Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel

The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure
Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform
postclosure criticality calculations. The validation process applies the criticality analysis methodology
approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report.1
The application systems for this validation consist of waste packages containing transport, aging, and
disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types,
initial enrichments, and burnup values that are expected from the waste stream and of varying degree of
internal component degradation that may occur over the 10,000-year regulatory time period.2 The
criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP)3
transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the
various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI
(ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit
(LBTL) functions for CSNF waste packages are determined based on the guidance in
ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1.
The development of this report is consistent with Test Plan for: Range of Applicability and Bias
Determination for Postclosure Criticality.6 This calculation report has been developed in support of
licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the
calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual
TAD canister.