pages 159-167
Technical Paper | Reactor Safety | doi.org/10.13182/NT06-A3782
Articles are hosted by Taylor and Francis Online.
The purpose of this study is the development and verification of the coupled code system SCAN and RELAP-CANDU for transient analysis of a Canada deuterium uranium (CANDU) reactor. For this purpose, a spatial kinetics calculation module is developed and implemented in SCAN, a three-dimensional (3-D) CANDU-pressurized heavy water reactor neutronics design and analysis code. Then, a dynamic linked library of the SCAN code is generated for the integration with RELAP-CANDU.The RELAP-CANDU code has been developed for best-estimate transient simulation of CANDU reactor coolant systems based on the RELAP5 code. The SCAN code is a 3-D neutronic calculation code, which is composed of both unified nodal methods based on coarse-mesh finite difference method solutions to the time-dependent two-group diffusion equations.To verify the reliability of the coupled code system RELAP-CANDU/SCAN, the 40% reactor inlet header break accident, the 100% reactor outlet header break accident, and the pump suction pipe break are analyzed. The proposed coupled thermal-hydraulic and neutronic analyses methodology shows that there is an important margin in the traditional accident analysis.
