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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.
Seok-Hee Ryu, Kil-Sup Um, Jae-Il Lee
Nuclear Technology | Volume 189 | Number 2 | February 2015 | Pages 163-172
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT14-28
Articles are hosted by Taylor and Francis Online.
To evaluate the effect of thermal conductivity degradation for high-burnup fuel, a postulated control element assembly (CEA) ejection accident is assessed with the SPACE (Safety and Performance Analysis CodE) code. The SPACE code, which is currently under development as a safety analysis code for nuclear power plants, can predict thermal-hydraulic responses of the nuclear fuel and nuclear steam supply system during design basis accidents with two-fluid, three-field governing equations. Fuel performance behaviors during the CEA ejection accident using six fuel conductivity models including the burnup-independent reference conductivity model, the Lyons model, are investigated and compared with results of the reference model within the range from 0 to 30 GWd/tonne U. The Oak Ridge National Laboratory model predicts the highest peak fuel centerline temperature of 4531°F at 0 GWd/tonne U, and the modified Nuclear Fuels Institute model shows the uppermost value of 4796°F, which is 364°F higher than the reference model at 30 GWd/tonne U. It is also observed that the peak fuel centerline temperature increases linearly with fuel burnup and that the maximum increase rate of the peak centerline temperature per fuel burnup is ∼11.6°F per GWd/tonne U. For all thermal conductivity models, the maximum radial average fuel enthalpies are <230 cal/g, and the rise in radial average fuel enthalpy during the CEA ejection accident still remains within the pellet-cladding-mechanical-interaction failure criterion.