Publications

Abstract not provided.

Abstract not provided.

The Nuclear Regulatory Commission (NRC) approves new package designs for shipping fissile quantities of UF{sub 6}. Currently there are three packages approved by the NRC for domestic shipments of fissile quantities of UF{sub 6}: NCI-21PF-1; UX-30; and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR Part 71. The primary objective of this project was to relate the conditions experienced by these packages in the tests described in 10 CFR Part 71 to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR Part 71 tests was achieved by means of computer modeling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from test and other fire scenarios. In addition, the likelihood of encountering bodies of water or sufficient rainfall to cause complete or partial immersion during transport over representative truck routes was assessed. Modeled effects, and their associated probabilities, were combined with existing event-tree data, plus accident rates and other characteristics gathered from representative routes, to derive generalized probabilities of encountering accident conditions comparable to the 10 CFR Part 71 conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents, i.e. the likelihood of UF{sub 6} being dispersed as a result of accident impact or fire is small. Moreover, given that an accident has occurred, exposure to water by fire-fighting, heavy rain or submersion in a body of water is even less probable by factors ranging from 0.5 to 8E-6.

A study of the errors introduced when one-dimensional inverse heat conduction techniques are applied to problems involving two-dimensional heat transfer effects was performed. The geometry used for the study was a cylinder with similar dimensions as a typical container used for the transportation of radioactive materials. The finite element analysis code MSC P/Thermal was used to generate synthetic test data that was then used as input for an inverse heat conduction code. Four different problems were considered including one with uniform flux around the outer surface of the cylinder and three with non-uniform flux applied over 360{degree}, 180{degree}, and 90{degree} sections of the outer surface of the cylinder. The Sandia One-Dimensional Direct and Inverse Thermal (SODDIT) code was used to estimate the surface heat flux of all four cases. The error analysis was performed by comparing the results from SODDIT and the heat flux calculated based on the temperature results obtained from P/Thermal. Results showed an increase in error of the surface heat flux estimates as the applied heat became more localized. For the uniform case, SODDIT provided heat flux estimates with a maximum error of 0.5% whereas for the non-uniform cases, the maximum errors were found to be about 3%, 7%, and 18% for the 360{degree}, 180{degree}, and 90{degree} cases, respectively.