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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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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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Latest News
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.
Keisuke Kobayashi, Hiroshi Nishihara
Nuclear Science and Engineering | Volume 28 | Number 1 | April 1967 | Pages 93-104
Technical Paper | doi.org/10.13182/NSE67-A18671
Articles are hosted by Taylor and Francis Online.
The group-diffusion equation in one-dimensional geometry is solved by using Green's function. In the first section, using Green's tensor, the group-diffusion equation is transformed into a system of linear equations which contain only the fluxes at the interfaces between the regions. Solving this equation, we obtain the fluxes at the interfaces and then the flux inside the regions with the aid of Green's tensor. This treatment is the same kind of approach as that of the response matrix method or the theory of invariant imbedding. In the second section, the group-diffusion equation is solved by the source iteration method. Using Green's function, the exact three-point difference equation is obtained and the explicit forms for the slab, cylindrical, and spherical geometry are given. It is shown that the usual three-point difference equation is obtained if the source term is approximated to be flat piecewise and if Green's function is expanded into Taylor's series neglecting all but the first two terms. Sample calculations for a thermal and a fast reactor show that the improved difference equation obtained by approximating the source term by a polynomial of second degree is more accurate than the usual three-point difference equation.