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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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2024 ANS Winter Conference and Expo
November 17–21, 2024
Orlando, FL|Renaissance Orlando at SeaWorld
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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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New laws offer nuclear industry incentives for existing power plant uprates
This year, the U.S. nuclear industry received a much-needed economic boost that could help preserve operating nuclear power plants and incentivize upgrades that extend their lifespan and power output.
Signed into law in 2022, the Inflation Reduction Act offers production tax credits (PTCs) for existing nuclear power plants and either PTCs or investment tax credits (ITCs) for new carbon-free generation. These credits could make power uprates—increasing the maximum power level at which a commercial plant may operate—a much more appealing option for utilities.
P. Benoist, M. Carta, G. Palmiotti, M. Salvatores, J. Tullett
Nuclear Science and Engineering | Volume 103 | Number 3 | November 1989 | Pages 254-264
Technical Paper | doi.org/10.13182/NSE89-A23676
Articles are hosted by Taylor and Francis Online.
A simple method to calculate the effectiveness of the control assembly in a fast neutron reactor is proposed. For each type of heterogeneous assembly (control or follower), a polar parameter, taking into account the assembly absorption and the axial leakage of neutrons inside the assembly, is defined. In a similar way, a bipolar parameter, taking into account the reaction of the assembly to a transverse flux gradient, is also defined. These two parameters, deduced from transport theory, are used to determine the absorption cross section and the diffusion coefficient of an equivalent homogeneous control or follower assembly. These new parameters are introduced in a one-group diffusion code, calculating the reactor as a whole with any number of control and follower assemblies. An approximate generalization to multigroup theory is proposed. Numerical comparisons show that this equivalent diffusion method gives results that are much closer to transport results than those obtained by the classical diffusion theory.