Presentation made at the Electric Power Research Institute (EPRI) Extended Storage Collaboration Project (ESCP) meeting November 2016 discussing the status of nondestructive examinations being performed on high burnup (HBU) sent nuclear fuel (SNF) rods at Oak Ridge National Laboratory and proposed destructive examinations that will be performed over the next several years.

The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases.

The U.S. industry’s limited efforts at licensing transportation packages characterized as “highcapacity,”
or containing “high-burnup” (>45 GWd/MTU) commercial spent nuclear fuel
(CSNF), or both, have not been successful considering existing spent-fuel inventories that will
have to be eventually transported. A holistic framework is proposed for resolving several CSNF
transportation issues. The framework considers transportation risks, spent-fuel and cask-design

Expanding interest in nuclear power and advanced fuel cycles indicate that use of mixed-oxide (MOX) fuel in the current and new U.S. reactor fleet could become an option for utilities in the coming decades. In light of this renewed interest, EPRI has reviewed the substantial knowledge base on MOX fuel irradiation in light water reactors (LWRs). The goal was to evaluate the technical feasibility of MOX fuel use in the U.S. reactor fleet for both existing and advanced LWR designs (Generation III/III+).

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance.

The objective of this analysis is to evaluate accumulations within the thermally altered tuff surrounding a drift. The evaluation examines accumulation of uranium minerals (soddyite), plutonium oxide (Pu01), and combinations of these materials. A hypothetical model of the tuff is used to provide insight into the factors that affect criticality for this near-field scenario. The factors examined include: the size of the accumulation, the fissile composition of the accumulation, the water or clayey material fraction in the accumulation and the water fraction in the tuff

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013