The purpose of this calculation is to perform the same reactivity calculations as performed in Reference 7.1 and Reference 7.2 for a set of Laboratory Critical Experiments (LCE) except to change some of the cross section libraries as specified here, and to perform sixteen additional calculations for U233 LCEs.

The purpose of this engineering calculation is to document the MCNP4B2LVevaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories.

The purpose of this document is to present the methodology to be used for development of the Subcritical Limit (SL) for post closure conditions for the Yucca Mountain repository. The SL is a value based on a set of benchmark criticality multiplier, keff> results that are outputs of the MCNP calculation method. This SL accounts for calculational biases and associated uncertainties resulting from the use of MCNP as the method of assessing kerr·

The purpose of this study is to evaluate dissolved concentration limits (also referred to as solubility limits) of elements with radioactive isotopes under probable repository conditions, based on geochemical modeling calculations using geochemical modeling tools, thermodynamic databases, field measurements, and laboratory experiments.

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 document the MCNP4B2L V evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 182 different cases with varied design parameters. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (keff) for various critical configurations.

The purpose of this calculation is to perform a parametric study to determine the effects of fission product leaching, assembly collapse, and iron oxide loss on the reactivity of a waste package containing mixed oxide spent nuclear fuel. Previous calculations (CRWMS M&O 1998a) have shown that the criticality control features of the waste package are adequate to prevent criticality of a flooded WP for all the enrichment/burnup pairs expected for the MOX SNF.

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).

The purpose of this calculation file is to document criticality calculations performed on two different rod consolidation waste package designs. The results presented in this calculation file may be used to support further evaluation of the rod consolidation waste package design.

The purpose of this analysis is to document the Commercial Reactor Critical (CRC) benchmark evaluation performed for the Quad Cities Unit 1 boiling water reactor (BWR). The CRC benchmark is performed at a beginning of life (BOL) statepoint representing reactor start-up critical conditions. The objective of this CRC benchmark analysis is to provide a validation benchmark for the MCNP 4A analytic tool for use in the disposal criticality analysis methodology.

The purpose of this analysis is to document the MCNP evaluations of benchmark lattice Laboratory Critical Experiments (LCE's). The objective of this analysis is to quantify the MCNP 4A (Reference 5.4) code system's ability to accurately calculate the effective neutron multiplication factor (keff) for various measured critical (i.e., keff= 1.0) configurations. This analysis quantifies the effectiveness of the MCNP criticality calculation for lattice configurations containing U02 and Pu02 fissile oxide fuel using two different cross section data libraries.

The objective of this analysis is to evaluate accumulations within the thermally altered tuff surrounding a drift. The evaluation examines accumulation of uranium minerals (soddyite), plutonium oxide (Pu01), and combinations of these materials. A hypothetical model of the tuff is used to provide insight into the factors that affect criticality for this near-field scenario. The factors examined include: the size of the accumulation, the fissile composition of the accumulation, the water or clayey material fraction in the accumulation and the water fraction in the tuff

The Disposal Criticality Analysis Methodology Topical Report (Reference 1) states that the accuracy of the criticality analysis methodology (MCNP Monte Carlo code and cross-section data) designated to assess the potential for criticality of various configurations in the Yucca Mountain proposed repository is established by evaluating appropriately selected benchmark critical experiments.

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17x17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi- Purpose Canister (MPC) PWR waste package concepts.

This report, Summary Report of Laboratory Critical Experiment Analyses Performed for the Disposal Criticality Analysis Methodology, contains a summary of the laboratory critical experiment (LCE) analyses used to support the validation of the disposal criticality analysis methodology.

The purpose of this analysis is to document Waste Package Development Department (WPPD) MCNP evaluations of benchmark solution Laboratory Critical Experiments (LCE's). The objective of this analysis is to quantify the ability of the MCNP 4A (Reference 5.4) code system to accurately calculate the effective neutron multiplication factor (keff) for various measured critical (i.e., keff=1.0) configurations.

The purpose of this study is to evaluate dissolved concentration limits (also referred to as solubility limits) of elements with radioactive isotopes under probable repository conditions, based on geochemical modeling calculations using geochemical modeling tools, thermodynamic databases, field measurements, and laboratory experiments.

Thep purpose of this calculation is to perform degraded mode criticality evaluations of plutonium disposed in a ceramic waste form and emplaced in a Monitored geologic Repository (MGR). A 5 Defense High-Level Waste (DHLW) Canister Waste Package (WP) design, incorporating the can-in-canister concept for plutonium immobilization is considered in this calculation. Each HLW glass pour canister contains 7 tubes. Each tube contains 4 cans, with 20 ceramic disks (immobilized plutonium) in each.

The purpose of this document is to evaluate the potential for criticality for the fissile material that could accumulate in the near-field (invert) and in the far-field (host rock) beneath the U.S. Department of Energy (DOE) spent nuclear fuel (SNF) codisposal waste packages (WPs) as they degrade in the proposed monitored geologic repository at Yucca Mountain. The scope of this calculation is limited to the following DOE SNF types: Shippingport Pressurized Water Reactor (PWR), Enrico Fermi, Fast Flux Test Facility (FFTF), Fort St.

The objective of this calculation is to characterize the nuclear criticality safety concerns
associated with the codisposal of the Department of Energy’s (DOE) Enrico Fermi (EF) Spent
Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP) and
placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to
the determination of the effective neutron multiplication factor (keff) for the degraded mode
internal configurations of the codisposal WP. The results of this calculation and those of Ref. 8

This calculation file uses the MCNP neutron transport code to determine the range of parameters for Pressurized Water Reactor Spent Nuclear Fuel contained with a 21 PWR waste package (WP). Four base geometry patterns were considered in this work and included the following: intact fuel assemblies with intact WP internal components, intact fuel assemblies with degraded WP internal components, degraded fuel assemblies with intact WP internal components, and degraded fuel assemblies with degraded WP internal components.

The objective of this calculation is to document the Grand Gulf Unit 1 (GG1) reactivity calculations for sixteen critical statepoints in cycles 4 through 8. The GG1 reactor is a boiling water reactor (BWR) owned and operated by Entergy Operations Inc. The Commercial Reactor Criticality (CRC) evaluations support the development and validation of the neutronic models used for criticality analyses involving commercial spent nuclear fuel to be placed in a geologic repository. This calculation is performed as part of the evaluation in the CRC program.

The objective of this calculation is to evaluate the criticality potential for co-disposal waste packages affected by an igneous intrusion disruptive event in the emplacement drifts. The scope of this calculation is limited to U.S. Department of Energy (DOE) Spent Nuclear Fuel (SNF) types in DOE standardized SNF canisters or Multi-Canister Overpack (MCO) Canisters.

The purpose of this calculation is to perform nuclear criticality calculations for High-Level Waste (HLW) glass to support the criticality safety analysis of normal operations and off-normal conditions associated with the receipt, handling and loading of HLW glass canisters into 5-DHLW/DOE SNF Waste Packages (WPs) and 2-MCO/2-DHLW WPs in the surface facilities, in addition to the emplacement of loaded and sealed WPs in the sub-surface facility.