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Mathematics & Computation
Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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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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Latest News
Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
John Loberg, Michael Österlund, Jan Blomgren, Klaes-Håkan Bejmer
Nuclear Science and Engineering | Volume 164 | Number 1 | January 2010 | Pages 69-79
Technical Paper | doi.org/10.13182/NSE09-17
Articles are hosted by Taylor and Francis Online.
The ratio between the thermal- and fast-neutron fluxes in a boiling water reactor depends on the void fraction. The density of the steam-water mixture present in the core determines the efficiency of the moderation of fast neutrons born in fission; therefore, the void fraction could be determined by means of a simultaneous measurement of the thermal- and fast-neutron fluxes. Such measurement could also be used to investigate channel bow of the nuclear fuel bundles surrounding the detector because of sensitivity of the thermal flux to geometry changes.Calculations have been performed with both lattice and nodal codes to study the behavior of the void fraction correlation to the ratio of the thermal- and fast-neutron fluxes. The results prove the correlation to be nearly linear and robust. The rate of change of the correlation is insensitive to standard reactor operating parameters such as control rods and burnable absorbers; the sensitivity of the ratio to void fraction changes primarily depends on the geometry of the fuel bundles. A linear prediction model was used to represent the nodal code results. The absolute void fraction at over 792 positions in the core could be predicted with an absolute uncertainty of ±1.5%.