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At least 1,711 records · Page 95
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Light Water Reactor Sustainability Program: Integrated Program Plan

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 60%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to grow by about 24% from 2013 to 2040 . At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license, for a total of 60 years of operation (the oldest commercial plants in the United States reached their 40th anniversary in 2009). Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity for 40- and 60-year license periods. If current operating nuclear power plants do not operate beyond 60 years (and new nuclear plants are not built quickly enough to replace them), the total fraction of generated electrical energy from nuclear power will rapidly decline. That decline will be accelerated if plants are shut down before 60 years of operation. Decisions on extended operation ultimately rely on economic factors; however, economics can often be improved through technical advancements. The U.S. Department of Energy Office of Nuclear Energy's 2010 Research and Development Roadmap (2010 Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: 1. Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; 2. Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; 3. Develop sustainable nuclear fuel cycles; and 4. Understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2010

Effect of Creep of Ferritic Interconnect on Long-Term Performance of Solid Oxide Fuel Cell Stacks

High-temperature ferritic alloys are potential candidates as interconnect (IC) materials and spacers due to their low cost and coefficient of thermal expansion (CTE) compatibility with other components for most of the solid oxide fuel cells (SOFCs) . However, creep deformation becomes relevant for a material when the operating temperature exceeds or even is less than half of its melting temperature (in degrees of Kelvin). The operating temperatures for most of the SOFCs under development are around 1,073 K. With around 1,800 K of the melting temperature for most stainless steel, possible creep deformation of ferritic IC under the typical cell operating temperature should not be neglected. In this paper, the effects of IC creep behavior on stack geometry change and the stress redistribution of different cell components are predicted and summarized. The goal of the study is to investigate the performance of the fuel cell stack by obtaining the changes in fuel- and air-channel geometry due to creep of the ferritic stainless steel IC, therefore indicating possible changes in SOFC performance under long-term operations. The ferritic IC creep model was incorporated into software SOFC-MP and Mentat-FC, and finite element analyses were performed to quantify the deformed configuration of the SOFC stack under the long-term steady-state operating temperature. It was found that the creep behavior of the ferritic stainless steel IC contributes to narrowing of both the fuel- and the air-flow channels. In addition, stress re-distribution of the cell components suggests the need for a compliant sealing material that also relaxes at operating temperature.

SOFC. creep, ferritic interconnects, long-term per↗
DOE OSTI2012

A Prototype Performance Assessment Model for Generic Deep Borehole Repository for High-Level Nuclear Waste - 12132

A deep borehole repository is one of the four geologic disposal system options currently under study by the U.S. DOE to support the development of a long-term strategy for geologic disposal of commercial used nuclear fuel (UNF) and high-level radioactive waste (HLW). The immediate goal of the generic deep borehole repository study is to develop the necessary modeling tools to evaluate and improve the understanding of the repository system response and processes relevant to long-term disposal of UNF and HLW in a deep borehole. A prototype performance assessment model for a generic deep borehole repository has been developed using the approach for a mined geological repository. The preliminary results from the simplified deep borehole generic repository performance assessment indicate that soluble, non-sorbing (or weakly sorbing) fission product radionuclides, such as I-129, Se-79 and Cl-36, are the likely major dose contributors, and that the annual radiation doses to hypothetical future humans associated with those releases may be extremely small. While much work needs to be done to validate the model assumptions and parameters, these preliminary results highlight the importance of a robust seal design in assuring long-term isolation, and suggest that deep boreholes may be a viable alternative to mined repositories for disposal of both HLW and UNF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2026

Fuel Heating System Design and Thermal Modeling for Aviation Combustion Research

Aviation fuels offer substantial heat-sink capability that can be applied to aircraft thermal-management strategies and advanced propulsion-system concepts. Accurate control of fuel temperature is critical for understanding combustion behavior and supporting high temperature fuel experimental studies. This work describes the modeling and experimental validation of a portable, high temperature fuel-conditioning system designed for multiple NASA Glenn Research Center’s combustion facilities. The system utilizes a circulating-oil heating unit and modular tube-in-tube heat exchangers capable of conditioning Jet-A to temperatures approaching 600 K. A Thermal Desktop model was created to analyze heat exchanger configurations and predict performance across a range of operating conditions. Model predictions were compared with measurements collected in the CE-13C test facility, showing good agreement across the tested flow range except at low fuel flow rate around 2 kg/hr. Results provide guidance for sizing future fuel heating system for combustion research applications.

