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PWR Axial Burnup Profile Analysis
PWR Axial Burnup Profile Analysis
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.
Westinghouse 17x17 MOX PWR Assembly- Waste Package Criticality Analysis (SCPB: N/A)
Westinghouse 17x17 MOX PWR Assembly- Waste Package Criticality Analysis (SCPB: N/A)
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.
Probability of a PWR Uncanistered Fuel Waste Package Postclosure Criticality
Probability of a PWR Uncanistered Fuel Waste Package Postclosure Criticality
The purpose of this calculation is to estimate the probability of criticality in a pressurized water reactor (PWR) uncanistered fuel waste package during the postclosure phase of the repository as a function of various waste package material, loading, and environmental parameters. Parameterization on the upper subcritical limit that is used to define the threshold for criticality will also be performed. The possibility of waste package misload due to human or equipment error during preclosure is also considered in estimating the postclosure criticality probability.
Radionuclide Screening
Radionuclide Screening
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.
Disposal Criticality Analysis for Aluminum-based Fuel in a Codisposal Waste Package - ORR and MIT SNF - Phase II
Disposal Criticality Analysis for Aluminum-based Fuel in a Codisposal Waste Package - ORR and MIT SNF - Phase II
The objective of this analysis is to characterize the criticality safety aspects of a degraded Department of Energy spent nuclear fuel (DOESNF) canister containing Masachusetts Institute of Technology (MIT) or Oak Ridge Research (ORR) fuel in the Five Pack defense high level waste (DHLW) waste package to demonstrate concept viability related to use in the Minded Geologic Disposal System (MGDS) environment for the postclosure time frame.
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations
Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations
The objective of the Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations is to determine the accuracy of the SAS2H control module of the baselined modular code system SCALE, Version 4.4A (STN: 10129-4.4A-00), in predicting the isotopic concentrations of spent fuel, and to quantify the overall effect that the differences between the calculated and measured isotopic concentrations have on the system reactivity. The scope of this calculation covers eight different spent fuel samples from a fuel assembly that was irradiated in the Limerick Unit 1 boiling water reactor (BWR).
MCNP Evaluation of Laboratory Critical Experiments: Homogeneous Mixture Criticals
MCNP Evaluation of Laboratory Critical Experiments: Homogeneous Mixture Criticals
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.
Slides - Retrievability, Cladding Integrity, and Safety Handling during Storage and Transportation
Slides - Retrievability, Cladding Integrity, and Safety Handling during Storage and Transportation
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
Nuclear Criticality Calculations for the Wet Handling Facility
Nuclear Criticality Calculations for the Wet Handling Facility
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)
Range of Parameters For PWR SNF in a 21 PWR WP
Range of Parameters For PWR SNF in a 21 PWR WP
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.
Waste Packages and Source Terms for the Commercial 1999 Design Basis Waste Streams
Waste Packages and Source Terms for the Commercial 1999 Design Basis Waste Streams
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.
Criticality Evaluation of Plutonium Disposition Ceramic Waste Form: Degraded Mode
Criticality Evaluation of Plutonium Disposition Ceramic Waste Form: Degraded Mode
The 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 for this calculation. Each HLW glass pour canister contains 7 tubes. Each tube contains 4 cans, with 20 ceramic disks (immobilized plutonium) in each.
Criticality Safety and Shielding Evaluations of the Codisposal Canister in the Five-Pack DHLW Waste Package
Criticality Safety and Shielding Evaluations of the Codisposal Canister in the Five-Pack DHLW Waste Package
The objective of this analysis is to characterize a codisposal canister containing MIT or ORR fuel in the Five-Pack defense high level waste (DHLW) waste package (WP) to demonstrate concept viability related to use in the Mined Geologic Disposal System (MGDS) environment for the postclosure time frame. The purpose of this analysis is to investigate the disposal criticality and shielding issues for the DHLW WP and establish DHLW WP and codisposal canister compatibility with the MGDS, and to provide criticality and shielding evaluations for the preliminary DHLW WP design.
Westinghouse MOX SNF Isotopic Source
Westinghouse MOX SNF Isotopic Source
The purpose of this calculation is to develop an estimate of the isotopic content as a function of time for mixed oxide (MOX) spent nuclear fuel (SNF) assemblies in a Westinghouse pressurized water reactor (PWR). These data will be used as source data for criticality, thermal, and radiation shielding evaluations of waste package (WP) designs for MOX assemblies in the Monitored Geologic Repository (MGR).
