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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
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Las Vegas, NV|Mandalay Bay Resort and Casino
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Glass strategy: Hanford’s enhanced waste glass program
The mission of the Department of Energy’s Office of River Protection (ORP) is to complete the safe cleanup of waste resulting from decades of nuclear weapons development. One of the most technologically challenging responsibilities is the safe disposition of approximately 56 million gallons of radioactive waste historically stored in 177 tanks at the Hanford Site in Washington state.
ORP has a clear incentive to reduce the overall mission duration and cost. One pathway is to develop and deploy innovative technical solutions that can advance baseline flow sheets toward higher efficiency operations while reducing identified risks without compromising safety. Vitrification is the baseline process that will convert both high-level and low-level radioactive waste at Hanford into a stable glass waste form for long-term storage and disposal.
Although vitrification is a mature technology, there are key areas where technology can further reduce operational risks, advance baseline processes to maximize waste throughput, and provide the underpinning to enhance operational flexibility; all steps in reducing mission duration and cost.
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.