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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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Latest News
First astatine-labeled compound shipped in the U.S.
The Department of Energy’s National Isotope Development Center (NIDC) on March 31 announced the successful long-distance shipment in the United States of a biologically active compound labeled with the medical radioisotope astatine-211 (At-211). Because previous shipments have included only the “bare” isotope, the NIDC has described the development as “unleashing medical innovation.”
Andrew E. Johnson, Dan Kotlyar
Nuclear Science and Engineering | Volume 194 | Number 2 | February 2020 | Pages 120-137
Technical Paper | doi.org/10.1080/00295639.2019.1661171
Articles are hosted by Taylor and Francis Online.
An adjoint-based method to predict the variation in spatial flux distribution during a depletion interval is presented in this paper. Burnup analyses require dividing a fuel cycle into multiple time intervals. At the start of each interval, the neutron transport equation is solved, and a subsequent depletion calculation is performed to obtain isotopic concentrations at the end of the interval. The most common approaches are to assume that either the flux or the power are constant through this depletion interval. In reality, changes in material compositions cause the flux and power distribution to change instantaneously, and thus, these assumptions are not valid in general except in the limit of infinitesimally small time steps. To overcome these assumptions, a method for predicting the spatial flux variation (SFV) due to changes in material compositions is derived, implemented, and verified. The formulation relies on the first-order perturbation formulation in conjunction with the forward and adjoint moments of the fission source, obtained from the fission matrix. Moreover, multiple adjoint modes are used to better predict the flux variation following materials transmutations. Such a prediction is capable of mimicking a transport calculation across a depletion interval based on the beginning-of-step transport solution and could be used to extend the simulated time between transport simulations in depletion and fuel cycle analysis. The SFV method is applied to a single three-dimensional fuel pin, depleted using a variety of depletion step sizes and verified against a reference simulation. The results show that the method produces accurate prediction of the end-of-step spatial flux distribution.