SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5

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 relevant and well-documented critical
configurations from commercial pressurized water reactors.
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 SCALE-4 SAS2H analytical sequence. The sources of data and
the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code family
was used to extract the necessary isotopic densities from SAS2H results and to provide the data in
the format required for SCALE 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 (keff) for the critical configuration.
The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission
products) data was used for all calculations.
This volume of the report documents a reactor critical calculation for GPU Nuclear
Corporation’s Three Mile Island Unit 1 (TMI-1) during hot, zero-power startup testing for the
beginning of cycle 5. This unit and cycle were selected because of their relevance in spent fuel
benchmark applications: (1) cycle 5 startup occurred after an especially long downtime of 6.6 years;
and (2) the core consisted primarily (75%) of burned fuel, with all fresh fuel loaded on the core outer
periphery. A keff value of 0.9978 ± 0.0004 was obtained using two million neutron histories in the
KENO V.a model. This result is close to the known critical keff of 1.0 for the actual core and is
consistent with other mixed-oxide criticality benchmarks. Thus this method is shown to be valid for
spent fuel applications in burnup credit analyses.