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Blink, James A.

Publications and source records attributed to Blink, James A..

18 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems (Rev.2)

This performance specification document addresses use of a standardized transportation, aging, and disposal (STAD) canister system in the spent nuclear fuel (SNF) waste management system. This document was developed to support STAD canister system studies and potential research, development, and demonstration activities. Requirements in this specification may evolve with time based on results from analyses, experiments, design studies, system evaluations and demonstrations, as well as other factors. A STAD canister system consists of a canister, together with a storage or aging overpack/module/vault, a shielded transfer cask (STC), a site transporter, a transportation overpack, and a transportation skid. Three sizes of circular cross section STAD canister variants are specified in this document. The small STAD canister capacity is four pressurized water reactor (PWR) or nine boiling water reactor (BWR) SNF assemblies. The medium STAD canister capacity is 12 PWR or 32 BWR SNF assemblies. The large STAD canister capacity is 21 PWR or 44 BWR SNF assemblies. Using standardized canisters for SNF facilitates the interface between storage at an independent spent fuel storage installation (ISFSI) or a DOEowned interim storage facility (ISF), transportation between facilities, and disposal at a repository. Multiple STAD canisters may be stored in a single storage overpack and/or a single transportation overpack. The US Nuclear Regulatory Commission (NRC) requirements currently used for storage and transportation are adopted for the STAD canister system. Approved approaches (guidance) in NRC storage and transportation documents are also adopted as requirements for the STAD canister system to facilitate obtaining NRC Certificates of Compliance (CoCs) for STAD canister systems. Performance specifications in this document are also informed by STAD canister disposability in various geologic media and consideration of ISF (and geologic repository) handling facility design, operations, and licensing requirements.

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DOE OSTI2015

Rationale for the Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems (Rev.1)

This document provides the rationale for the specifications for the standardized transportation, aging, and disposal (STAD) canister systems provided in Performance Specification for Standardized Transportation, Aging, and Disposal Canister Systems, FCRD-NFST-2014-000579, Rev. 2, referred to as STAD Spec herein. The STAD Spec was developed to support STAD canister system studies and potential research, development, and demonstration activities. Requirements in the specification may evolve with time based on results from analyses, experiments, design studies, system evaluations and demonstrations, as well as other factors. The STAD canister system consists of a canister, together with a storage or aging overpack/module/vault, a shielded transfer cask (STC), a site transporter, a transportation overpack, and a transportation skid. Three sizes of circular cross section STAD canister variants are specified. The small STAD canister capacity is four pressurized water reactor (PWR) or nine boiling water reactor (BWR) spent nuclear fuel (SNF) assemblies. The medium STAD canister capacity is 12 PWR or 32 BWR SNF assemblies. The large STAD canister capacity is 21 PWR or 44 BWR SNF assemblies. There are 60 specifications in the STAD Spec that cover the STAD canister general design attributes, canister safety functional requirements, operational considerations, materials, storage and aging system requirements, and transportation system requirements. The 60 specifications are derived from regulatory requirements, regulatory guidance, STAD canister disposability considerations, and interim storage facility (ISF) and geologic repository handling facility considerations.

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DOE OSTI2013

Disposal Systems Evaluation Framework DSEF User Manual (V.3.0)

The Disposal Systems Evaluation Framework (DSEF) has been developed at Lawrence Livermore National Laboratory (LLNL) to formalize and facilitate the development and documentation of repository conceptual design options for a range of waste forms, geologic environments, repository design concepts, and repository operating modes. DSEF is one component of the Engineered Barrier System (EBS) family of work packages at several national laboratories led by Sandia National Laboratories (SNL), in support of the DOE Office of Nuclear Energy, specifically of the Used Fuel Disposition (UFD) Campaign within the Office of Used Nuclear Fuel Disposition Research and Development (NE-53). DSEF is a knowledge management system that allows the user to intelligently access and draw data from a case library of hundreds of completed thermal analyses (currently 721 cases in the library), and draw input from databases of material properties and repository development cost data (currently the data in these databases are drawn from more than 100 references). The core functionality of DSEF is provided by a Microsoft Office Excel 2010 workbook with macros and form controls that create a structured environment and walk the user through the steps of creating the input data required to perform disposal system evaluations internally and by interfacing with external programs. The user can choose to work with built-in data from the case library of previous analysis cases, from the material property databases, or can define their own input descriptions and data. DSEF uses the Excel worksheet protection (implemented on a sheet by sheet basis) to prevent inadvertent over-writing of formulas with data. When worksheet protection is enabled, the user can see all underlying formulas, and enter data in unprotected cells (the unprotected cells generally correspond to the yellow highlighted cells). Worksheet protection also prevents restructuring the files (adding or deleting rows or columns) and changing the formatting, as well as preventing access to the Visual Basic Editor to prevent modification of the macros. If the user needs to modify the worksheet structure, formulas, or macros, they can contact LLNL to get the password to unprotect the worksheets. However, care should be taken in making any modifications, since the macros and formulas are dependent in some cases on the current structure, range names, and macro subroutines DSEF is set up to retrieve the results of external thermal analysis program runs, summarize them graphically, and provide a template for adding the new results to the Case Library. The Case Library worksheet in DSEF is set up to create a Case Library data record for each new analysis case, at the top of the list of prior cases. Therefore, a separate DSEF file (with unique name) should be saved for each run that is to be archived. Users should submit a copy of their DSEF files to LLNL so that their results can be added to the overall Case Library and made available to other laboratories and stakeholders. LLNL will maintain the DSEF Case Library and make the most current files available on the SNL SharePoint site. The Revision History sheet documents the file evolution. One of the key objectives of DSEF is to allow the user to compare the results of thermal analyses to the thermal constraints of the design components. These comparisons can then be used to consider design, layout, and operating mode modifications to find combinations that can meet the thermal constraints for a given combination of waste form, geologic medium, surface storage time, and system concept of operation.

