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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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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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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Candidates for leadership provide statements: ANS Board of Directors
With the annual ANS election right around the corner, American Nuclear Society members will be going to the polls to vote for a vice president/president-elect, treasurer, and members-at-large for the Board of Directors. In January, Nuclear News published statements from candidates for vice president/president-elect and treasurer. This month, we are featuring statements from each nominee for the Board of Directors.
C. M. Walter, P. G. Shewmon, J. P. Bacca
Nuclear Technology | Volume 11 | Number 1 | May 1971 | Pages 38-44
Technical Paper | Fuel | doi.org/10.13182/NT71-A30900
Articles are hosted by Taylor and Francis Online.
A metallic fuel-element modeling code (BEMOD) has been developed to describe the irradiation behavior of EBR-II driver fuels. BEMOD has been applied to both the present Mark LA and the advanced Mark II driver fuels. Good agreement on cladding diameter changes as a function of burnup is obtained between calculations and measurements on irradiated fuel elements. At a reactor power of 50 MW(th), the code calculations indicate that the Mark IA element is capable of about 3.5 at.% before a cladding ΔD/D of 2% is expected, while the Mark II design should be capable of about twice that burnup before a similar cladding ΔD/D is attained. The increase in reactor power to 62.5 MW(th) appears to have no appreciable effect on the above values.
