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Title: Validation of MCNP: SPERT-D and BORAX-V fuel

This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assemblies or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.
Authors:
;
Publication Date:
OSTI Identifier:
10146373
Report Number(s):
WINCO--1112
ON: DE93012546; IN: CSS-92-015
DOE Contract Number:
AC07-84ID12435
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: Nov 1992
Research Org:
Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States)
Sponsoring Org:
USDOE, Washington, DC (United States)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE; FUEL ELEMENTS; TESTING; FUEL ASSEMBLIES; M CODES; VALIDATION; FUEL PLATES; SLABS; REACTOR LATTICES; URANIUM ALLOYS; ALUMINIUM ALLOYS; MODERATORS; COMPUTER CALCULATIONS 220300; 990200; FUEL ELEMENTS; MATHEMATICS AND COMPUTERS