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)

This model report documents the abstraction of drift seepage, conducted to provide seepage relevant parameters and their probability distributions for use in Total System Performance Assessment for License Application (TSPA-LA). Drift seepage refers to the flow of liquid water into waste emplacement drifts.

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 purpose of this calculation is to simulate the processing of an incoming waste stream into waste packages, simulating the required aging as applicable, and the emplacement of the waste packages into the Yucca Mountain repository.

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.

This model report documents the drift scale coupled thermal-hydrological-mechanical (THM) processes model development and presents simulations of the THM behavior in fractured rock close to emplacement drifts. The modeling and analyses are used to evaluate the impact of THM processes on permeability and flow in the near-field of the emplacement drifts.

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

The purpose of this report is to demonstrate that postclosure temperature limits can be met, and certain thermal characteristics of the postclosure thermal reference case can be preserved, with alternative thermal loading schemes. The analysis considers certain variations from the base case.waste stream, the predicted postclosure temperatures that develop within the rock mass due to these waste stream variations, and then compares these temperatures to postclosure temperature limits.

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.

This calculation is prepared by the Monitored Geologic Repository Waste Package Requirements & Integration Department. The purpose of this calculation is to compile source term and commercial waste stream information for use in the analysis of waste package (WP) designs for commercial fuel. Information presented will consist of the number of WPs, source terms, metric tons of uranium, and the average characteristics of assemblies to be placed in each WP design. The source terms provide thermal output, radiation sources, and radionuclide inventories.

The objective of this calculation is to evaluate the peak temperatures due to thermal loading and boundary conditions of the TAD Waste Package design under nominal Monitored Geologic Repository conditions.

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 report was prepared in accordance with Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]). This report provides information on the phase stability of Alloy 221, the current waste package outer barrier material. The goal of this model is to determine whether the single-phase solid solution is stable under repository conditions and, if not, how fast other phases may precipitate.

This study provides results supporting the conclusion that the repository can be operated over a varying range of thermal modes and therefore temperatures. In particular, this work focused on limiting the peak, postclosure waste package surface temperature to less than 85 degrees Celsius, a possible limit due to corrosion considerations. These operating modes were compared by varying the waste package in drift spacing (0.1-2.83 meters), drift pitch (drift spacing centerline to centerline of 40-120 meters), ventilation duration (75-300 years), and ventilation efficiency (50-80%).

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 intended purpose of the multiscale thermohydrologic model (MSTHM) is to predict the possible range of thermal-hydrologic conditions, resulting from uncertainty and variability, in the repository emplacement drifts, including the invert, and in the adjoining host rock for the repository at Yucca Mountain. The goal of the MSTHM is to predict a reasonable range of possible thermal-hydrologic conditions within the emplacement drift.

This evaluation is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide analyses of disposal of aluminum (AI)-based Department of Energy-owned research reactor spent nuclear fuel (DOE-SNF) in a codisposal waste package with five canisters of high-level waste (HLW). The analysis was performed in sufficient detail to establish the technical viability of the Al-based DOE-SNF codisposal canister option.