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Title: Visualizing MCNP Tally Segment Geometry and Coupling Results with ABAQUS

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