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Members are devoted to applying nuclear science and engineering technologies involving isotopes, radiation applications, and associated equipment in scientific research, development, and industrial processes. Their interests lie primarily in education, industrial uses, biology, medicine, and health physics. Division committees include Analytical Applications of Isotopes and Radiation, Biology and Medicine, Radiation Applications, Radiation Sources and Detection, and Thermal Power Sources.
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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.
Akitoshi Hotta, Hiroshi Shirai, Shinya Mizokami
Nuclear Science and Engineering | Volume 152 | Number 3 | March 2006 | Pages 292-305
Technical Paper | doi.org/10.13182/NSE06-A2583
Articles are hosted by Taylor and Francis Online.
A postulated single control rod drop transient was calculated for a typical boiling water reactor plant taking into account effects of detailed void distributions in those bundles neighboring the withdrawn control blade. Time-dependent pin power distributions were reconstructed by the plant simulator TRAC/BF1-ENTRÉE and were exported to the subchannel code NASCA.Macroscopic cross-section libraries based on flat and distorted void distributions were allocated in accordance with fuel location in a simplified two-way coupling method. Exposure trends of bundle neutronic properties were compared between two void distributions. Although the infinite multiplication factor was not influenced, the radial peaking factor increased significantly because of the void distortion caused by pin-by-pin exposure of fissile materials.The result with the combined cross sections was compared with those with the flat void cross sections. Application of the combined cross sections lowered the initial local peaking because of larger neutron leakage around the withdrawn control blade. The transient linear power density at the critical fuel rod increased more rapidly. A change in the fuel heat flux was attenuated because of the heat conduction delay. As a consequence of these effects, the peak cladding temperature became slightly lower than that of the flat void model.