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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.
Tom Burr, Jeremy Conlin, Jianwei Hu, Jack Galloway, Vladimir Henzl, Howard Menlove, Martyn Swinhoe, Stephen Tobin, Holly Trellue, Timothy Ulrich
Nuclear Science and Engineering | Volume 172 | Number 2 | October 2012 | Pages 180-192
Technical Paper | doi.org/10.13182/NSE11-73
Articles are hosted by Taylor and Francis Online.
Estimating plutonium (Pu) mass in spent nuclear fuel assemblies (SFAs) helps inspectors ensure that no Pu is diverted. Therefore, nondestructive assay (NDA) methods are being developed to assay Pu mass in SFAs. Uncertainty quantification is an important task in most assay methods, and particularly for SFA assay. A computer model (MCNPX) is being used to predict isotope masses and the spatial distribution of masses in virtual SFAs for 64 combinations of initial fuel enrichment (IE), fuel utilization [burnup (BU)], and cooling time (CT) values. Additional MCNPX modeling for the same 64 virtual SFAs provided the expected detector responses (DRs) for several NDA techniques such as the passive neutron albedo reactivity method and the 252Cf interrogation with prompt neutrons method.A previous paper describes one uncertainty quantification approach involving Monte Carlo (MC) simulation using individually any of six new NDA options together with IE, BU, and CT. This paper provides an interpretation of the MC approach that is suited for a numerical Bayesian alternative, separately assesses the impact of MCNPX interpolation error, and compares several options to use subsets of IE, BU, CT, and one DR.
