Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Technical Design Report”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 703 records · Page 39
All sourcesNASA NTRSDOE OSTI
NASA NTRS1971

Aircraft Propulsion

The Lewis Research Center is the National Aeronautics and Space Administration's principal field installation for research and development of advanced-aerospace-propulsion and power-generating systems. More specifically, a substantial part of the Center's activities is devoted to progress in the technology of aircraft propulsion. Work includes such diverse areas as components, controls and other aspects of installation, V/STOL and low-cost-engine designs and noise reduction. The results of this work are published as NASA reports and as articles in the technical journals. In addition, an occasional technical conference assists us in communicating more directly with others in the engineering fraternity. Accordingly, this conference - in which the NASA Ames, Flight, and Langley Research Centers are represented where they are working in related areas - is held to present the results of our recent and current work.

Source record↗
NASA NTRS1972

Space Storable Propellant Performance Gas/Liquid Like-Doublet Injector Characterization

A 30-month applied research program was conducted, encompassing an analytical, design, and experimental effort to relate injector design parameters to simultaneous attainment of high performance and component (injector/thrust chamber) compatibility for gas/liquid space-storable propellants. The gas/liquid propellant combination selected for study was FLOX (82.6% F2)/ambient temperature gaseous methane. The injector pattern characterized was the like-(self)-impinging doublet. Program effort was apportioned into four basic technical tasks: injector and thrust chamber design, injector and thrust chamber fabrication, performance evaluation testing, and data evaluation and reporting. Analytical parametric combustion analyses and cold flow distribution and atomization experiments were conducted with injector segment models to support design of injector/thrust chamber combinations for hot fire evaluation. Hot fire tests were conducted to: (1) optimize performance of the injector core elements, and (2) provide design criteria for the outer zone elements so that injector/thrust chamber compatibility could be achieved with only minimal performance losses.

Falk, A. Y.↗
NASA NTRS1974

MSFC Skylab Orbital Workshop, volume 1

The technical aspects of the Skylab-Orbital Workshop are discussed. Original concepts, goals, design philosophy, hardware, and testing are reported. The final flight configuration, overall test program, and mission performance are analyzed. The systems which are examined are: (1) the structural system, (2) the meteoroid shield systems, and (3) the environmental/thermal control subsystem.

Source record↗
NASA NTRS1995

Space Mechanisms Lessons Learned Study. Volume 1: Summary

Hundreds of satellites have been launched to date. Some have operated extremely well and others have not. In order to learn from past operating experiences, a study was conducted to determine the conditions under which space mechanisms (mechanically moving components) have previously worked or failed. The study consisted of: (1) an extensive literature review that included both government contractor reports and technical journals; (2) communication and visits (when necessary) to the various NASA and DOD centers and their designated contractors (this included contact with project managers of current and prior NASA satellite programs as well as their industry counterparts); (3) requests for unpublished information to NASA and industry; and (4) a mail survey designed to acquire specific mechanism experience. The information obtained has been organized into two volumes. Volume 1 provides a summary of the lessons learned, the results of a needs analysis, responses to the mail survey, a listing of experts, a description of some available facilities and a compilation of references. Volume 2 contains a compilation of the literature review synopsis.

Shapiro, Wilbur↗
NASA NTRS2011

A Case Study of Measuring Process Risk for Early Insights into Software Safety

In this case study, we examine software safety risk in three flight hardware systems in NASA's Constellation spaceflight program. We applied our Technical and Process Risk Measurement (TPRM) methodology to the Constellation hazard analysis process to quantify the technical and process risks involving software safety in the early design phase of these projects. We analyzed 154 hazard reports and collected metrics to measure the prevalence of software in hazards and the specificity of descriptions of software causes of hazardous conditions. We found that 49-70% of 154 hazardous conditions could be caused by software or software was involved in the prevention of the hazardous condition. We also found that 12-17% of the 2013 hazard causes involved software, and that 23-29% of all causes had a software control. The application of the TPRM methodology identified process risks in the application of the hazard analysis process itself that may lead to software safety risk.

Layman, Lucas↗
NASA NTRS1970

Reliability Abstracts and Technical Reviews January - December 1970: R70-14805 - R70-15438 - Volume 10, Nos. 1-12

Reliability Abstracts and Technical Reviews is an abstract and critical analysis service covering published and report literature on reliability. The service is designed to provide information on theory and practice of reliability as applied to aerospace and an objective appraisal of the quality, significance, and applicability of the literature abstracted.

