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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 Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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
Proving DRACO will deliver
The United States is now closer than it has been in over five decades to launching the first nuclear thermal rocket into space, thanks to DRACO—the Demonstration Rocket for Agile Cislunar Orbit.
Tsung-Kuang Yeh, Digby D. Macdonald, Arthur T. Motta
Nuclear Science and Engineering | Volume 121 | Number 3 | December 1995 | Pages 468-482
Technical Paper | doi.org/10.13182/NSE95-A24148
Articles are hosted by Taylor and Francis Online.
A computer code with the capability of simultaneously estimating the concentrations of radiolysis species, the electrochemical corrosion potential, and the kinetics of growth of a reference crack in sensitized Type 304 stainless steel is developed for the heat transport circuits of boiling water reactors (BWRs). The primary objective of this code, DAMAGE-PREDICTOR, is to theoretically evaluate the effectiveness of hydrogen water chemistry (HWC) in the BWRs as a function of feedwater hydrogen concentration and reactor power level. The power level determines various important thermal-hydraulic parameters and the neutron and gamma energy deposition rate in the core and near-core regions. These input parameters are estimated using well-established algorithms, and the simulations are carried out for full-power conditions for two reactors that differ markedly in their responses to HWC. The DAMAGE-PREDICTOR code is found to successfully account for plant data from both reactors using a single set of model parameter values.