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DOE OSTI2013

Development of a Generalized Methodology for Soil-Structure Interaction Analysis Using Nonlinear Time-Domain TechniquesNEAMS Program, DOE Office of Nuclear Energy (NE-41)

A generalized time-domain method for Soil-Structure Interaction Analysis is developed, based upon an extension of the Bielak Method. The methodology is combined with the use of a simple hysteretic soil model based upon the Ramberg-Osgood formulation and applied to a notional Small Modular Reactor. These benchmark results compare well with those obtained by using the industry-standard frequency-domain code SASSI. The methodology provides a path forward for investigation of other sources of nonlinearity, including those associated with the use of more physically-realistic material models incorporating pore-pressure effects, gap opening/closing, the effect of nonlinear structural elements, and 3D seismic inputs.

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DOE OSTI2013

Nuclear Energy Return on Energy Investment

This report provides a methodology and requisite data to assess the potential Energy Return on (Energy) Investment (EROI) for nuclear fuel cycle alternatives, and applies that methodology to an example “once-through” fuel cycle using low enrichment uranium (LEU) in conventional light water cooled reactors (LWRs). Within the DOE Office of Nuclear Energy Science and Technology - Fuel Cycle Technologies program, the Fuel Cycle Options Campaign “performs integrating analyses of nuclear energy and fuel cycle systems to inform fuel cycle R&D, programmatic decisions, strategy formulation, and policy development”. Campaign objectives include development of relevant fuel cycle metrics, and development of tools and associated data for analysis of fuel cycle systems. This study represents an extension of a prior evaluation of EROI as a metric for fuel cycle facilities, processes and technologies. That prior study [Simon 2011] addressed the energy return on the addition of fuel recycle to an existing nuclear energy system. Limited to just the addition of fuel recycle, that study did not include all the energy investments required to create, operate and decommission the underlying nuclear fuel cycle, such as uranium mining, fuel fabrication, reactor construction, and used fuel disposition. This extension of the prior work adds these remaining pieces of the fuel cycle to provide a basic evaluation framework and initial data to enable evaluation of EROI for nuclear energy in general. It is intended as a basis for evaluation of alternative fuel cycle options in the future with the addition of pertinent details for other fuel cycle scenarios. In the prior work, a spreadsheet tool was constructed for the specific purpose of this energy input/output analysis of nuclear fuel cycle facilities. This tool allows the user to enter the parameters of the nuclear fuel cycle and assumptions about energy use in reprocessing. In the current analysis, energy consumption for an entire nuclear energy enterprise is considered, including facility construction, materials, facility operation, and decommissioning. The analysis tool has been extended to include the entire fuel cycle for a representative scenario of a once-through LWR nuclear energy system. Energy content data have been developed for an initial representation. In many cases where there is a valid range of plausible energy content, the range is discussed and a representative value selected for demonstration of the evaluation methodology. In this analysis, energy produced is the energy output of power reactors within the fuel cycle. The spreadsheet calculates both Primary EROI and Final EROI for the fuel cycle entered. The EROI is the ratio of the output energy divided by the consumed energy. The intent for this study was to develop the methodology and analysis tool for a complete fuel cycle. It is not the intent of the study to fully explore the wide range of potential energy intensities, or to reconcile the disparate values found in the literature. Representative numbers were used in this demonstration. The methodology and tool do provide a framework for future exploration of the key energy intensity values and for conducting sensitivity studies on specific values, either to assess improved understanding of the values, or to explore the potential for alternative technologies to impact EROI. The Primary and Final Energy EROI values calculated for the representative scenario were ~ 52 and 24, respectively.