Source record↗
DOE OSTI2015

Maine PACE Program Final Technical Report

The ARRA EECBG BetterBuilding helped augment the existing Home Energy Savings Programs (HESP) and incentives with financing through a subordinate lien PACE and HUD PowerSaver programs. The program was designed to document innovative techniques to dramatically increase the number of homes participating in weatherization programs in participating towns. Maine will support new energy efficiency retrofit pilots throughout the state, designed to motivate a large number of homeowners to invest in comprehensive home energy efficiency upgrades to bring real solutions to market.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2021

New High-Energy & Safe Battery Technology with Extreme Fast Charging Capability for Automotive Applications (Final Technical Report)

Being able to refill a cars gas tank when low on fuel is incredibly convenient and is a lacking feature for battery elective vehicles. New High-Energy & Safe Battery Technology with Extreme Fast Charging Capability for Automotive Applications was a research project designed to help correct this lacking feature by developing automotive sized Li-ion battery cells and demonstrating their ability to charge in under 10-minutes. To address the technical challenges Microvast, a cell manufacturer, was joined by Argonne National Labs, battery technology innovator, and BMW, a renown OEM, worked together to define and validate materials and cells for this project. By projects end, Microvast had successfully built and demonstrated a 240 Wh/kg, 35 Ah pouch cell that was capable of over 800 10-minute charge, full discharge cycles. From the outset of the project the plan was to use large amp-hour cells (ie automotive relevant sizes) to ensure the project findings and outcomes were most relevant for electric vehicle applications. The manufacture of cells was done using test manufacturing lines at Microvast and BMW that closely resemble the procedures of a Li-ion battery cell factory.

25 ENERGY STORAGE↗
DOE OSTI2011

Improved Creep-­Fatigue Models on Advanced Materials for SFR Applications (Final Report)

