The purpose of this calculation is to apply the process described in the TDR-DS0-NU-000001 Rev. 02, Preclosure Criticality Analysis Process Report (Ref. 2.2.25) to aid in establishing design and operational criteria important to criticality safety and to identify potential control parameters and their limits important to the criticality safety of commercial spent nuclear fuel (CSNF) handling operations in the Wet Handling Facility (WHF)

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.

The purpose of this calculation is to apply the process described in the Preclosure Criticality Analysis Process Report (Ref. 2.2.12) to establish the bias for keff calculations performed for commercial nuclear fuels using the MCNP code system. This bias will be used in criticality safety analyses as part of the basis for establishing the upper subcritical limit (USL). This calculation also defines the range of applicability (ROA) for which the bias may be used directly without need to consider additional penalties on the USL.

The means to prevent and control criticality must be addressed as part of the Preclosure Safety Analysis (PCSA) required for compliance with 10 CFR Part 63 [DIRS 180319], where the preclosure period covers the time prior to permanent closure activities. This technical report presents the nuclear criticality safety evaluation that documents the achievement of this objective.

This report proposes and documents a computational benchmark for the estimation of the
additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor
actinides in a burnup-credit storage/transport environment, relative to SNF compositions
containing only the major actinides. The benchmark problem/configuration is a generic burnupcredit
cask designed to hold 68 boiling water reactor (BWR) spent nuclear fuel assemblies. The
purpose of this computational benchmark is to provide a reference configuration for the

This report has been prepared to support technical discussion of and planning for future
research supporting implementation of burnup credit for boiling-water reactor (BWR) spent fuel
storage in spent fuel pools and storage and transport cask applications. The review and
discussion in this report are based on knowledge and experience gained from work performed
in the United States and other countries, including experience with burnup credit for
pressurized-water reactor (PWR) spent fuel. Relevant physics and analysis phenomena are

This report presents a best-estimate probabilistic risk assessment (PRA) to quantify the frequency of criticality accidents during railroad transportation of spent nuclear fuel casks. The assessment is of sufficient detail to enable full scrutiny of the model logic and the basis for each quantitative parameter contributing to criticality accident scenario frequencies. The report takes into account the results of a 2007 peer review of the initial version of this probabilistic risk assessment, which was published as EPRI Technical Report 1013449 in December 2006.

Utilization of burnup credit in criticality safety analysis for long-term disposal of spent
nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile
material that will be present in the repository. Burnup-credit calculations are based on depletion
calculations that provide a conservative estimate of spent fuel contents (in terms of criticality
potential), followed by criticality calculations to assess the value of the effective neutron

This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance.

The purpose of this calculation is to establish the relative change in the effective neutron multiplication factor (keff) due to the use of MCNP unique identifiers (ZAIDs) in Nuclear Criticality Calculations for Canister-Based Facilities - DOE SNF (Reference 2.2.1, Attachment 3, MCNP inputs.zip) that are different to the ZAIDs used in the Analysis of Critical Benchmark Experiments and Critical Limit Calculation for DOE SNF (Reference 2.2.5, Table 5-3).

This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the Project Design Criteria (PDC) Document (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in Project Requirements Document (Canori and Leitner 2003 [DIRS 166275], p.

The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the Disposal Criticality Analysis Methodology Topical Report (YMP 2000).

The purpose of this calculation is to evaluate the transient behavior and consequences of a worst- case criticality event involving intact pressurized water reactor (PWR) mixed-oxide (MOX) spent nuclear fuel (SNF) in a degraded basket configuration inside a 21 PWR waste package (WP). This calculation will provide information necessary for demonstrating that the consequences of a worst-case criticality event involving intact PWR MOX SNF are insignificant in their effect on the overall radioisotopic inventory and on the integrity of the repository.

The purpose of this activity is to develop a representative “limiting” axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the “end-effect”. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package.

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).

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.

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.

This paper presents an evaluation of the amount of burnup credit needed for high-density casks to
transport the current U.S. inventory of commercial spent nuclear fuel (SNF) assemblies. A prototypic
32-assembly cask and the current regulatory guidance were used as bases for this evaluation.
By comparing actual pressurized-water-reactor (PWR) discharge data (i.e., fuel burnup and initial
enrichment specifications for fuel assemblies discharged from U.S. PWRs) with actinide-only-based

The purpose of this calculation is to characterize the criticality aspect of the codisposal of TRIGA (Training, Research, Isotopes, General Atomic) reactor spent nuclear fuel (SNF) with Savannah River Site (SRS) high-level waste (HLW). The TRIGA SNF is loaded into a Department of Energy (DOE) standardized SNF canister which is centrally positioned inside five-canister defense SRS HLW waste package (WP). The objective of the calculation is to investigate the criticality issues for the WP containing the five SRS HLW and DOE SNF canisters in various stages of degradation.

The purpose of this analysis is to evaluate the transient behavior and consequences of a worst case criticality event involving intact pressurized water reactor (PWR) spent nuclear fuel (SNF) in a degraded basket configuration inside a 21 PWR assembly waste package (WP). The objective of this analysis is to demonstrate that the consequences of a worst case criticality event involving intact PWR SNF are insignificant in their effect on the overall radioisotopic inventory in a WP. An internal WP criticality is modeled in a manner analogous to transient phenomena in a nuclear reactor core.