Review of Experimental Data for Validating Computer Codes Used in Shielding Calculations for Spent Fuel Storage and Transportation Systems

Abstract

It should be noted that a relatively small subset of the identified experimental data (e.g., criticality alarm experiments) is available in a standard format established by the international community participating in experimental isotopic and shielding data evaluations. An effort of the SFCOMPO Technical Review Group (TRG) is underway to publish first isotopic evaluations of individual assay data using a standard data evaluation format. The SINBAD TRG has recently initiated benchmark evaluations and modernization of the database. Therefore, more relevant information is expected in the future that will enable users to select quality experimental data in depletion code and shielding code validations for SNF applications.

Updated Recommendations Related to Spent Fuel Transport and Dry Storage Shielding Analyses

“…The analyzed average fuel assembly burnup and maximum initial UO2 fuel enrichment values were 80 GWd/MTU and 12%, respectively. ORNL/SPR-2022/2518 [17] provides a summary of available experimental data for validating computer codes used in source term and shielding calculations for spent fuel storage and transportation systems. ORNL/SPR-2022/2692 [18] provides a review of SCALE validations applicable to SNF shielding calculations.…”

“…At long cooling times (100 years or longer), 238 Pu, 239 Pu, 240 Pu, and 242 Pu, 241 Am, and 246 Cm are primary neutron source contributors through SF and/or (α,n) reactions. In UO2 fuel, 17 O(α,n) 20 Ne and 18 O(α,n) 21 Ne reactions are the principal source of neutrons from (α,n) reactions because of the large amount of oxygen compared with other light elements that might be present in the fuel in trace quantities as impurities. However, nuclides with larger cross sections for neutron production might be present as impurities in the fuel oxide.…”

“…Light nuclei-such as 6,7 Li, 9 Be, 10,11 B, 12,13 C, 14,15 N, and 17,18 O-have large cross sections for neutron production [48,49]. NUREG/CR-6802 [9] indicates that for situations in which the (α,n) contribution to the total neutron source is significant, it is necessary to accurately quantify or establish bounding values for all impurities, especially light elements that are known to produce significant (α,n) quantities.…”

“…Thus, the neutron source depends on the actinides present in the waste and the resin elemental composition. A significant neutron source can be produced from 17 O (α,n) and 18 O (α,n) reactions, as well as (α,n) reactions with light-element impurities in the waste matrix material. Actinide neutron and gamma source terms were calculated in ORNL/TM-2020/1586 [14] based on spent resin actinide loading data provided in NUREG/CR-2830 [88] and DOE/EIS-0375 [90].…”

“…This report provides updated recommendations relevant to current shielding designs and analyses, which are primarily based on a series of ORNL research studies sponsored by the NRC. These research studies are documented in sponsor reports [13,14,15,16,17,18] that are publicly available from the DOE Office of Scientific and Technical Information (OSTI.GOV). More justification for updated shielding recommendations is provided in Section 1.1.…”