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SAS2H Generated Isotopic Concentrations for B&W 15xl5 PWR Assembly (SCPB: N/A)
SAS2H Generated Isotopic Concentrations for B&W 15xl5 PWR Assembly (SCPB: N/A)
This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.
Initial Radionuclide Inventories
Initial Radionuclide Inventories
The purpose of this analysis is to provide an initial radionuclide inventory (in grams per waste package) and associated uncertainty distributions for use in the Total System Performance Assessment for the License Application (TSPA-LA) in support of the license application for the repository at Yucca Mountain, Nevada. This document is intended for use in postclosure analysis only.
Probabilistic Criticality Consequence Evaluation
Probabilistic Criticality Consequence Evaluation
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.
Drift-Scale THC Seepage Model
Drift-Scale THC Seepage Model
The purpose of this report is to document the thermal-hydrologic-chemical (THC) seepage model and model simulations. The simulations predict the composition of fracture water that could potentially seep into repository emplacement drifts and the composition of the associated gas phase. The THC seepage model is not used to feed the total system performance assessment (TSPA) for the license application (LA).
THC Sensitivity Study of Heterogeneous Permeability and Capillarity Effects
THC Sensitivity Study of Heterogeneous Permeability and Capillarity Effects
The purpose of this report is to <,locument the sensitivity of the drift-scale thermal-hydrologic- chemical (THC) seepage model (SNL 2007 [DIRS 177404]) to heterogeneities in permeability and capillarity, which could affect predicted fluxes and chemistries of water and gases seeping into the emplacement drifts. This report has been developed following Technical Work Plan for: Revision of Model Reports for Near-Field and In-Drift Water Chemistry (SNL 2007 [DIRS 179287]).
Nuclear Fuel Data Survey, Form GC-859, Rev. May 2013
Nuclear Fuel Data Survey, Form GC-859, Rev. May 2013
This is the May 2013 revision to Form GC-859, Nuclear Fuel Data Survey, downloaded from the EIA site on September 12, 2014.
This form requires license holders to provide the data defined within the form for spent nuclear fuel as of June 30, 2013. This survey is similar to the older RW-859 surveys.
Axial Burnup Profile Database for Pressurized Water Reactors
Axial Burnup Profile Database for Pressurized Water Reactors
The data were obtained directly from utilities whose reactors represent the range of commercial PWR fuel lattices. The work was performed by Yankee Atomic Electric for Sandia National Laboratory. All axial burnup profiles were calculated from 3-D depletion analyses of the core configuration. The organizations and utilities providing axial burnup profiles for the database used different model codes for the 3D-depletion calculations. The model codes used were: SIMULATE-3, NEMO, ANC, and PRESTO-II. Cross-section inputs describing the assemblies are derived from assembly lattice calculations.
Summary of 2017 Public Comments on the Draft Consent-Based Siting Process for Consolidated Storage and Disposal Facilities for Spent Nuclear Fuel and High-Level Radioactive Waste
Summary of 2017 Public Comments on the Draft Consent-Based Siting Process for Consolidated Storage and Disposal Facilities for Spent Nuclear Fuel and High-Level Radioactive Waste
This document summarizes comments received on the U.S. Department of Energy’s (DOE) Draft Consent-Based Siting Process for Consolidated Storage and Disposal Facilities for Spent Nuclear Fuel and High-Level Radioactive Waste (referred to throughout this document as the Draft Consent-Based Siting Process). DOE published a request for public comment on the Draft Consent-Based Siting Process in the Federal Register on January 13, 2017 (82 FR 4333).
Consent-Based Siting Process for Federal Consolidated Interim Storage of Spent Nuclear Fuel
Consent-Based Siting Process for Federal Consolidated Interim Storage of Spent Nuclear Fuel
In this document, the U.S. Department of Energy (DOE) outlines its road map for implementing a consent-based siting process to site one or more federal consolidated interim storage facilities for spent nuclear fuel.
Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste 3
Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste 3
The Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste is a framework for moving toward a sustainable program to deploy an integrated system capable of transporting, storing, and disposing of used nuclear fuel1 and high-level radioactive waste from civilian nuclear power generation, defense, national security and other activities. The Strategy addresses several important needs.
Designing a Consent-Based Siting Process: Summary of Public Input, Final Report
Designing a Consent-Based Siting Process: Summary of Public Input, Final Report
The successful siting of nuclear waste management facilities, through a consent-based process, is predicated on a robust and transparent dialogue between the Department and its stakeholders. From December 23, 2015 through July 31, 2016, DOE solicited and received public comments and heard input from a wide range of individuals, communities, states, Tribes, and stakeholders through the multiple avenues described above.
Yucca Mountain Project Map, YMP–03–024.2, “Proposed Land Withdrawal” July 21, 2005.
