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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.

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At least 1,657 records · Page 92
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Unified Creep Plasticity Damage (UCPD) Model for Rigid Polyurethane Foams

Numerous experiments were performed to characterize the mechanical response of several different rigid polyurethane foams (FR3712, PMDI10, PMDI20, and TufFoam35) to large deformation. In these experiments, the effects of load path, loading rate, and temperature were investigated. Results from these experiments indicated that rigid polyurethane foams exhibit significant volumetric and deviatoric plasticity when they are compressed. Rigid polyurethane foams were also found to be very strain-rate and temperature dependent. These foams are also rather brittle and crack when loaded to small strains in tension or to larger strains in compression. Thus, a new Unified Creep Plasticity Damage (UCPD) model was developed and implemented into SIERRA with the name Foam Damage to describe the mechanical response of these foams to large deformation at a variety of temperatures and strain rates. This report includes a description of recent experiments and experimental findings. Next, development of a UCPD model for rigid, polyurethane foams is described. Selection of material parameters for a variety of rigid polyurethane foams is then discussed and finite element simulations with the new UCPD model are compared with experimental results to show behavior that can be captured with this model.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Sample EP Flow Analysis of Severely Damaged Networks

These are slides for a presentation at the working group meeting of the WESC SREMP Software Product Integration Team on sample EP flow analysis of severely damaged networks. The following topics are covered: ERCOT EP Transmission Model; Zoomed in to Houston and Overlaying StreetAtlas; EMPACT Solve/Dispatch/Shedding Options; QACS BaseCase Power Flow Solution; 3 Substation Contingency; Gen. & Load/100 Optimal Dispatch; Dispatch Results; Shed Load for Low V; Network Damage Summary; Estimated Service Areas (Potential); Estimated Outage Areas (potential).

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Concrete Materials with Ultra-High Damage Resistance and Self- Sensing Capacity for Extended Nuclear Fuel Storage Systems

The objective of this project is to develop a new class of multifunctional concrete materials (MSCs) for extended spent nuclear fuel (SNF) storage systems, which combine ultra-high damage resistance through strain-hardening behavior with distributed multi-dimensional damage self-sensing capacity. The beauty of multifunctional concrete materials is two-fold: First, it serves as a major material component for the SNF pool, dry cask shielding and foundation pad with greatly improved resistance to cracking, reinforcement corrosion, and other common deterioration mechanisms under service conditions, and prevention from fracture failure under extreme events (e.g. impact, earthquake). This will be achieved by designing multiple levels of protection mechanisms into the material (i.e., ultrahigh ductility that provides thousands of times greater fracture energy than concrete and normal fiber reinforced concrete; intrinsic cracking control, electrochemical properties modification, reduced chemical and radionuclide transport properties, and crack-healing properties). Second, it offers capacity for distributed and direct sensing of cracking, strain, and corrosion wherever the material is located. This will be achieved by establishing the changes in electrical properties due to mechanical and electrochemical stimulus. The project will combine nano-, micro- and composite technologies, computational mechanics, durability characterization, and structural health monitoring methods, to realize new MSCs for very long-term (greater than 120 years) SNF storage systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Concrete Materials with Ultra-High Damage Resistance and Self- Sensing Capacity for Extended Nuclear Fuel Storage Systems

The objective of this project is to develop a new class of multifunctional concrete materials (MSCs) for extended spent nuclear fuel (SNF) storage systems, which combine ultra-high damage resistance through strain-hardening behavior with distributed multi-dimensional damage self-sensing capacity. The beauty of multifunctional concrete materials is two-fold: First, it serves as a major material component for the SNF pool, dry cask shielding and foundation pad with greatly improved resistance to cracking, reinforcement corrosion, and other common deterioration mechanisms under service conditions, and prevention from fracture failure under extreme events (e.g. impact, earthquake). This will be achieved by designing multiple levels of protection mechanisms into the material (i.e., ultrahigh ductility that provides thousands of times greater fracture energy than concrete and normal fiber reinforced concrete; intrinsic cracking control, electrochemical properties modification, reduced chemical and radionuclide transport properties, and crack-healing properties). Second, it offers capacity for distributed and direct sensing of cracking, strain, and corrosion wherever the material is located. This will be achieved by establishing the changes in electrical properties due to mechanical and electrochemical stimulus. The project will combine nano-, micro- and composite technologies, computational mechanics, durability characterization, and structural health monitoring methods, to realize new MSCs for very long-term (greater than 120 years) SNF storage systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

The Finite Strain Johnson Cook Plasticity and Damage Constitutive Model in ALEGRA.

A finite strain formulation of the Johnson Cook plasticity and damage model and it's numerical implementation into the ALEGRA code is presented. The goal of this work is to improve the predictive material failure capability of the Johnson Cook model. The new implementation consists of a coupling of damage and the stored elastic energy as well as the minimum failure strain criteria for spall included in the original model development. This effort establishes the necessary foundation for a thermodynamically consistent and complete continuum solid material model, for which all intensive properties derive from a common energy. The motivation for developing such a model is to improve upon ALEGRA's present combined model framework. Several applications of the new Johnson Cook implementation are presented. Deformation driven loading paths demonstrate the basic features of the new model formulation. Use of the model produces good comparisons with experimental Taylor impact data. Localized deformation leading to fragmentation is produced for expanding ring and exploding cylinder applications.