Jet fuel heating system↗
DOE OSTI2023

Multi-scale fission product release model with comparison to AGR data

TRistructural ISOtropic (TRISO) particle fuel is central to several advanced, high-temperature reactor designs. Each particle consists of a fuel kernel encapsulated by three layers of carbon and ceramics that prevent the release of fission products and ensure physical integrity. Despite outstanding retention properties, fission product release has been observed from intact particles. To better understand and quantify fission product release from TRISO particles, a multiscale, mechanistic model of fission product transport is being developed by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Previous work focused on silver (Ag) transport and improved Ag release predictions. The work described in this report builds on this experience to better understand cesium (Cs) transport in silicon carbide (SiC), the main barrier to the release of fission products. Atomistic simulations provide bulk and grain boundary (GB) Cs diffusivities in SiC, which are used by phase field simulations in the mesoscale code Marmot to determine the temperature, microstructure, and irradiation-dependent Cs diffusivity at the mesoscale in SiC. This approach attributes the different temperature regimes experimentally observed for Cs diffusivities in SiC to a transition from bulk-dominated diffusivity at high temperatures to a GB-dominated regime at low temperatures, providing new insight. The multiscale, mechanistic effective diffusivity is then implemented in the fuel performance code BISON and further validated by comparing Cs release predictions from Advanced Gas Reactor (AGR)-1 and AGR-2 post-irradiation measurements. The new model improves BISON’s predictability. This document also reports improvements made on Ag transport modeling by accounting for different GB types having different diffusivities. Moreover, this report details preliminary efforts to model palladium (Pd) attack of the SiC at the mesoscale using a phase field approach. Pd attack and its impact on accelerated Ag transport remains a misunderstood phenomenon, and we use the model to demonstrate that the formation of lamellae that has been observed in experiments can be explained by the reaction of Pd with SiC to form alternating layers of graphite and Pd 2 Si. This effort aims to improve our understanding of the reaction and eventually provide a model for BISON to account for Pd penetration and its effects on fission product release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

ASSESSMENT OF RESIDUAL STRESSES ON U10MO ALLOY BASED MONOLITHIC MINI-PLATES DURING HOT ISOSTATIC PRESSING

This article presents an assessment of the residual stress field on U10Mo alloy based monolithic fuel plates and their elasto-plastic response to the thermo-mechanical fabrication process. Monolithic plate-type fuel for research and test reactors is a new fuel form proposed and developed by Idaho National Laboratory (INL) to accomplish higher uranium densities in the reactor core and allow multi-fold reduction of fuel enrichment. Identification of the 3D residual field is critical for benchmarking the new design and understanding the overall performance of the new plate-type fuels. To define fuel-cladding stress-strain characteristics, a new thermo-mechanical finite element model was developed. Material properties at high temperatures and material plasticity were considered carefully to improve the accuracy as the Hot Pressing temperature reaches the melting temperature of the cladding material. By using elasto-thermo-plastic material models, it was determined that the cladding material (AL6061-TO) is exposed to tensile stress field and exceeds its proportional limits, while the fuel foil (U10Mo) is under compression and remains below its yield limit. It was shown that the residual stresses would govern the overall behavior of the plates, and therefore, should not be neglected. In particular, the simulations have revealed the existence of stress gradients at the fuel/cladding interface, thus emphasizing the necessity of excellent bonding quality at the bond region for mechanical endurance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Start-up and Emergency Shutdown Modeling for a Coal-fired 10MWe sCO₂ Power Plant

As part of a Phase II Front End Engineering Design (FEED) study for the DOE Fossil Fuel Large Scale Pilots program, investigations of novel start-up and emergency shutdown methods for a stoker-fed, coal-fired, 10MWe sCO₂ power plant proposal were performed. These investigations were performed using 1D transient system models of the plant created within the GT-SUITE system modeling platform. The system model components were created using vendor quotation data (heat exchangers), performance maps (turbomachines), or FMU models (fired heater). Previously, Echogen Power Systems (Echogen) has used a sCO₂ power cycle start-up method that fills the system with liquid CO₂ to facilitate easy use of an electrically-driven start pump to transition the system from initial CO₂ fill to turbo-compressor initialization. The novel start-up method presented attempts this transition with a minimal filling of the system with liquid CO₂, to reduce the total CO₂ inventory required. This method leaves the start pump vulnerable to a two phase inlet condition as system pressure is below the CO₂ saturation pressure for the condenser cooling water temperature (Tcw). System model cases at Tcw of 18°C, 26°C, and 32°C, corresponding to cold, design, and hot ambient days, were analyzed to determine if the SP inlet condition could be kept as a subcooled liquid and estimate the CO₂ inventory reduction amount./p> Stoker-fed, coal-fired heaters continue to emit heat for some time even after emergency shutdown from events such as a power failure. This heat emission would lead to a failure of the fired heater, as the metal overheats, if the CO₂ flow is shut off. The emergency shutdown method presented utilizes CO₂ vented by a controllable vent valve (CRV) to provide CO₂ cooling flow to the fired heater. To determine the effectiveness of this method at keeping the fired heater peak tubing metal temperature below the ASME material temperature limit, for pressures below 5 MPa, of 816 °C, system model cases with CRV diameters ranging from 4” to 10”, and an alternative CO₂ vent routing with CRV size of 6”, were investigated./p>

