Unirradiated reactor fuel has a well-specified nuclide composition that provides a
straightforward and bounding approach to the criticality safety analysis of transport and storage
casks. As the fuel is irradiated in the reactor, the nuclide composition changes and, ignoring
the presence of burnable poisons, this composition change will cause the reactivity of the fuel to
decrease. Allowance in the criticality safety analysis for the decrease in fuel reactivity resulting

Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2 - Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport
and Storage Casks

Subcritical source-driven noise measurements are simultaneous Rossi-a and randomly
pulsed neutron measurements that provide measured quantities that can be related to the
subcritical neutron multiplication factor. In fact, subcritical source-driven noise
measurements should be performed iii lieu of Rossi-a rneasurements because of the
additional information that is obtained from noise measurements such as the spectral ratio
and the coherencc functions. The basic understanding of source-driven noisc analysis

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
criticality safety analyses be validated against experimental measurements. If credit for the negative
reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark
computational methods against spent fuel critical configurations. This report summarizes a portion
of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical
configurations from commercial pressurized-water reactors.

Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been
modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system.
The SAS2H sequence uses transport methods combined with the depletion and decay capabilities
of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup
history. Results of these calculations are compared with chemical assay measurements of spent fuel
inventories for each sample. Results show reasonable agreement between measured and predicted

This paper presents an assessment of the benefits for extended burnup credit in transporting
pressurized-water-reactor (PWR) spent nuclear fuel (SNF) in the United States. A prototypic 32-
assembly cask and the current regulatory guidance were used as bases for this assessment. By
comparing recently released PWR discharge data with actinide-only-based loading curves, this
evaluation shows that additional negative reactivity (through either increased credit for fuel burnup or

Disposal Criticality Analysis Methodology Topical Report describes a methodology for performing postclosure criticality analyses within the repository at Yucca Mountain, Nevada. An important component of the postclosure criticality analysis is the calculation of conservative isotopic concentrations for spent nuclear fuel. This report documents the isotopic calculation methodology. The isotopic calculation methodology is shown to be conservative based upon current data for pressurized water reactor and boiling water reactor spent nuclear fuel.

Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor
have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code
system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the
Waste Package Design team at the Yucca Mountain Project in the United States, who performed
the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical
measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The

The attached documents are an extensive list of references relevant to burnup credit criticality analysis. Some of the references may be available within the CURIE document collection.

This report documents the work performed by ORNL for the Yucca Mountain Project (YMP)
M&O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k values for inf
infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various
burnup/enrichment and cooling-time combinations. These scenarios simulate expected limiting
criticality loading conditions (for a given assembly type) for drift emplacements in a repository. Upon