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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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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Latest News
2024: The Year in Nuclear—July through September
Another calendar year has passed. Before heading too far into 2025, let’s look back at what happened in 2024 in the nuclear community. In today's post, compiled from Nuclear News and Nuclear Newswire are what we feel are the top nuclear news stories from July through September 2024.
Stay tuned for the top stories from the rest of the past year.
Karl Verfondern, Werner Schenk, Heinz Nabielek
Nuclear Technology | Volume 91 | Number 2 | August 1990 | Pages 235-246
Technical Paper | Safety of Next Generation Power Reactor / Fuel Cycle | doi.org/10.13182/NT90-A34431
Articles are hosted by Taylor and Francis Online.
The high fission product retention potential of coated particle fuel combined with inherently passive temperature controls guarantee almost complete fission product retention during an accident in a small modular high-temperature reactor. Extensive experimental results provide the basis for this claim to inherent safety. Models and codes have been developed to (a) predict realistic, or at least conservative, overall release rates from the primary circuit, (b) reduce the large number of experimental results to a small set of characteristic coefficients, and (c) predict release beyond experimental conditions. Conservative predictions of release from the core have been done using a traditional pressure vessel model for release from fuel particles and simplified diffusion models for fission product transport. This approach is based on experimental work that has been done on nearly all possible accident conditions and is limited by the finite number of experiments. Data reduction has been achieved with two different modeling approaches combined into a new model that is equally relevant to all volatile fission products. The safety design of the 200-MW(thermal) HTR-Modul is based on Kernforschungsanlage Jülich experimental results from fuel accident condition performance testing and the modeling effort has been applied to a safety review.