01 COAL, LIGNITE, AND PEAT↗
NASA NTRS2015

Fuel Distribution Estimate via Spin Period to Precession Period Ratio for the Advanced Composition Explorer

The spin period to precession period ratio of a non-axisymmetric spin-stabilized spacecraft, the Advanced Composition Explorer (ACE), was used to estimate the remaining mass and distribution of fuel within its propulsion system. This analysis was undertaken once telemetry suggested that two of the four fuel tanks had no propellant remaining, contrary to pre-launch expectations of the propulsion system performance. Numerical integration of possible fuel distributions was used to calculate moments of inertia for the spinning spacecraft. A Fast Fourier Transform (FFT) of output from a dynamics simulation was employed to relate calculated moments of inertia to spin and precession periods. The resulting modeled ratios were compared to the actual spin period to precession period ratio derived from the effect of post-maneuver nutation angle on sun sensor measurements. A Monte Carlo search was performed to tune free parameters using the observed spin period to precession period ratio over the life of the mission. This novel analysis of spin and precession periods indicates that at the time of launch, propellant was distributed unevenly between the two pairs of fuel tanks, with one pair having approximately 20% more propellant than the other pair. Furthermore, it indicates the pair of the tanks with less fuel expelled all of its propellant by 2014 and that approximately 46 kg of propellant remains in the other two tanks, an amount that closely matches the operational fuel accounting estimate. Keywords: Fuel Distribution, Moments of Inertia, Precession, Spin, Nutation

Moment of Inertia↗
DOE OSTI2010

Low-temperature fuel cell systems for commercial airplane auxiliary power.

This presentation briefly describes the ongoing study of fuel cell systems on-board a commercial airplane. Sandia's current project is focused on Proton Exchange Membrane (PEM) fuel cells applied to specific on-board electrical power needs. They are trying to understand how having a fuel cell on an airplane would affect overall performance. The fuel required to accomplish a mission is used to quantify the performance. Our analysis shows the differences between the base airplane and the airplane with the fuel cell. There are many ways of designing a system, depending on what you do with the waste heat. A system that requires ram air cooling has a large mass penalty due to increased drag. The bottom-line impact can be expressed as additional fuel required to complete the mission. Early results suggest PEM fuel cells can be used on airplanes with manageable performance impact if heat is rejected properly. For PEMs on aircraft, we are continuing to perform: (1) thermodynamic analysis (investigate configurations); (2) integrated electrical design (with dynamic modeling of the micro grid); (3) hardware assessment (performance, weight, and volume); and (4) galley and peaker application.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2011

Nuclear Material Control and Accountability System Effectiveness Tool (MSET)

