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Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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2024 ANS Winter Conference and Expo
November 17–21, 2024
Orlando, FL|Renaissance Orlando at SeaWorld
Standards Program
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.
C. K. Cheng, B. M. Ma
Nuclear Science and Engineering | Volume 48 | Number 2 | June 1972 | Pages 139-158
Technical Paper | doi.org/10.13182/NSE72-A22467
Articles are hosted by Taylor and Francis Online.
The time-dependent radius of the central void and the extents of the columnar grain, the equiaxed grain, and the unaffected grain regions of a typical oxide cylindrical fuel rod in a fast reactor at constant power level are determined. The temperature distributions in the fuel element are obtained. A model postulated to analyze and calculate the irradiation swelling and fission-gas release for oxide fuels of fast reactors is developed. The mechanical analysis is based on the thermal and radiation dilatations and on an elastoplastic approach for the Prandtl-Reuss material. An iteration method of successive approximation is used to compute the stresses and strains developed in the fuel elements. The computed results are shown by curves for the unsteady-state fuel restructuring of the fuel element.