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Waste Package Probabilistic Criticality Analysis: Summary Report of Evaluations in 1997
Waste Package Probabilistic Criticality Analysis: Summary Report of Evaluations in 1997
The emplacement of nuclear waste in the proposed geologic repository must satisfy relevant regulatory requirements with respect to criticality, 10CFR60. I31 (h) (Ref. 25). The waste packages for the various waste forms will be designed to preclude criticality (typically by the inclusion of neutron absorbers) even if the waste package becomes filled with water. Criticality may, however, be possible if the contents of the waste package become degraded in such a way that the fissile material can be separated from the neutron absorbers, while sufficient moderator is retained.
Frequency of SNF Misload for Uncanistered Fuel Waste Package
Frequency of SNF Misload for Uncanistered Fuel Waste Package
The purpose ofthis engineering calculation is to estimate the frequency of misloading spent nuclear fuel (SNF) assemblies that would result in exceeding the criticality design basis of a waste package (WP). This type of misload - a reactivity misload - results from the incorrect placement of one or more fuel assemblies into a waste package such that the criticality controls do not match the required controls for the fuel assemblies.
External Criticality Risk of Immobilized Plutonium Waste Form in a Geologic Repository
External Criticality Risk of Immobilized Plutonium Waste Form in a Geologic Repository
This technical report provides an updated summary of the waste package (WP) external criticalityrelated
risk of the plutonium disposition ceramic waste form, which is being developed and
evaluated by the Office of Fissile Materials Disposition of the U.S. Department of Energy (DOE).
The ceramic waste form consists of Pu immobilized in ceramic disks, which would be embedded
in High-Level Waste (HLW) glass in the HLW glass disposal canisters, known as the "can-incanister"
THE REPORT TO THE PRESIDENT AND THE CONGRESS BY THE SECRETARY OF ENERGY ON THE NEED FOR A SECOND REPOSITORY
THE REPORT TO THE PRESIDENT AND THE CONGRESS BY THE SECRETARY OF ENERGY ON THE NEED FOR A SECOND REPOSITORY
The Nuclear Waste Policy Act of 1982, as amended (NWPA), establishes a process for the siting, construction and operation of one or more national repositories for permanent disposal of the Nation’s spent nuclear fuel (SNF) and high-level radioactive waste (HLW). In 1987, after the Department of Energy (the Department or DOE) had conducted studies of nine potential repository sites located throughout the United States, Congress amended the NWPA and selected the Yucca Mountain site in Nye County, Nevada as the only site for further study for the first national repository.
Second Waste Package Probabilistic Criticality Analysis: Generation and Evaluation of Internal Criticality Configurations
Second Waste Package Probabilistic Criticality Analysis: Generation and Evaluation of Internal Criticality Configurations
This analysis is prepared by the Mined Geologic Disposal System (MODS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having some or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment.
Dry Transfer Facility Criticality Safety Calculations
Dry Transfer Facility Criticality Safety Calculations
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.
Nuclear Criticality Calculations for Canister-Based Facilities - DOE SNF
Nuclear Criticality Calculations for Canister-Based Facilities - DOE SNF
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. It is intended that the results of the criticality safety calculations provided in this document will
Evaluation of Codisposal Viability for Aluminum-Clad DOE-Owned Spent Fuel: Phase I Intact Codisposal Canister
Evaluation of Codisposal Viability for Aluminum-Clad DOE-Owned Spent Fuel: Phase I Intact Codisposal Canister
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.
EQ6 Calculations for Chemical Degradation of TRIGA Codisposal Waste PacKages
EQ6 Calculations for Chemical Degradation of TRIGA Codisposal Waste PacKages
The Monitored Geologic Repository Waste Package Operations 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 Training, Research, Isotopes, General Atomics (TRIGA) reactor (Ref. 1). The TRIGA SNF has been considered for disposal at the potential Yucca Mountain site.
Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package
Intact and Degraded Mode Criticality Calculations for the Codisposal of TMI-2 Spent Nuclear Fuel in a Waste Package
The objective of these calculations is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) Three Mile Island- Unit 2 (TMI-2) spent nuclear fuel (SNF) in canisters. This analysis evaluates codisposal in a 5-Defense High-Level Waste (5-DHLW/DOE SNF) Long Waste Package (Civilian Radioactive Waste Management System Management and Operating Contractor [CRWMS M&O] 2000b, Attachment V), which is to be placed in a potential monitored geologic repository (MGR).
EQ6 Calculation for Chemical Degradation of Shippingport LWBR (Th/U Oxide) Spent Nuclear Fuel Waste Packages
EQ6 Calculation for Chemical Degradation of Shippingport LWBR (Th/U Oxide) Spent Nuclear Fuel Waste Packages
The Monitored Geologic Repository (MGR) Waste Package Department 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 Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site.
Second Waste Package Probabilistic Criticality Analysis: Generation and Evaluation of Internal Criticality Configurations
Second Waste Package Probabilistic Criticality Analysis: Generation and Evaluation of Internal Criticality Configurations
This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having some or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment.
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.
Criticality Consequence Calculation Involving Intact PWR MOX SNF in a Degraded 21 PWR Assembly Waste Package
Criticality Consequence Calculation Involving Intact PWR MOX SNF in a Degraded 21 PWR Assembly Waste Package
Data Analysis for Infiltration Modeling: Extracted Weather Station Data Used to Represent Present-Day and Potential Future Climate Conditions in the Vicinity of Yucca Mountain
Data Analysis for Infiltration Modeling: Extracted Weather Station Data Used to Represent Present-Day and Potential Future Climate Conditions in the Vicinity of Yucca Mountain
The purpose of this analysis is to identify, extract, and reformat weather (meteorological) data that is appropriate for use as input to an infiltration model, within the Yucca Mountain region. The analysis uses relevant meteorological data (e.g., precipitation and temperature) from source stations, and reformats or converts the data into a form suitable for the generation of meteorological conditions for a 10,000-year future climate in the Yucca Mountain region.
General Corrosion and Localized Corrosion of the Drip Shield
General Corrosion and Localized Corrosion of the Drip Shield
The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall.
The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared ·according to Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]).
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.
Summary Report of Commercial Reactor Criticality Data for McGuire Unit 1
Summary Report of Commercial Reactor Criticality Data for McGuire Unit 1
The "Summary Report of Commercial Reactor Criticality Data for McGuire Unit 1" contains the detailed information necessary to perform commercial reactor criticality (CRC) analyses for the McGuire Unit 1 reactor.
Summary Report of Laboratory Critical Experiment Analyses Performed for the Disposal Criticality Analysis Methodology
Summary Report of Laboratory Critical Experiment Analyses Performed for the Disposal Criticality Analysis Methodology
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.
Bias and Range of Applicability Determinations for Commercial Nuclear Fuels
Bias and Range of Applicability Determinations for Commercial Nuclear Fuels
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.
Commercial Reactor Reactivity Analysis for Grand Gulf, Unit 1
Commercial Reactor Reactivity Analysis for Grand Gulf, Unit 1
The objective of this calculation is to document the Grand Gulf Unit 1 (GGl) 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.
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.