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Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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2024 ANS Annual Conference
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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.
Michael L. Fensin, John S. Hendricks, Samim Anghaie
Nuclear Technology | Volume 164 | Number 1 | October 2008 | Pages 3-12
Technical Paper | Icapp '06 | doi.org/10.13182/NT08-A4003
Articles are hosted by Taylor and Francis Online.
As advanced reactor concepts challenge the accuracy of current modeling technologies, a higher-fidelity depletion calculation is necessary to model time-dependent core reactivity properly for accurate cycle length and safety margin determinations. The recent integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a completely self-contained Monte Carlo-linked depletion capability. Two advances have been made in the latest MCNPX capability based on problems observed in prereleased versions: continuous-energy collision density tracking and adequate fission yield selection.Prereleased versions of the MCNPX depletion code calculated the reaction rates for (n,2n), (n,3n), (n,p), and (n,) by matching the MCNPX steady-state 63-group flux with 63-group cross sections inherent in the CINDER90 library and then collapsing to one-group collision densities for the depletion calculation. The accuracy of this procedure is therefore dictated by the adequacy of the 63-group energy structure of the cross-section set to accurately capture the spectrum of a specific model. Different types of models would therefore require different types of cross-section energy group structure. MCNPX 2.6.A eliminates this dependency by using the continuous-energy reaction rates determined during the MCNPX steady-state calculation to calculate energy-integrated collision rates to be used by CINDER90.MCNPX 2.6.A now also determines the proper fission yield to be used by the CINDER90 code for the depletion calculation. The CINDER90 code offers a thermal, fast, and high-energy fission yield for each fissile isotope contained in the CINDER90 data file. MCNPX 2.6.A determines which fission yield to use for a specified problem by calculating the integral fission rate for the defined energy boundaries (thermal, fast, and high energy), determining which energy range contains the majority of fissions, and then selecting the appropriate fission yield for the energy range containing the majority of fissions.The MCNPX depletion capability enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. This study focuses on the methodology development of the two improvements described here. Further improvements are under development to enhance the usefulness of this new capability.
