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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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Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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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NRC cuts fees by 50 percent for advanced reactor applicants
The Nuclear Regulatory Commission has announced it has amended regulations for the licensing, inspection, special projects, and annual fees it will charge applicants and licensees for fiscal year 2025.
C. H. Lee, Y. J. Kim, J. W. Song, C. O. Park
Nuclear Science and Engineering | Volume 124 | Number 1 | September 1996 | Pages 160-166
Technical Paper | doi.org/10.13182/NSE96-A24231
Articles are hosted by Taylor and Francis Online.
The spectral history problem encountered in reconstructing local homogeneous power distributions is investigated. Because of difficulties in most nodal codes concerning spectral interactions between neighboring assemblies when rebuilding the local power distribution, nodal codes assume a constant spectrum or do not properly consider local spectrum variations within an assembly. A simple, fuel-type-independent method is presented to eliminate the spectrum-induced errors from local homogeneous powers within an assembly over the entire burnup range. The method, which is generalized for its application to any fuel type in the entire assembly burnup domain, uses the proportional relationship between macroscopic cross sections and average spectral history indices. Verification results through embedded calculations and an actual core calculation show that local homogeneous power errors are reduced to the same magnitude as flux errors. The error reduction is conspicuous in the cases of mixed-oxide and highly poisoned fuel assemblies.