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

Technical Report · Sun Nov 01 00:00:00 EST 1992

OSTI ID:10146373

Crawford, C; Palmer, B

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.

• Research Organization: Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC07-84ID12435
• OSTI ID: 10146373
• Report Number(s): WINCO-1112; ON: DE93012546; IN: CSS-92-015
• Resource Relation: Other Information: PBD: Nov 1992
• Country of Publication: United States
• Language: English

Related Subjects

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
MATHEMATICS AND COMPUTERS