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Extra info for Waste Package Design Methodology Rpt
Example text
Out of each group, a representative fuel type is chosen as a bounding case for the group. Generally, burnup, fissile enrichments, and total fuel mass determine the selection of the representative fuel used for the criticality analysis. For shielding calculations, the bounding photon and neutron source terms for these criticality analyses are also used. Currently, DOE has completed six of the nine criticality analyses and thus has identified six bounding shielding source terms for the six fuel groups.
Macroscopic cross sections have units of reciprocal length (cm-I), and they act as linear attenuation coefficients in the photon transport. Shielding analyses require macroscopic cross sections in order to determine interactions of radiation with the materials of the system. The codes calculate the macroscopic cross sections based on material compositions and densities, and, in the case of a Monte Carlo radiation transport code, they select the target element. The chemical compositions and densities of the materials in a waste package shielding calculation are available in the American Society for Testing and Materials (ASTM) standards database.
Damage is expressed in terms of diffusion-controlled cavity growth of micro-flaws in the zircaloy matrix. 3 Shielding and Source Term Codes and Standards c Shielding analysis methodologies are discussed in the SRPs for storage and transportation. Shielding analysis methodologies are often validated along with criticality safety software systems. Specific standards include the following: NUREG/CR-6484 (Broadhead et al. 1996), which describes verification and validation of the shielding sequences of the SCALE analysis systems NUREG/CR-5625 (Hermann et al.