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Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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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.”
John E. Suich, Henry C. Honeck
Nuclear Science and Engineering | Volume 20 | Number 1 | September 1964 | Pages 93-110
Technical Paper | doi.org/10.13182/NSE64-A19279
Articles are hosted by Taylor and Francis Online.
A method is developed for calculating the temperature coefficient of ηf for heterogeneous reactor lattice cells on a fairly rigorous basis, using only microscopic material constants as input data. The method is based on the integral transport equation, and involves flux and adjoint weighting the temperatures derivatives of the kernels of the integral operators. Temperature coefficients are obtained for a localized temperature increase, as well as for a uniform increase in cell temperature. The coefficients are separated, on physical grounds, into ‘spectrum’ and ‘transport’ effects. The numerical accuracy of the method is found to be limited, at the present time, by the uncertainties in fuel reaction cross sections. The method is used in a brief survey of temperature effects in natural-uranium/graphite lattices. The transport temperature coefficients are shown to yield the dependence of the thermal multiplication factor on a velocity-averaged diffusion coefficient. The spectrum temperature coefficients give the dependence of the thermal multiplication factor on average neutron velocity and disadvantage factor. Non-diffusion effects are noticed when the region near the fuel is heated. The results of the method are compared with published experimental results for natural-uranium/graphite lattices. Good agreement between theory and experiment is obtained. The influence of reactor operating conditions on temperature coefficients is reproduced by the theory.