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Conference Spotlight
2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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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.
Yu-Chih Ko, Ching-Hui Wu, Min Lee
Nuclear Technology | Volume 155 | Number 1 | July 2006 | Pages 22-33
Technical Paper | Reactor Safety | doi.org/10.13182/NT06-A3743
Articles are hosted by Taylor and Francis Online.
Probabilistic safety assessment (PSA) uses a systematic approach to estimate the reliability and risk of a nuclear power plant (NPP). Over the past few years, severe accident management guidance (SAMG), which delineates the mitigation actions of core melt accidents of an NPP, has been developed to support operators and staff in the technical support center in dealing with those misfortunes. It can be expected that the implementation of SAMG will lower the containment failure frequency and reduce the amount of radionuclides released to the environment during the accident. The plant studied is the Maanshan NPP of Taiwan Power Company, which employs a Westinghouse-designed three-loop pressurized water reactor (PWR) with large dry containment.The containment system event trees and containment phenomenological event trees of the Level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The HCR (Human Cognitive Reliability) and THERP (Technique for Human Error Rate Prediction) models are used to quantify the human error probability (HEP) of all the actions in the Level-2 PSA model. The MAAP4 (Module Accident Analysis Program version 4) code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the frequency of most of the source term categories is reduced except the one in which both the reactor pressure vessel and containment are intact. The containment failure frequency is reduced by 14.8% after the implementation of SAMG. The frequency of containment early failure is reduced by 16.2%. Most of the reduction in the containment early failure frequency comes from the reduction in the induced steam generator tube rupture (STGR). The frequency of induced SGTR was reduced from 2.3 × 10-7/reactoryr to 1.0 × 10-8/reactoryr.
