Preliminary Validation Studies of Existing Neutronics Analysis Tools for Advanced Burner Reactor Design Applications

Technical Report · Mon Apr 30 04:00:00 EDT 2007

DOI:https://doi.org/10.2172/1508296· OSTI ID:1508296

Yang, W. S. ^[1]; Kim, T. K. ^[1]; Kim, S. J. ^[1]; Lee, C. H. ^[1]

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1. Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

Validation studies of ANL neutronics analysis tools and nuclear data for Advanced Burner Reactor design applications have been initiated. As an initial effort, a numerical benchmark problem based on the reference metal fuel core concept of 250 MW_t Advanced Burner Test Reactor (ABTR) and the six ZPPR-21 critical assemblies were analyzed. Preliminary tests of the ENDF/B-VII.0 data were also performed using a set of fast system criticality benchmark problems. Multi-group cross sections were generated using the ETOE-2/MC²-2 codes based on the ENDF/B-V, ENDF/B-VI, and ENDF/B-VII nuclear data. Deterministic core calculations were performed with the VARIANT nodal transport, TWODANT discrete ordinate transport, and DIF3D diffusion theory codes. MCNP5 simulations for the ABTR numerical benchmark problem showed a significant dependence of Monte Carlo solutions on the base cross section libraries. The core multiplication factors determined with the ENDF/B-VI and VII data were different by almost 1.0 %Δk. Even different signs were observed among the sodium void worths obtained with different ENDF/B versions, although the sodium void worth of ABTR was estimated to be less than 1$.

• Research Organization: Argonne National Laboratory (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: AC02-06CH11357
• OSTI ID: 1508296
• Report Number(s): ANL-AFCI--186; 148747
• Country of Publication: United States
• Language: English

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