SOURCES-3A: A code for calculating ({alpha}, n), spontaneous fission, and delayed neutron sources and spectra

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In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay {alpha}-particles in ({alpha},n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO{sub 2}, ThO{sub 2}, MOX, etc.), enrichment plant operations (UF{sub 6}, PuF{sub 4}, etc.), waste tank studies, waste products in ... continued below

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11 p.

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Perry, R.T.; Wilson, W.B. & Charlton, W.S. April 1, 1998.

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In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay {alpha}-particles in ({alpha},n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO{sub 2}, ThO{sub 2}, MOX, etc.), enrichment plant operations (UF{sub 6}, PuF{sub 4}, etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of {alpha}-emitting source material and low-Z target material) and in interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material). 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. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 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. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an analysis of the contributions to that source by each nuclide in the problem.

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11 p.

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INIS; OSTI as DE98005278

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  • Radiation protection and shielding topical meeting: technologies for the new century, Nashville, TN (United States), 19-23 Apr 1998

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  • Other: DE98005278
  • Report No.: LA-UR--98-87
  • Report No.: CONF-980403--
  • Grant Number: W-7405-ENG-36
  • Office of Scientific & Technical Information Report Number: 658395
  • Archival Resource Key: ark:/67531/metadc711596

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  • April 1, 1998

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  • Sept. 12, 2015, 6:31 a.m.

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  • Feb. 26, 2016, 5:59 p.m.

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Perry, R.T.; Wilson, W.B. & Charlton, W.S. SOURCES-3A: A code for calculating ({alpha}, n), spontaneous fission, and delayed neutron sources and spectra, article, April 1, 1998; New Mexico. (digital.library.unt.edu/ark:/67531/metadc711596/: accessed August 23, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.