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DOE OSTI · 23055119

Generating shield response functions using a source independent method - 25587

Lum, E. S.·Fensin, M. L.·Kelley, K. C.
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Abstract

Generalized questions about shield materials are common and yet difficult to answer analytically. Most forms of quick analysis use simple attenuation. The drawback to such methods is the use of mono-energetic averaged cross sections. Averaging a cross section requires knowledge of the flux a priori. If the flux was known, then some more in-depth calculation was performed negating the need for quick scoping calculations. Source independent methods for assessing the transport medium provide a means of creating response functions that yield general statements about the effectiveness of a shielding material. The following paper uses MCNP 6.2 to generate source independent, energy dependent, response functions for the transport medium that allow for quick scoping of effectiveness of various neutron shields. The source independent method here leveraged FT SCX tally treatment, to assess the performance of various shields such that energy response is preserved. (authors)

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61 RADIATION PROTECTION AND DOSIMETRY73 NUCLEAR PHYSICS AND RADIATION PHYSICS97 MATHEMATICS AND COMPUTINGATTENUATIONCROSS SECTIONSENERGY DEPENDENCENEUTRONSPERFORMANCERADIATION TRANSPORTRESPONSE FUNCTIONSSHIELDINGSHIELDING MATERIALS

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BibTeXRIS

Lum, E. S., Fensin, M. L., Kelley, K. C.. 2018-08-15. Generating shield response functions using a source independent method - 25587. https://www.osti.gov/biblio/23055119

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