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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.
Ian C. Gauld, Georgeta Radulescu, Germina Ilas, Brian D. Murphy, Mark L. Williams, Dorothea Wiarda
Nuclear Technology | Volume 174 | Number 2 | May 2011 | Pages 169-195
Technical Paper | Special Issue on the SCALE Nuclear Analysis Code System / Fuel Cycle and Management | doi.org/10.13182/NT11-3
Articles are hosted by Taylor and Francis Online.
The calculation of fuel isotopic compositions is essential to support design, safety analysis, and licensing of many components of the nuclear fuel cycle - from reactor physics and severe accident analysis to back-end fuel cycle issues, including spent-fuel storage and transportation, reprocessing, and radioactive waste management. Versions of the ORIGEN code, developed by Oak Ridge National Laboratory, have been used worldwide for isotopic depletion and decay analysis for more than three decades. The supported version of ORIGEN, maintained as the depletion analysis module for SCALE 6, performs detailed time-dependent isotopic generation and depletion for 1946 nuclides for reactor fuel and activation analysis. Stand-alone ORIGEN calculations can be performed using cross-section libraries developed for a wide range of reactor types and fuel designs used worldwide, including light water reactors UO2 and MOX, CANDU, VVER 440 and 1000, RBMK, and graphite reactors. Alternatively, within SCALE 6, ORIGEN can be automatically coupled to two-dimensional discrete ordinates or three-dimensional Monte Carlo transport solvers that provide problem-dependent cross sections for use in the ORIGEN depletion calculation. The hybrid ability to function as either a stand-alone or coupled depletion code provides ORIGEN advanced capabilities to simulate a broad range of applications for various reactor systems. The nuclear data libraries in ORIGEN have been significantly improved recently, using modern ENDF/B nuclear data evaluations. The most recent developments in SCALE 6.1 include the addition of ENDF/B-VII decay data, energy-dependent fission yields, and fine-group ORIGEN neutron cross sections based on the JEFF-3.0/A special purpose activation files. Advanced methods and data for neutron and gamma source energy spectral analysis are also available in the current version of the code. The ORIGEN code and associated nuclear data libraries have been extensively validated against experimental data that include spent nuclear fuel isotopic assay data for actinides and fission products, radiation source spectra, and decay heat measurements.