SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary

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 is to be taken for
the reduced reactivity of burned or spent fuel relative to its original $fresh# composition, it is
necessary to benchmark computational methods used in determining such reactivity worth against
spent fuel reactivity measurements. 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 (PWR).
The analysis methodology utilized for all calculations in this report is based on the modules
and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in
the core were calculated using the SAS2H analytical sequence in SCALE-4. The sources of data and
the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code
sequence was used to extract the necessary isotopic densities from SAS2H results and to provide the
data in the format required for SCALE-4 criticality analysis modules. The CSASN analytical sequence
in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module
of SCALE-4 was used to calculate the effective multiplication factor (k ) for the critical eff
configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and
ENDF/B-V (fission products) data was used for analysis of each critical configuration.
Each of the five volumes comprising this report provides an overview of the methodology
applied. Subsequent volumes also describe in detail the approach taken in performing criticality
calculations for these PWR configurations: Volume 2 describes criticality calculations for the
Tennessee Valley Authority’s Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis
of Virginia Power’s Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations
performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly,
Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the
reactor-specific volumes provides the details of calculations performed to determine the effective
multiplication factor for each reactor core for one or more critical configurations using the SCALE-4
system; these results are summarized in this volume. Differences between the core designs and their
possible impact on the criticality calculations are also discussed. Finally, results are presented for
additional analyses performed to verify that solutions were sufficiently converged. All calculations
show the ability to predict a k value very close to 1.0 for various conditions and cooling times. eff
Thus, the methodology applied is shown to be a valid approach for calculating the value of k for eff
systems with spent PWR fuel.