Evaluation of shielding analysis methods in spent-fuel cask environments
- Oak Ridge National Lab., TN (United States). Computational Physics and Engineering Div.
The three-dimensional Monte Carlo code MORSE-SGC, as implemented in the SCALE system calculational sequence SAS4, is applied 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. A total of 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. The general trends seen in this work are in agreement within 30% or better with the measurements for neutron dose rates along the ask side, lid, and bottom. The gamma-ray dose rates with substantial contributions from the top endfitting, plenum, and bottom endfitting regions also are in good agreement. based on the latest results, gamma-ray dose rate calculations with major contributions due to the active fuel region show a consistent factor of 1.6 overprediction of the measured quantities for casks with iron and concrete shields. Major uncertainties exist in the quantification of {sup 59}Co concentrations in endfitting hardware materials. The results presented support the accuracy of source generation methods and dose estimation methods in these regions given accurate impurity characterizations. Thus, it is felt that the practice of using upper bounds for {sup 59}Co initial concentrations should ensure conservative cask designs.
- DOE Contract Number:
- AC05-96OR22464
- OSTI ID:
- 449567
- Journal Information:
- Nuclear Technology, Vol. 117, Issue 2; Other Information: PBD: Feb 1997
- Country of Publication:
- United States
- Language:
- English
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