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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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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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
Securing the advanced reactor fleet
Physical protection accounts for a significant portion of a nuclear power plant’s operational costs. As the U.S. moves toward smaller and safer advanced reactors, similar protection strategies could prove cost prohibitive. For tomorrow’s small modular reactors and microreactors, security costs must remain appropriate to the size of the reactor for economical operation.
T. Ozawa, T. Abe
Nuclear Technology | Volume 156 | Number 1 | October 2006 | Pages 39-55
Technical Paper | Fuel Cycle and Management | doi.org/10.13182/NT156-39
Articles are hosted by Taylor and Francis Online.
Annular fuel is very beneficial for fast reactors because of its availability for both high power and high burnup. Most of the annular pellets irradiated up to high burnup showed central-hole shrinkage due to deformation and restructuring during irradiation. This shrinkage has a great influence on power-to-melt, which is a main factor in deciding the maximum power in the fuel design. To predict precisely the central-hole shrinkage during irradiation, the CEPTAR code was developed and verified by using the results of various experiments. In this code, the central-hole diameter is decided in accordance with the law of conservation of mass by using the radial profile of fuel density computed with the void migration model, and its deformation caused by the thermal expansion, swelling, and creep is computed by stress-strain analysis using the approximation of plane strain. Furthermore, this code can also estimate the effect of joint oxide gain (JOG) observed in a gap between the cladding and the fuel pellet with high burnup, which tends to decrease the fuel swelling and to improve the gap conductance due to deposition of solid fission product to the JOG layer. In this paper, an outline of the CEPTAR code and the results of verification are presented.