This report provides an update on the materials performance criteria and methodology relevant to sodium-cooled fast reactors (SFRs). The report is the second deliverable (Level 2) in FY11 (M2A11AN040303) under the work package A-11AN040303 “Materials Performance Criteria and Methodology” as part of the Advanced Structural Materials Program for the Advanced Reactor Concepts. The overall objective of the Materials Performance Criteria and Methodology work project is to evaluate the key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of advanced structural materials in support of the design and performance of sodium-cooled fast reactors. Advanced materials are a critical element in the development of fast reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced fast reactors. Qualification and licensing of advanced materials are prominent needs for the development and implementation of advanced fast reactor technologies. Nuclear structural component designs in the U.S. comply with the ASME Boiler and Pressure Vessel (B&PV) Code Section III (Rules for Construction of Nuclear Facility Components), and the NRC grants licensing. As the SFRs will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Section III Subsection NH (Class 1 Components in Elevated Temperature Service). A number of technical issues relevant to materials performance criteria and high temperature design methodology in the SFR were identified and presented in earlier reports. A viable approach to resolve these issues and the R&D priority were also recommended. The development of mechanistically based creep-fatigue interaction models for life prediction and reliable data extrapolation was chosen to be the central focus in near-term efforts. Our current focus is on the creep-fatigue damage issue in high-strength ferritic/martensitic steels such as mod.9Cr-1Mo (G91) and NF616 (G92) steels. The current ASME creep-fatigue design rule puts severe limits of fatigue and creep loads for G91 steel, the lead structural material for fast reactors. High-strength ferritic/martensitic steels behave fundamentally differently from austenitic stainless steels, for which the current ASME creep-fatigue design rules were developed. The unique deformation and damage characteristics in G91 steel, e.g. cyclic softening, degradation of creep and rupture strength during cyclic service, demands a new creep-fatigue design procedure that explicitly accounts for the material’s unique creep-fatigue behavior. G92 steel, a variant of G91 steel, is the lead candidate in the Advanced Alloy Development program. There is currently no creep-fatigue design rule available for this advanced alloy. To support the predictive model development and resolve the over-conservative issue with the ASME design rule for G91 steel, we recovered stress relaxation data from thirteen creep-fatigue tests conducted in the late 1980s and early 1990s at Oak Ridge National Laboratory, and conducted extensive data analysis in FY10. Based on this database and available literature data, we have developed a Cyclic Softening Model and a Stress Relaxation Model specifically applied to G91 steel. The Cyclic Softening Model describes the cyclic stress variation as a function of cycle number during creep-fatigue loading, and well captures the cyclic softening effect in G91 steel. The Stress Relaxation Model predicts the stress relaxation curve during the hold time of cyclic loading, and its dependence on the cyclic softening effect of G91 steel. Two creep-fatigue test systems have been set up to conduct model-oriented creep-fatigue experiments. Results from three creep-fatigue experiments on Heat 30176 of G91 steel are presented in the report: (1) a test performed at 550°C with total strain amplitude of 0.25% and a hold time of 60 s applied at peak strains in both tension and compression, (2) a test performed at 400°C with total strain amplitude of 0.5% and a tension and compression hold time of 60 s, and (3) a test performed at 20°C with total strain amplitude of 0.5% and a tension and compression hold time of 60 s. Three different test temperatures were chosen for two primary reasons: to determine the transition temperature of creep-fatigue interaction from a plasticity-dominated process to a creep-dominated process; to develop unified creep-fatigue models applicable for a wide range of temperatures. With the detailed information obtained in the creep-fatigue experiments, we have further refined the Cyclic Softening Model and Stress Relaxation Model. The Cyclic Softening Model and Stress Relaxation Model were incorporated into the Bilinear Creep-Fatigue Damage Model, forming a powerful framework for the life prediction by an improved creep-fatigue evaluation methodology. The cyclic softening and consecutive stress relaxation responses in G91 steel can be considered for each cycle of the entire cyclic process in the model calculation. The creep-fatigue damage in G91 steel was evaluated using the improved Bilinear Creep-Fatigue Damage Model without any safety margins. It was found that due to the significant cyclic softening in G91 steel, the unit creep damage in each cycle decreased continuously with increasing number of cycles. This continuous change in creep damage with cycling has significant implication in the evaluation of the total creep-fatigue damage. The total creep damage can be one order of magnitude lower when the unit creep damage at half-life was used in the evaluation in comparison with the calculations using the unit creep damage at cycle 10. An important conclusion was that the accurate assessment of the creep-fatigue damage in G91 steel must consider the history of cycling. This analysis provides a more realistic assessment of creep-fatigue damage and identifies critical factors that must be considered in the development of future advanced models for high-strength ferritic/martensitic alloys. An accelerated creep-fatigue testing methodology is proposed for the design of model-oriented experiments to guide and validate model developments. Future work will focus on creep-fatigue experiments of both G91 and G92 steels, and development of creep-fatigue models applicable to advanced ferritic/martensitic alloys. The Cyclic Softening Model and Stress Relaxation Model will be improved by incorporating microstructural parameters of dislocation density, subgrain structure, and second-phase particles. Collaborative modeling of combined Cyclic Softening Model, Stress Relaxation Model, and Interactive Damage Rate Model will be pursued for life prediction under creep-fatigue loading with extended hold time. Accelerated, well-controlled creep-fatigue tests will be conducted to provide high-quality experimental data to refine key material parameters and validate models.

