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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.
Degraded Waste Package Criticality: Summary Report of Evaluations Through 1996
Degraded Waste Package Criticality: Summary Report of Evaluations Through 1996
The purpose of this document is to summarize the degraded waste package disposal criticality evaluations which were performed in fiscal years I995 and I996. These evaluations were described in detail in 4 previous documents (Refs. I through 4). The initial version of this summary has been described in the I996 Disposal Criticality Analysis Methodology Technical Report (Ref. 5). A topical report planned for 1998 will present the methodology in its final form for approval by the US Nuclear Regulatory Commission.
slides - Generic Communications and Guidance on Spent Fuel Storage & Transportation
slides - Generic Communications and Guidance on Spent Fuel Storage & Transportation
Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013
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
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.
UCF Waste Package Criticality Analysis
UCF Waste Package Criticality Analysis
This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the UCF waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan5·1 for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives.
Preliminary Criticality Analysis of Degraded SNF Accumulations External to a Waste Package (SCPB: N/A)
Preliminary Criticality Analysis of Degraded SNF Accumulations External to a Waste Package (SCPB: N/A)
This study is prepared by the Mined Geologic Disposal System (MODS) Waste Package Development Department (WPDD) to provide input to a separate evaluation on the probablility of criticality in the far- field environment. These calculations are performed in sufficient detail to provide conservatively bounding configurations to support separate probabilistic analyses.
EQ6 Calculation for Chemical Degradation of Enrico Fermi Codisposal Waste Packages: Effects of Updated Design and Rates
EQ6 Calculation for Chemical Degradation of Enrico Fermi Codisposal Waste Packages: Effects of Updated Design and Rates
The Monitored Geologic Repository (MGR) Waste Package Project of the BSC Management and Operating Contractor for the Department of Energy's Office of Civilian Radioactive Waste Management performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Enrico Fermi Reactor owned by the DOE (Ref. 9). The Fermi SNF has been considered for disposal at the proposed Yucca Mountain site.
Analysis of Dust Deliquescence for FEP Screening
Analysis of Dust Deliquescence for FEP Screening
The purpose of this report is to evaluate the potential for penetration of the Alloy 22 (UNS N06022) waste package outer barrier by localized corrosion due to the deliquescence of soluble constituents in dust present on waste package surfaces. The results support a recommendation to exclude deliquescence-induced localized corrosion (pitting or crevice corrosion) of the outer barrier from the total system performance assessment for the license application (TSPA-LA).
Isotopic Models for Commercial SNF Burnup Credit
Isotopic Models for Commercial SNF Burnup Credit
Disposal Criticality Analysis Methodology Topical Report1 describes a methodology for performing postclosure criticality analyses within the repository at Yucca Mountain, Nevada. An important component of the postclosure criticality analysis is the calculation of conservative isotopic concentrations for spent nuclear fuel. This report documents the isotopic calculation methodology. The isotopic calculation methodology is shown to be conservative based upon current data for pressurized water reactor and boiling water reactor spent nuclear fuel.
U.S. Regulatory Recommendations for Actinide-Only Burnup Credit in Transport and Storage Casks
U.S. Regulatory Recommendations for Actinide-Only Burnup Credit in Transport and Storage Casks
In July 1999, the U.S. Nuclear Regulatory Commission (NRC) Spent Fuel Project Office
(SFPO) issued Interim Staff Guidance 8 Revision 1 (ISG8R1) to provide recommendations for the use
of burnup credit in storage and transport of pressurized-water reactor (PWR) spent fuel. Subsequent to
the issuance of ISG8R1, the NRC Office of Regulatory Research (RES) has directed an effort to
investigate the technical basis for extending the criteria and recommendations of ISG8R1 to allow
Drift Scale THM Model
Drift Scale THM Model
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.
Criticality Evaluation of Degraded Internal Configurations for the PWR AUCF WP Designs
Criticality Evaluation of Degraded Internal Configurations for the PWR AUCF WP Designs
The purpose of this analysis is to provide input on the criticality potential of various degraded configurations to an analysis on the probability of a criticality event in a Pressurized Water Reactor (PWR) Advanced Uncanistered Fuel (AUCF) Waste Package (WP).
