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Radiation Protection Dosimetry 2005 115(1-4):117-121; doi:10.1093/rpd/nci260
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Published by Oxford University Press 2005.

SOURCES: a code for calculating ({alpha},n), spontaneous fission, and delayed neutron sources and spectra

W. B. Wilson1, R. T. Perry1, W. S. Charlton2, T. A. Parish2 and E. F. Shores1,*

1 Los Alamos National Laboratory, Los Alamos, NM 87545, USA
2 Nuclear Engineering Department, Texas A&M University, College Station, TX 77843, USA

* Corresponding author. eshores{at}lanl.gov

SOURCES is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission and delayed neutron emission owing to the decay of radionuclides in homogeneous media, interface problems and three-region interface problems. The code is also capable of calculating the neutron production rates due to ({alpha},n) reactions induced by a monoenergetic beam of alpha particles incident on a slab of target material. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the centre-of-mass system with a library of 107 nuclide decay alpha-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional alpha particle stopping cross sections for Z < 106. Spontaneous fission sources and spectra are calculated with evaluated half-life, spontaneous fission branching and Watt spectrum parameters for 44 actinides. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron sources. It also provides an analysis of the contributions to that source by each nuclide in the problem.


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