Li, Meimei↗
DOE OSTI2021

Molten Salt Reactor Signatures and Modeling Study

Molten salt reactor (MSR) technologies, either liquid fueled and cooled or only liquid cooled, pose specific, unique challenges for safeguards of the special nuclear material during the operation, fueling, and maintenance of the reactor. MSRs are one type of Generation IV technologies being invested in and considered for U.S. domestic fabrication primarily for electricity and process heat production. These designs have generated growing commercial interest for several reasons, including high (≈40%) thermal efficiency, ease of fueling, improved use of uranium fuel, potential utilization of thorium fuel, and proposed inherent safety features. U.S. companies have several planned designs that differ in fuel, cooling, and neutron energy spectrum. Driven by commercial interest and the intent of licensing MSRs, the U.S. Nuclear Regulatory Commission (NRC) has developed a vision and strategy to accommodate non-light water reactors (LWR), which include MSRs (U.S. Nuclear Regulatory Commission 2019). In addition, the Department of Energy (DOE) through the Office of Nuclear Energy (NE) Office of Advanced Reactor Technologies (ART), …sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. Reactor types considering the use of salts, liquid metals, or gases for coolant fall under both ARC and aSMR. Therefore, Research Design & Development is being pursued by DOE-NE through national laboratories, universities, and international and industrial collaborations. Additionally, the U.S. is a member of the Gen IV International Forum (GIF). The GIF is a cooperative, multinational organization to guide and carry out research and development needed for the GEN IV reactor systems (Forum 2018). GIF evaluated numerous reactor concepts and down-selected to the six most feasible advanced reactor technologies: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), MSR, supercritical watercooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). In support of the growing interest domestically and internationally, the Materials Protection, Accounting, and Control Technologies (MPACT) campaign, under the DOE-NE Fuel Cycle Technologies (FCT) program, engages in R&D activities by developing advanced instrumentation and analysis for safeguards and security of modern, advanced nuclear fuel cycle (non-LWR) facilities. Because of the historic experience in the operation of the Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) (Robertson, MSRE Design and Operations Report Part I 1965), Oak Ridge National Laboratory (ORNL) is heavily engaged in the various R&D activities through the DOE complex related to MSRs including national technical leadership of the DOE-NE MSR campaign. This report discusses and presents the outcomes of the FY19 MPACT MSR Safeguards task. The challenges presented by MSRs for nuclear material accountancy and control (NMAC) and associated safeguards will be investigated. The objective of this research is to explore and compile the safeguards requirements and identify measurement signatures through an initial high-level MSR design and develop complementary advanced simulation and modeling capabilities. A high-level ORNL-developed MSR design called the Molten Salt Demonstration Reactor (MSDR) (Bettis, Alexander and Watts 1972) was used as the target reactor design for this research. The MSDR model incorporates technology from the MSRE and the Molten Salt Breeder Reactor (Robertson, Conceptual Design of a Single-Fluid Molten-Salt Breeder Reactor 1971). But the MSDR is a 750 MWth graphite moderated liquid fueled (low-enriched uranium) MSR compared to the MSRE’s of 7.5 MWth. The focus of this report is to discuss the evaluation of novel signatures, correlations, and indicators to understand the applicability of current safeguards instrumentation to MSRs using the modeling results from the MSDR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Proposed Risk-Informed Regulatory Framework for Approval of Microreactor Transportation Packages

Microreactors are very small nuclear reactors with a power output of about 20 megawatts electric or less that are designed to be factory-built, modular in nature, and highly portable. These compact reactors will be small enough to be transported by truck or even air and could help solve energy challenges in a number of areas, ranging from remote commercial or residential locations to military bases. Pacific Northwest National Laboratory is tasked to develop and evaluate transportation licensing options for microreactors. The work is funded by the National Reactor Innovation Center a National Department of Energy program led by Idaho National Laboratory for the Office of Nuclear Energy Research and Development which support demonstration of microreactor technology. Key transportation steps include the (1) initial movement of high-assay low enriched uranium fresh fuel, (2) transportation of an intact, but never-operated microreactor, and (3) transportation of an intact, previously-operated microreactor. The deliverables on the project consists of a documentation of applicable regulations and regulatory authority for transportation. The objective of this report is to propose a risk-informed regulatory framework for the licensing of the transportation of microreactors, including the transportation of irradiated nuclear fuel that is assumed to be an integral component of the microreactor transportation package. The framework lays out a viable regulatory pathway, including decision points for regulatory options and the supporting technical evaluations for those options in phases from near to long term. This report includes discussion of the (1) general microreactor design concepts including representative microreactor source terms, (2) options for regulatory approval of microreactor transportation based on current regulation and historical precedence, (3) regulatory basis for including risk information in microreactor transportation licensing activities, and (4) description of a risk-informed regulatory framework.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Proposed Risk-Informed Regulatory Framework for Approval of Microreactor Transportation Packages

Microreactors are very small nuclear reactors with a power output of about 20 megawatts electric or less that are designed to be factory-built, modular in nature, and highly portable. These compact reactors will be small enough to be transported by truck or even air and could help solve energy challenges in a number of areas, ranging from remote commercial or residential locations to military bases. Pacific Northwest National Laboratory is tasked to develop and evaluate transportation licensing options for microreactors. The work is funded by the National Reactor Innovation Center a National Department of Energy program led by Idaho National Laboratory for the Office of Nuclear Energy Research and Development which support demonstration of microreactor technology. Key transportation steps include the (1) initial movement of high-assay low enriched uranium fresh fuel, (2) transportation of an intact, but never-operated microreactor, and (3) transportation of an intact, previously-operated microreactor. The deliverables on the project consists of a documentation of applicable regulations and regulatory authority for transportation. The objective of this report is to propose a risk-informed regulatory framework for the licensing of the transportation of microreactors, including the transportation of irradiated nuclear fuel that is assumed to be an integral component of the microreactor transportation package. The framework lays out a viable regulatory pathway, including decision points for regulatory options and the supporting technical evaluations for those options in phases from near to long term. This report includes discussion of the (1) general microreactor design concepts including representative microreactor source terms, (2) options for regulatory approval of microreactor transportation based on current regulation and historical precedence, (3) regulatory basis for including risk information in microreactor transportation licensing activities, and (4) description of a risk-informed regulatory framework.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

