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Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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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.
P. Leconte, C. Vaglio-Gaudard, R. Eschbach, M. Antony, J. Di-Salvo, A. Pépino
Nuclear Science and Engineering | Volume 175 | Number 3 | November 2013 | Pages 308-317
Technical Paper | doi.org/10.13182/NSE12-56
Articles are hosted by Taylor and Francis Online.
The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a pressurized water reactor (PWR) between five and seven cycles, and also on the experimental validation of the spent fuel reactivity loss with burnup, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and the nuclear data responsible for the reactivity loss. This program also offers unique experimental data for fuels with a burnup reaching 85 GWd/tonne, as spent fuels in French PWRs have never exceeded 70 GWd/tonne up to now.The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first step, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists of the self-shielding of cross sections on the 281-energy-group SHEM mesh, followed by flux calculation by the method of characteristics in a two-dimensional exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between experiment and calculation shows satisfactory results with the JEFF3.1.1 library, which predicts the reactivity loss within 2% for burnup of ~75 GWd/tonne and within 4% for burnup of ~85 GWd/tonne.