Visualizing MCNP Tally Segment Geometry and Coupling Results with ABAQUS

Conference · Thu Nov 01 00:00:00 EDT 2007

OSTI ID:912910

Parry, J R; Galbraith, J A

The Advanced Graphite Creep test, AGC-1, is planned for irradiation in the Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant program. The experiment requires very detailed neutronics and thermal hydraulics analyses to show compliance with programmatic and ATR safety requirements. The MCNP model used for the neutronics analysis required hundreds of tally regions to provide the desired detail. A method for visualizing the hundreds of tally region geometries and the tally region results in 3 dimensions has been created to support the AGC-1 irradiation. Additionally, a method was created which would allow ABAQUS to access the results directly for the thermal analysis of the AGC-1 experiment.

• Research Organization: Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: DOE - NE
• DOE Contract Number: DE-AC07-99ID-13727
• OSTI ID: 912910
• Report Number(s): INL/CON-07-12844; TRN: US0800609
• Resource Relation: Conference: 2007 ANS Winter Meeting,Washington, DC,11/11/2007,11/15/2007
• Country of Publication: United States
• Language: English

Related Subjects

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS , 99 - GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
COMPLIANCE
CREEP
DIMENSIONS
GEOMETRY
GRAPHITE
IRRADIATION
SAFETY
TEST REACTORS
THERMAL ANALYSIS
THERMAL HYDRAULICS
graphite
neutronics
NGNP
thermal-hydraulics