Evaluation of shielding analysis methods in spent fuel cask environments. Final report

Technical Report · Mon May 01 04:00:00 EDT 1995

OSTI ID:95554

Broadhead, B L; Tang, J S; Childs, R L; Parks, C V; Taniuchi, H ^[1]

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1. Oak Ridge National Lab., TN (United States)

This report describes the application of the three-dimensional (3-D) Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, to the analysis of a series of simple geometry benchmark experiments and prototypic spent fuel storage cask measurements. The simple geometry experiments were performed in Japan and at the General Electric-Morris Operation facility; the cask measurements were performed at the Idaho National Engineering Laboratory (INEL). The quantification of uncertainties in a typical shielding analysis process for transport/storage casks can be accomplished by a comparison of consistent trends between calculated and measured dose-rate quantities in both benchmark and prototypic environments. Benchmark results typically measure the validity of cross-section data and computer code adequacy; prototypic environments generally measure the overall validity of the calculational procedure. Five storage cask problems and two simple geometry problems were analyzed to determine the expected accuracies of computational analyses using well-established source-generation and Monte Carlo codes.

• Research Organization: Electric Power Research Inst., Palo Alto, CA (United States); Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: Electric Power Research Inst., Palo Alto, CA (United States)
• OSTI ID: 95554
• Report Number(s): EPRI-TR--104329
• Country of Publication: United States
• Language: English

Related Subjects

05 NUCLEAR FUELS
42 ENGINEERING
BENCHMARKS
COMPUTER CALCULATIONS
CROSS SECTIONS
D CODES
FUEL ASSEMBLIES
GAMMA RADIATION
M CODES
MONTE CARLO METHOD
O CODES
RADIATION DOSES
RADIATION SOURCES
RECOMMENDATIONS
S CODES
SHIELDING
SOURCE TERMS
SPENT FUEL CASKS
SPENT FUEL STORAGE