WTP Waste Feed Qualification: Glass Fabrication Unit Operation Testing Report

The waste feed qualification program is being developed to protect the Hanford Tank Waste Treatment and Immobilization Plant (WTP) design, safety basis, and technical basis by assuring waste acceptance requirements are met for each staged waste feed campaign prior to transfer from the Tank Operations Contractor to the feed receipt vessels inside the Pretreatment Facility. The Waste Feed Qualification Program Plan describes the three components of waste feed qualification: 1. Demonstrate compliance with the waste acceptance criteria 2. Determine waste processability 3. Test unit operations at laboratory scale. The glass fabrication unit operation is the final step in the process demonstration portion of the waste feed qualification process. This unit operation generally consists of combining each of the waste feed streams (high-level waste (HLW) and low-activity waste (LAW)) with Glass Forming Chemicals (GFCs), fabricating glass coupons, performing chemical composition analysis before and after glass fabrication, measuring hydrogen generation rate either before or after glass former addition, measuring rheological properties before and after glass former addition, and visual observation of the resulting glass coupons. Critical aspects of this unit operation are mixing and sampling of the waste and melter feeds to ensure representative samples are obtained as well as ensuring the fabrication process for the glass coupon is adequate. Testing was performed using a range of simulants (LAW and HLW simulants), and these simulants were mixed with high and low bounding amounts of GFCs to evaluate the mixing, sampling, and glass preparation steps in shielded cells using laboratory techniques. The tests were performed with off-the-shelf equipment at the Savannah River National Laboratory (SRNL) that is similar to equipment used in the SRNL work during qualification of waste feed for the Defense Waste Processing Facility (DWPF) and other waste treatment facilities at the Savannah River Site. It is not expected that the exact equipment used during this testing will be used during the waste feed qualification testing for WTP, but functionally similar equipment will be used such that the techniques demonstrated would be applicable. For example, the mixing apparatus could use any suitable mixer capable of being remoted and achieving similar mixing speeds to those tested.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2019

Astrophysics and Technical Study of a Solar Neutrino Spacecraft: NIAC (NASA Innovative Advanced Concepts) Grant - NASA 2018-Phase-1

We report on our study of the design of a neutrino detector, shielding and veto array needed to operate a neutrino detector in space close to the Sun. This study also took into account the expected rates of Galactic gamma and cosmic rays in addition to the particles from the Sun.These preliminary studies show that we can devise a detector such that a small signal of neutrino interactions can be extracted from a large random number of events from the background sources using a double timing method from the conversion electron produced in the neutrino interaction and a secondary delayed signal from the nuclear excited state produced from the initial neutrino interaction; in our case the conversion of Ga 69 or 71 into Ge 69 or 71, but this method could apply to other nuclei with large neutrino cross sections such as Ir 115. Although these types of events need to be above 0.405 megaelectronvolt (MeV) neutrino energy and are only 66 percent of all conversion neutrino interactions on Gallium, this is a small price to pay for an increase of 10,000 by going close to the Sun to enhance the neutrino rate over the background combatorical fake-signal events. The conclusion of this Phase-1 study is very positive in that we can get the backgrounds less than 20 percent fake signals, and in addition to this we have devised another shielding method that makes the Galactic gamma-ray rate a hundred fold less which will make further improvements over these initial estimates. Although these studies are very encouraging it suggests that the next step is a NIAC Phase-II to actually build a test device,measuring basic principles such as light attention within the scintillator with high dopants and to take data in the lab with a cosmic-ray test stand and triggered X-ray source for comparison with simulated expected performance of the detector. This would be the perfect lead into a future proposal beyond a NIAC (NASA Innovative Advanced Concepts) Phase-II for a test flight of a small one-pint detector in orbit of the detector concept beyond Earth outside of the radiation belts.

Astrophysics↗
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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