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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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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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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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Urenco USA feeds UF6 into new U.S. commercial enrichment cascade
Urenco USA has initiated production of enriched uranium in its newest gas centrifuge enrichment cascade—the first in a planned expansion of its Eunice, N.M., facility announced in July 2023. When the expansion is complete, early in 2027, the site will have increased its capacity by about 15 percent, adding about 700,000 separative work units (SWU) per year, the company said May 19.
Maria Pusa, Jaakko Leppänen
Nuclear Science and Engineering | Volume 164 | Number 2 | February 2010 | Pages 140-150
Technical Paper | doi.org/10.13182/NSE09-14
Articles are hosted by Taylor and Francis Online.
The topic of this paper is the computation of the matrix exponential in the context of burnup equations. The established matrix exponential methods are introduced briefly. The eigenvalues of the burnup matrix are important in choosing the matrix exponential method, and their characterization is considered. Based on the characteristics of the burnup matrix, the Chebyshev rational approximation method (CRAM) and its interpretation as a numeric contour integral are discussed in detail. The introduced matrix exponential methods are applied to two test cases representing an infinite pressurized water reactor pin-cell lattice, and the numerical results are presented. The results suggest that CRAM is capable of providing a robust and accurate solution to the burnup equations with a very short computation time.