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DOE OSTI2012

Application of Analytical Heat Transfer Models of Multi-layered Natural and Engineered Barriers in Potential High-Level Nuclear Waste Repositories - 12435

A combination of transient heat transfer analytical solutions for a finite line source, a series of point sources, and a series of parallel infinite line sources were combined with a quasi-steady-state multi-layered cylindrical solution to simulate the temperature response of a deep geologic radioactive waste repository with multi-layered natural and engineered barriers. This evaluation was performed to provide information to scientists and decision makers to compare candidate geologic media for a repository (crystalline rock [granite], clay, salt, and deep borehole), and to provide input for the future evaluation of the trade-off between pre-emplacement surface storage time, waste package size, and repository footprint. This approach was selected in favor of the finite element solution typically used to analyze the temperature response because it allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. The analytical solution approach used to analyze the repository temperature response allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. This approach allowed investigation of the sensitivity of the results to combinations of parameters that show that there is much flexibility to be gained in terms of spent fuel management options by varying a few key parameters. This initial analysis used representative design concepts and thermal constraints based on international design concepts, and it also included waste forms representing future fuel cycles with high burnup fuels. Unlike repository designs with large open tunnels and pre-closure ventilation, all of the disposal concepts analyzed in this study used enclosed emplacement modes, where the waste packages were in direct contact with encapsulating engineered or natural materials. The deep borehole repository concept limits the size of the SNF waste packages and may require rod consolidation to fit within the drill casing diameter. A single assembly waste package, assuming rod consolidation, was evaluated in the current analysis. Similar size restrictions apply for the HLW canisters. At this time no thermal constraints have been defined for the deep borehole repository concept. Representative EBS materials and properties were evaluated. However, changes in EBS design concepts and materials can also have significant effects on the maximum waste package surface temperature. Increased thermal conductivity of the buffer layer can be achieved by using an engineered buffer consisting of a mixture of graphite, sand, and bentonite [14]. One of the advantages of the analytical model is that it highlights the sensitivity of the results to the parameters that define the repository layout, including spacing between axial and lateral neighboring waste packages and drifts. It is clear that repository layout adjustments can be made to reduce the calculated peak temperatures. The results also show that significant reductions in required surface storage times can be achieved if higher thermal constraints can be justified Additional studies are planned to evaluate the trade-offs between surface storage times, repository layout parameters, and variations in EBS design concepts. Model validation and uncertainties will also be addressed. It is expected that shorter surface storage times and more optimized repository design configurations may be achieved. (authors)

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DOE OSTI2012

Advanced Nuclear Fuel Cycle Effects on the Treatment of Uncertainty in the Long-Term Assessment of Geologic Disposal Systems - EBS Input

The Used Fuel Disposition (UFD) Campaign within the Department of Energy's Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation's spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. The planning, construction, and operation of a nuclear disposal facility is a long-term process that involves engineered barriers that are tailored to both the geologic environment and the waste forms being emplaced. The UFD Campaign is considering a range of fuel cycles that in turn produce a range of waste forms. The UFD Campaign is also considering a range of geologic media. These ranges could be thought of as adding uncertainty to what the disposal facility design will ultimately be; however, it may be preferable to thinking about the ranges as adding flexibility to design of a disposal facility. For example, as the overall DOE-NE program and industrial actions result in the fuel cycles that will produce waste to be disposed, and the characteristics of those wastes become clear, the disposal program retains flexibility in both the choice of geologic environment and the specific repository design. Of course, other factors also play a major role, including local and State-level acceptance of the specific site that provides the geologic environment. In contrast, the Yucca Mountain Project (YMP) repository license application (LA) is based on waste forms from an open fuel cycle (PWR and BWR assemblies from an open fuel cycle). These waste forms were about 90% of the total waste, and they were the determining waste form in developing the engineered barrier system (EBS) design for the Yucca Mountain Repository design. About 10% of the repository capacity was reserved for waste from a full recycle fuel cycle in which some actinides were extracted for weapons use, and the remaining fission products and some minor actinides were encapsulated in borosilicate glass. Because the heat load of the glass was much less than the PWR and BWR assemblies, the glass waste form was able to be co-disposed with the open cycle waste, by interspersing glass waste packages among the spent fuel assembly waste packages. In addition, the Yucca Mountain repository was designed to include some research reactor spent fuel and naval reactor spent fuel, within the envelope that was set using the commercial reactor assemblies as the design basis waste form. This milestone report supports Sandia National Laboratory milestone M2FT-12SN0814052, and is intended to be a chapter in that milestone report. The independent technical review of this LLNL milestone was performed at LLNL and is documented in the electronic Information Management (IM) system at LLNL. The objective of this work is to investigate what aspects of quantifying, characterizing, and representing the uncertainty associated with the engineered barrier are affected by implementing different advanced nuclear fuel cycles (e.g., partitioning and transmutation scenarios) together with corresponding designs and thermal constraints.

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