A nuclear material control and accountability (MC&A) system effectiveness tool (MSET) has been developed in the United States for use in evaluating material protection, control, and accountability (MPC&A) systems in nuclear facilities. The project was commissioned by the National Nuclear Security Administration's Office of International Material Protection and Cooperation. MSET was developed by personnel with experience spanning more than six decades in both the U.S. and international nuclear programs and with experience in probabilistic risk assessment (PRA) in the nuclear power industry. MSET offers significant potential benefits for improving nuclear safeguards and security in any nation with a nuclear program. MSET provides a design basis for developing an MC&A system at a nuclear facility that functions to protect against insider theft or diversion of nuclear materials. MSET analyzes the system and identifies several risk importance factors that show where sustainability is essential for optimal performance and where performance degradation has the greatest impact on total system risk. MSET contains five major components: (1) A functional model that shows how to design, build, implement, and operate a robust nuclear MC&A system (2) A fault tree of the operating MC&A system that adapts PRA methodology to analyze system effectiveness and give a relative risk of failure assessment of the system (3) A questionnaire used to document the facility's current MPC&A system (provides data to evaluate the quality of the system and the level of performance of each basic task performed throughout the material balance area [MBA]) (4) A formal process of applying expert judgment to convert the facility questionnaire data into numeric values representing the performance level of each basic event for use in the fault tree risk assessment calculations (5) PRA software that performs the fault tree risk assessment calculations and produces risk importance factor reports on the facility's MC&A (software widely used in the aerospace, chemical, and nuclear power industries) MSET was peer reviewed in 2007 and validated in 2008 by benchmark testing at the Idaho National Laboratory in the United States. The MSET documents were translated into Russian and provided to Rosatom in July of 2008, and MSET is currently being evaluated for potential application in Russian Nuclear Facilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Modeling of a spatially resolved ion temperature diagnostic for inertial confinement fusion

Here, the performance of modern laser-driven inertial confinement fusion (ICF) experiments is degraded by contamination of the deuterium–tritium (DT) fuel with high-Z material during compression. Simulations suggest that this mix can be described by the ion temperature distribution of the implosion, given that such contaminants deviate in temperature from the surrounding DT plasma. However, existing neutron time-of-flight (nTOF) diagnostics only measure the spatially integrated ion temperature. This paper describes the techniques and forward modeling used to develop a novel diagnostic imaging system to measure the spatially resolved ion temperature of an ICF implosion for the first time. The technique combines methods in neutron imaging and nTOF diagnostics to measure the ion temperature along one spatial dimension at yields currently achievable on the OMEGA laser. A detailed forward model of the source and imaging system was developed to guide instrument design. The model leverages neutron imaging reconstruction algorithms, radiation hydrodynamics and Monte Carlo simulations, optical ray tracing, and more. The results of the forward model agree with the data collected on OMEGA using the completed diagnostic. The analysis of the experimental data is still ongoing and will be discussed in a separate publication.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

X501 Experiment Simulation and PIE Data Comparison

Fuel performance simulations have been developed for X501 MA bearing fuel in BISON. The results of these simulations are compared with PIE results to assess the applicability of metallic fuel models in BISON to MA bearing metallic fuel.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Evaluation of used fuel disposition in clay-bearing rock.

The R&D program from the DOE Used Fuel Disposition Campaign (UFDC) has documented key advances in coupled Thermal-Hydrological-Mechanical-Chemical (THMC) modeling of clay to simulate its complex dynamic behavior in response to thermal and hydrochemical feedbacks. These efforts have been harnessed to assess the isolation performance of heatgenerating nuclear waste in a deep geological repository in clay/shale/argillaceous rock formations. This report describes the ongoing disposal R&D efforts on the advancement and refinement of coupled THMC process models, hydrothermal experiments on barrier clay interactions, used fuel and canister material degradation, thermodynamic database development, and reactive transport modeling of the near-field under non-isothermal conditions. These play an important role to the evaluation of sacrificial zones as part of the EBS exposure to thermallydriven chemical and transport processes. Thermal inducement of chemical interactions at EBS domains enhances mineral dissolution/precipitation but also generates mineralogical changes that result in mineral H2O uptake/removal (hydration/dehydration reactions). These processes can result in volume changes that can affect the interface / bulk phase porosities and the mechanical (stress) state of the bentonite barrier. Characterization studies on bentonite barrier samples from the FEBEX-DP international activity have provided important insight on clay barrier microstructures (e.g., microcracks) and interactions at EBS interfaces. Enhancements to the used fuel degradation model outlines the need to include the effects of canister corrosion due the strong influence of H 2 generation on the source term.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask

This report provides a best-estimate evaluation of residual water content (post-dry out) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask, and evaluates the radiolysis of the residual free water, and the physisorbed and chemisorbed waters on the surfaces of the fuel and cask internal contents. The evaluation of radiolytic breakdown of those waters with gamma radiation causing the generation of hydrogen gas (H 2 ) is made using available literature data and models. This evaluation is part of the overall materials performance evaluation of the SNF-in-canister system, and is part of the technical bases for their continued safe dry storage. The TN-32 cask contents included 32 HBU LWR spent fuel assemblies each with 264 fuel rods clad in zirconium alloys, aluminum neutron absorber components, and aluminum and stainless steel structural components. The residual free and surface (physisorbed/chemisorbed) waters are ascribed to water vapor in the free volume and to components’ surfaces, respectively. The total potential radiolytic hydrogen inventory from the water vapor and from waters ascribed to surfaces has been calculated assuming all the water produced molecular H 2 . The residual water that is chemically incorporated into the bulk of a hydrated oxide, i.e., chemisorbed water, and its total potential hydrogen inventory has been calculated. These calculations are at the physical limit of material available and are used for a bounding assessment purpose only.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1993