Yucca Mountain Project Map, YMP–03–024.2, “Proposed Land Withdrawal” July 21, 2005.
Yucca Mountain Project Map, YMP–03–024.2, “Proposed Land Withdrawal” and dated July 21, 2005. This map is referenced in House of Representatives Bill H.R.3053 — 115th Congress (2017-2018) dated June 26,2017
Nuclear Power Plant Infrastructure Evaluations for Removal of Spent Nuclear Fuel (2024)
Nuclear Power Plant Infrastructure Evaluations for Removal of Spent Nuclear Fuel (2024)
This report fulfills the M3 milestone M3SF-23PN0203020614, “Updated NPP Site Evaluation Report (2).” This report is an update of the 2021 report Nuclear Power Plant Infrastructure Evaluations for Removal of Spent Nuclear Fuel and includes expansion of the site evaluations to include operating nuclear power plant (NPP) sites and to incorporate updated site inventory data. Figures that include the number of spent nuclear fuel (SNF) assemblies and metric tons heavy metal (MTHM) in a single figure have also been added to the report.
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).
Frequency of SNF Misload for Uncanistered Fuel Waste Package
Frequency of SNF Misload for Uncanistered Fuel Waste Package
The purpose of this 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. An actual criticality event can not occur in a WP unless a moderator (e.g. water) is present.
In-Drift Precipitates/Salts Model
In-Drift Precipitates/Salts Model
This report documents the development and validation of the in-drift precipitates/salts (IDPS) process model. The IDPS process model is a geochemical model designed to predict the postclosure effects of evaporation and deliquescence on the chemical composition of water within the Engineered Barrier System (EBS) in support of the total system performance assessment (TSPA). Application of the model in support of TSPA is documented in Engineered Barrier System: Physical and Chemical Environment (BSC 2005 [DIRS 175083]).
Fast Flux Test Facility (FFTF) Reactor Fuel Criticality Calculations
Fast Flux Test Facility (FFTF) Reactor Fuel Criticality Calculations
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.
DHLW Glass Waste Package Criticality Analysis (SCPB: N/A)
DHLW Glass Waste Package Criticality Analysis (SCPB: N/A)
This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the Defense High-Level Waste (DHLW) Glass waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan (Ref. 5.1) for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives.
Criticality Evaluation of Degraded Internal Configurations for a 44 BWR Waste Package
Criticality Evaluation of Degraded Internal Configurations for a 44 BWR Waste Package
The purpose of this calculation is to perform an example criticality evaluation for degraded internal configurations of a boiling water reactor (BWR) waste package (WP) containing 44 spent nuclear fuel (SNF) assemblies.
Description of Fracture Systems for External Criticality Reports
Description of Fracture Systems for External Criticality Reports
The purpose of this Analysis/Model Report (AMR) is to describe probabilistically the main features of the geometry of the fracture system in the vicinity of the repository. They will be used to determine the quantity of fissile material that could accumulate in the fractured rock underneath a waste package as it degrades. This AMR is to feed the geochemical calculations for external criticality reports. This AMR is done in accordance with the technical work plan (BSC (Bechtel SAIC Company) 2001 b). The scope of this AMR is restricted to the relevant parameters of the fracture system.
EQ6 Calculation for Chemical Degradation of Shippingport PWR (HEU Oxide) Spent Nuclear Fuel Waste Packages
EQ6 Calculation for Chemical Degradation of Shippingport PWR (HEU Oxide) 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 Shippingport Pressurized Water Reactor (PWR) (Ref. 1). The Shippingport PWR SNF has been considered for disposal at the proposed Yucca Mountain site.
Isotopic Generation and Confirmation of the BWR Appl. Model
Isotopic Generation and Confirmation of the BWR Appl. Model
The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from boiling water reactors (BWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the Disposal Criticality Analysis Methodology Topical Report (Reference 7.1).
Range of Neutronic Parameters Calculation File
Range of Neutronic Parameters Calculation File
The purpose of this engineering calculation is to document the benchmark range, over a variety of parameters, for the validation of the criticality calculations supporting the Monitored Geologic Repository (MGR). This engineering calculation accomplishes this by characterizing the Laboratory Critical Experiments (LCE) and the Pressurized Water Reactor (PWR) Commercial Reactor Criticals (CRC), and summarizing the significant parameters. This engineering calculation supports the Disposal Criticality Analysis Methodology program.
Operational Waste Stream Assumption for TSLCC Estimates
Operational Waste Stream Assumption for TSLCC Estimates
This document provides the background and basis for the operational waste stream used in the 2000 Total System Life Cycle Cost (TSLCC) estimate for the Civilian Radioactive Waste Management System (CRWMS). This document has been developed in accordance with its Development Plan (CRWMS M&O 2000a), and AP-3.11Q, ''Technical Reports''.