Calculation of neutron spectra for a {sup 252}Cf transport cask using ANISN running on a PC
Journal Article
·
· Transactions of the American Nuclear Society
OSTI ID:411745
- Univ. of Arkansas, Fayetteville, AR (United States)
Neutron spectra have been calculated using the ANISN one-dimensional discrete ordinates code for the case of a {sup 152}Cf source in a transport cask of a particular design. All computations were done on personal computers (PCs) (mostly 486 models) with the ANISN-ORNL (486 version) computer code. With a source of {sup 252}Cf fission neutrons, the neutron flux spectrum in the cask cannot be characterized as {open_quotes}moderated.{close_quotes} Concern about an appropriate choice for the cross-section data set has led to a comparison, for this application, of three different cross-section libraries: DABL, HILO, and BUGLE-80. Although the cross-section sets were not originally designed for PC use, the libraries have been successfully employed for PC computations. Work with yet another data library, BUGLE-93, is incomplete at this stage. From neutron flux spectra on the surface of the cask, personnel dosimetric quantities (such as dose equivalent) have been determined for the DABL, HILO, and BUGLE-80 ANISN calculations.
- OSTI ID:
- 411745
- Report Number(s):
- CONF-951006--
- Journal Information:
- Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 73; ISSN TANSAO; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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