Integrated model development for liquid fueled rocket propulsion systems

As detailed in the original statement of work, the objective of phase two of this research effort was to develop a general framework for rocket engine performance prediction that integrates physical principles, a rigorous mathematical formalism, component level test data, system level test data, and theory-observation reconciliation. Specific phase two development tasks are defined.

Santi, L. Michael↗
DOE OSTI2023

Comparison Between Pin-by-Pin Subchannel and System Level Thermal Hydraulic Results for High Burnup Loss-of-Coolant Applications

This milestone report summarizes recent work to investigate higher fidelity modeling and simulation practices for large-break loss-of-coolant accident (LBLOCA) analysis in high-burnup pressurized water reactor (PWR) cores. Because of current industry interest in extending fuel cycle lengths ranging from 18 to 24-months, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has been investigating the susceptibility of high-burnup core designs to fuel fragmentation, release, and dispersal (FFRD) during accident conditions such as LBLOCA. This work has prioritized developing and demonstrating a methodology for calculating core-wide susceptibility to FFRD and addressing uncertainties identified in the analysis and the US Nuclear Regulatory Commission (NRC) Research Information Letter on FFRD. Part of this investigation seeks to quantify differences between system-level thermal hydraulic behavior and higher fidelity subchannel modeling methods to identify potential safety concerns or opportunities to minimize FFRD susceptibility. To this end, the NEAMS subchannel code, CTF, is being used, along with the NRC system analysis code TRACE, for analysis of LBLOCA in a core containing high-burnup fuel. This project includes two thrusts: (1) improving on the existing TRACE model for a 4-loop PWR for LBLOCA so that the model is higher fidelity and consistent with current USNRC best practices, and (2) using CTF to perform pin-resolved modeling of the core region of the reflood phase of a LBLOCA in a high-burnup PWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Separations and safeguards model integration.

Research and development of advanced reprocessing plant designs can greatly benefit from the development of a reprocessing plant model capable of transient solvent extraction chemistry. This type of model can be used to optimize the operations of a plant as well as the designs for safeguards, security, and safety. Previous work has integrated a transient solvent extraction simulation module, based on the Solvent Extraction Process Having Interaction Solutes (SEPHIS) code developed at Oak Ridge National Laboratory, with the Separations and Safeguards Performance Model (SSPM) developed at Sandia National Laboratory, as a first step toward creating a more versatile design and evaluation tool. The goal of this work was to strengthen the integration by linking more variables between the two codes. The results from this integrated model show expected operational performance through plant transients. Additionally, ORIGEN source term files were integrated into the SSPM to provide concentrations, radioactivity, neutron emission rate, and thermal power data for various spent fuels. This data was used to generate measurement blocks that can determine the radioactivity, neutron emission rate, or thermal power of any stream or vessel in the plant model. This work examined how the code could be expanded to integrate other separation steps and benchmark the results to other data. Recommendations for future work will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2016

Comparison of Fuel Consumption of Descent Trajectories Under Arrival Metering

This paper compares fuel consumption of descent trajectories from cruise altitude to meter fix when the required time of arrival is later than the nominal time of arrival at the meter fix. The required delay, which is the difference between the nominal and the required times of arrival, is achieved by either slowing down the aircraft in the cruise and descent phases or flying a longer route at a constant altitude. Performance models of ten different Boeing and Airbus aircraft, obtained from the Base of Aircraft Data, are employed for generating the results. It is demonstrated that the most fuel-efficient speed control strategy for absorbing delay is first reducing descent speed as much as possible and then reducing cruise speed. This is a common finding for all ten aircraft considered. For some aircraft, flying at a fixed flight path angle and constant Mach-calibrated-airspeed results in lower fuel consumption compared to standard descent at idle-thrust and constant Mach-calibrated- airspeed. Finally, for the cases examined, it is shown that executing a path stretch maneuver at cruise altitude and descent at a reduced speed is more fuel efficient than inserting an intermediate-altitude cruise segment.

Nikoleris, Tasos↗