Preclosure Criticality Analysis Process Report
Preclosure Criticality Analysis Process Report
The preclosure criticality analysis process described in this technical report provides a systematic approach for determining the need for criticality controls and for evaluating their effectiveness during the preclosure period of the Monitored Geologic Repository at Yucca Mountain, Nevada.
General Corrosion and Localized Corrosion of Waste Package Outer Barrier
General Corrosion and Localized Corrosion of Waste Package Outer Barrier
The purpose and scope of this model report is to document models for general and localized corrosion of the waste package outer barrier (WPOB) to be used in evaluating long-term waste package performance in the total system performance assessment (TSPA). The waste package design for the license application is a double-wall waste package placed underneath a protective drip shield (SNL 2007 [DIRS 179394]; SNL 2007 [DIRS 179354]). The WPOB will be constructed of Alloy 22 (UNS N06022) (SNL 2007 [DIRS 179567], Section 4.1.1.6), a highly corrosion-resistant nickel-based alloy.
44-BWR WASTE PACKAGE LOADING CURVE EVALUATION
44-BWR WASTE PACKAGE LOADING CURVE EVALUATION
The objective of this calculation is to evaluate the required minimum burnup as a function of initial boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an application of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU.
Waste Package Probabilistic Criticality Analysis: Summary Report of Evaluations in 1997
Waste Package Probabilistic Criticality Analysis: Summary Report of Evaluations in 1997
The purpose of this document is to summarize the degraded waste package disposal criticality evaluations which were reported in FY 1997 (Refs. 2-6), and to explain how those evaluations have served to further develop various aspects of the overall methodology for such evaluations.
Probabilistic Criticality Consequence Evaluation (SCPB: N/A)
Probabilistic Criticality Consequence Evaluation (SCPB: N/A)
This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department with the objective of providing a comprehensive, conservative estimate of the consequences of the criticality which could possibly occur as the result of commercial spent nuclear fuel emplaced in the underground repository at Yucca Mountain. The consequences of criticality are measured principally in terms of the resulting changes in radionuclide inventory as a function of the power level and duration of the criticality.
Criticality Analysis of Pu and U Accumulations in a Tuff Fracture Network
Criticality Analysis of Pu and U Accumulations in a Tuff Fracture Network
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
Multiscale Thermohydrologic Model
Multiscale Thermohydrologic Model
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.
Commercial Spent Nuclear Fuel Waste Package Misload Analysis
Commercial Spent Nuclear Fuel Waste Package Misload Analysis
The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design.
Analysis of Critical Benchmark Experiments for Configurations External to WP
Analysis of Critical Benchmark Experiments for Configurations External to WP
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.
Long-Term Criticality Control Issues for the MPC (SCPB: N/A)
Long-Term Criticality Control Issues for the MPC (SCPB: N/A)
This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) in response to a request received via a QAP-3-12 Design Input Data Request (Reference 5.1) from Waste Acceptance, Storage, & Transportation (WAST) Design (formerly MRSMPC Design). This design analysis is an answer to the Design Input Data Request to provide: Specific requirements for long-term criticality control.
End Effect keff Cias Curve for Actinide-Only Burnup Credit Casks
End Effect keff Cias Curve for Actinide-Only Burnup Credit Casks
A conservative end effect keff bias curve for actinide-only burnup credit casks is presented
in this paper. Rather than performing axially burnup-dependent analysis, cask designers can, if
they choose to, analyze casks with a uniform axial burnup (at assembly average burnup value) and
add the keff bias values to conservatively bound the actinide-only end effect. Earlier studies
suggested 1-3% increase in keff to account for the end effect, but they included fission products
as well as actinides for their analyses.
Calculation of Isotopic Bias and Uncertainty for BWR SNF
Calculation of Isotopic Bias and Uncertainty for BWR SNF
The objective of Calculation of Isotopic Bias and Uncertainty for BWR SNF is to quantify the computational bias and uncertainty in the multiplication factor (keff) to be used for Boiling Water Reactor (BWR) spent nuclear fuel (SNF) burn-up credit. The scope of this bias and uncertainty determination covers 38 different radiochemical assay (RCA) spent fuel samples from 14 different fuel assemblies that were irradiated in four different BWRs. The irradiated fuel samples evaluated span an enrichment range of 2.53 weight percent U-235 through 3.95 weight percent U-235.