Taking credit for the reduced reactivity of spent nuclear fuel in criticality analyses is referred to
as burnup credit. Criticality safety evaluations employing burnup credit require validation of the
depletion and criticality calculation methods and computer codes with available measurement
data. To address the issues of burnup credit criticality validation, the U.S. Nuclear Regulatory
Commission initiated a project with Oak Ridge National Laboratory to (1) develop and establish

The purpose of this calculation is to perform waste-form specific nuclear criticality safety calculations to aid in establishing criticality safety design criteria, and to identify design and process parameters that are potentially important to the criticality safety of Department of Energy (DOE) standardized Spent Nuclear Fuel (SNF) canisters.

This paper provides insights into the neutronic similarities between a representative high-capacity rail-transport cask containing typical pressurized water reactor (PWR) spent nuclear fuel assemblies and critical reactor state-points, referred to as commercial reactor critical (CRC) state-points. Forty CRC state-points from five PWRs were analyzed, and the characteristics of CRC state-points that may be applicable for validation of burnup-credit criticality safety calculations for spent fuel transport/storage/disposal systems were identified.

The purpose of this study is to provide insights into the neutronic similarities that may exist between a
generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-
points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the
type of CRC state-points that may be applicable for validation of burnup credit criticality safety
calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-

The waste forms under consideration for disposal in the repository at Yucca Mountain contain scores of radionuclides. It would be impractical and highly inefficient to model all of these radionuclides in a total system performance assessment (TSPA). Thus, the purpose of this radionuclide screening analysis is to remove from further consideration (screen out) radionuclides that are unlikely to significantly contribute to radiation dose to the public from a nuclear waste repository at Yucca Mountain.

Spent fuel transportation and storage cask designs based on a burnup credit approach must
consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For
example, the spent fuel composition must be adequately characterized and the criticality analysis
model can be complicated by the need to consider axial burnup variations. Parametric analyses are
needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel

The purpose of this calculation is to demonstrate that the preclosure performance objectives specified in 10 CFR 63.111(a) and 10 CFR 63.111(b) (Reference 2.2.1) have been met for the proposed design and operations in the geologic repository operations area (GROA) during normal operations and Category 1 event sequences, and following Category 2 event sequences. Category 1 event sequences are those natural and human-induced event sequences that are expected to occur one or more times before permanent closure of the repository.

This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC (Bechtel SAIC Company) 2004 (DIRS 167614).

Taking credit for the reduced reactivity of spent nuclear fuel (SNF) in criticality analyses is referred to as burnup credit (BUC). Criticality safety evaluations require validation of the computational methods with critical experiments that are as similar as possible to the safety analysis models, and for which the keff values are known. This poses a challenge for validation of BUC criticality analyses, as critical experiments with actinide and fission product (FP)

The purpose of this calculation is to perform waste-form specific nuclear criticality safety calculations to aid in establishing criticality safety design criteria, and to identify design and process parameters that are potentially important to the criticality safety of the transportation, aging and disposal (TAD) canister-based systems.