Disposal and Storage of Spent Nuclear Fuel — Finding the Right Balance
Disposal and Storage of Spent Nuclear Fuel — Finding the Right Balance
The Nuclear Waste Policy Act of 1982, as amended, established a statutory basis
for managing the nation’s civilian (or commercially produced) spent nuclear
fuel. The law established a process for siting, developing, licensing, and constructing
an underground repository for the permanent disposal of that waste.
Utilities were given the primary responsibility for storing spent fuel until it is
accepted by the Department of Energy (DOE) for disposal at a repository —
originally expected to begin operating in 1998. Since then, however, the repository
EQ6 Calculations for Chemical Degradation of Enrico Fermi Spent Nuclear Fuel Waste Packages
EQ6 Calculations for Chemical Degradation of Enrico Fermi Spent Nuclear Fuel Waste Packages
The Monitored Geologic Repository (MGR) Waste Package Operations (WPO) of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Emico Fermi Atomic Power Plant (Ref. 1). The Fermi fuel has been considered for disposal at the potential Yucca Mountain site.
January 2013 Presentation to the Institute fo Nuclear Materials Management on Near Term Planning for Storage and Transportation of Used Nuclear Fuel
January 2013 Presentation to the Institute fo Nuclear Materials Management on Near Term Planning for Storage and Transportation of Used Nuclear Fuel
This is the Nuclear Fuels Storage and Transportation Project Director's presentation on Near Term Planning for Stroage and Transportation of Used Nuclear Fuel presented to the Institute of Nuclear Materials Management on January 14, 2013 in Arlington Va.
Fast Flux Test Facility (FFTF) Reactor Fuel Degraded Criticality Calculations: Intact SNF Canister
Fast Flux Test Facility (FFTF) Reactor Fuel Degraded Criticality Calculations: Intact SNF Canister
The purpose of these calculations is to characterize the criticality safety concerns for the storage of Fast Flux Test Facility (FFTF) nuclear fuel in a Department of Energy spent nuclear fuel (DOE SNF) canister in a co-disposal waste package. These results will be used to support the analysis that will be done to demonstrate concept viability related to use in the Monitored Geologic Repository (MGR) environment.
Summary Report of SNF Isotopic Comparisons for the Disposal Criticality Analysis Methodology
Summary Report of SNF Isotopic Comparisons for the Disposal Criticality Analysis Methodology
The "Summary Report of SNF Isotopic Comparisons for the Disposal Criticality Analysis Methodology" contains a summary of the analyses that compare SNF measured isotopic concentrations (radiochemical assays) to calculated SNF isotop~c concentrations (SAS2H module ·orScale4.3). The results of these analyses are used to support the validation of the isotopic models for spent commercial light water reactor (LWR) fuel.
Criticality Potential of Waste Packages Affected by Igneous Intrusion
Criticality Potential of Waste Packages Affected by Igneous Intrusion
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.
UFD Storage and Transportation - Transportation Working Group Report
UFD Storage and Transportation - Transportation Working Group Report
The Used Fuel Disposition (UFD) Transportation Task commenced in October 2010. As its first task, Pacific Northwest National Laboratory (PNNL) compiled a list of structures, systems, and components (SSCs) of transportation systems and their possible degradation mechanisms during extended storage. The list of SSCs and the associated degradation mechanisms [known as features, events, and processes (FEPs)] were based on the list of used nuclear fuel (UNF) storage system SSCs and degradation mechanisms developed by the UFD Storage Task (Hanson et al. 2011).
Evaluation of Internal Criticality of the Plutonium Disposition MOX SNF Waste Form
Evaluation of Internal Criticality of the Plutonium Disposition MOX SNF Waste Form
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 (Me203) on the reactivity of a waste package (WP) containing mixed oxide (MOX) spent nuclear fuel (SNF). Previous calculations (CRWMS M&O 1998a) have shown that the criticality control features of the WP are adequate to prevent criticality of a flooded WP for all the enrichment/ burnup pairs expected for the MOX SNF.
Configuration Model Generator
Configuration Model Generator
The Disposal Criticality Analysis Methodology Topical Reporta prescribes an approach to the methodology for performing postclosure criticality analyses within the monitored geologic repository at Yucca Mountain, Nevada. An essential component of the methodology is the Configuration Generator Model for In-Package Criticality that provides a tool to evaluate the probabilities of degraded configurations achieving a critical state.
Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel
Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel
The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package.