42 ENGINEERING↗
DOE OSTI2018

Glass Striae and Laser Shock Damage

This study sought to examine if striae are preferential locations for damage initiation in vitreous silicate float glasses when they are subjected to shock. Striae are typically native to vitreous silicate float glasses and their formation is a consequence of incomplete or inhomogeneous mixing prior to glass solidification. Striae typically have a lamellar-like habit and their planar orientation is usually perpendicular to the glass tile thickness. F loat glass, such as BOROFLOATâ borosilicate glass, is often used in transparent protective systems and therefore can be subjected to shock in a ballistic impact. Previous work by the Authors indicated that shock-induced damage in such glasses could be co-located with striae, so the objective of this study was to more closely examine that potential relationship.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Basic Mechanisms: Displacement Damage [Slides]

This report contains a set of slides intended for the 2019 Los Alamos National Laboratory (LANL) radiation effects summer school. On the topic of the basic mechanisms of displacement damage, this presentation discusses the consequences, physical mechanisms, electrical degradation, modeling displacement damage, and current research.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Linking the Codependence of Grain Boundary Structure and Density to Defect Evolution Mechanisms during Radiation Damage

There is little understanding of the direct link between grain boundary character and density, and defect denuded zones. For example, the role of reduced grain size to the nanocrystalline regime and subsequent increase in grain boundary density is not fully understood in the context of dislocation or void behavior near grain boundaries. In addition, little is understood experimentally on how defect absorption changes the structure of the grain boundary during irradiation and may play a role in the migration of defects and solutes. This work aims to define a framework for understanding the origin of radiation damage mechanisms in model BCC and FCC nanocrystalline materials, on which a foundation of radiation tolerant material development can be built. The key questions that will be addressed by this work are: 1. What is the influence of GB density and GB character (GBC) on radiation damage accumulation near grain boundaries in model nanocrystalline FCC and BCC alloys; how are these two effects connected? 2. In the context of radiation induced segregation (RIS), what effect does species dependent interstitial diffusion have on the GB chemistry in light of these two dependencies? 3. How does grain boundary sink efficiency evolve with radiation over time? In situ and analytical transmission electron microscopy techniques are used to study both pure and alloy nanocrystalline materials under irradiation. TEM samples prepared from thin films will be irradiated in-situ, allowing for dynamic observation of the behavior of the alloy under ion irradiation. The key findings from this program were (1) the control of thin film texture via sputtering; (2) the development of characterization tools for defects and grain boundary quantification; (3) the development of new techniques for defect characterization toward high throughput analysis; (4) a mapped out parameter space for defect evolution as a function of grain size and grain boundary character; and finally (5) an indication of sink efficiency and its stability during irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Pool Fire Analysis Methodology for Assessing Damage to Waste Containers

This document provides guidelines, whereby the physical parameters associated with postulated flammable/combustible liquid pool fires are determined (thermal analysis) in order to estimate the number of waste containers impacted. Pool fire damage to waste containers may occur due to a container being fully engulfed (within the periphery of the pool), partially engulfed, or within the critical heat flux distance. The methodology described here for estimating the size of the pool fire is developed using spill characteristics outlined in Chapter 65 of the SFPE Handbook (Ref. 7) and testing done by Hansen (Ref. 25). For exposures, the critical flux height is determined using the detailed method of Shokri and Beyler (Ref. 8) from the flame height, calculated as described by Heskestad (Ref. 7) and the average black body emissive power, determined using the method of Shokri and Beyler (Ref. 8). Three types of liquid pool fires should be considered; unconfined crash/instantaneous spill fires, unconfined continuously flowing (also called metered leak) spill fires, and confined spill fires. All fuel sources should be considered consistent with the facility Fire Hazard Analysis (FHA). However, the most likely liquid fuel sources present in TRU waste facilities are those present on TRU waste handling vehicles (forklifts, pallet movers, transport trucks, etc.). This method can also be used for assessing damage to items other than waste containers or other methods of analysis may be used with appropriate justification.

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

Potential Formation Damage: An Integrated Reservoir Characterization Study of the Naturally Fractured Carbonate Middle Duperow Formation at the Kevin Dome, Montana

In this study, we integrate geologic and engineering data of a naturally fractured carbonate reservoir at the Kevin Dome, Montana. Well test data are correlated with core description, geochemical and lithology study to determine the flow behavior and communication within the injection test interval and to the surrounding area. Based on a dual-continuum geologic model, numerical brine injection simulations are carried out to validate the interpretation results from our well test analytical models and forecast the probability of CO 2 injection success using current reservoir properties. As a result, our well test analytical models as well as lithology/core description suggest that fluid flow may be mainly restricted to the injection interval and the assumption of radial (horizontal) flow may be appropriate. The well test models also indicate that there is potentially formation damage with a positive skin factor although prior to brine injection well tests, well stimulation through acid treatment was performed. Our numerical simulation results appear to confirm this formation damage by showing additional pressure buildup in the injection data during later test periods. To explain this, acid may have dissolved dolomite then dolomite or calcite may have been formed again further into the matrix/fracture system. Another possible explanation is mechanical clogging of the fractures due to acid dissolving dolomite and dislodging fine grains. Our work also predicts that if no additional well stimulation is performed, the project will have a lower probability of successfully injecting 1 million tons of CO 2 into the Middle Duperow formation over 4 years.

42 ENGINEERING↗
DOE OSTI2021

Physics-Infused AI/ML Based Digital-Twin Framework for Flow-Induced-Vibration Damage Prediction in a Nuclear Reactor Heat Exchanger

This report summarizes some of the ongoing work related to the development of an expert-elicitation-digital-twin framework for real time damage state prediction in heat exchanger components of a nuclear reactor. The framework is targeted towards predicting damage associated with coupled low cycle fatigue (associated with regular heat-up, cool-down and power operation transients) and high cycle fatigue (associated with flow induced vibration transients). The overall framework will be based on a NoSQL based database, physics-infused-geometry-dependent virtual-sensor data, different AI/ML techniques-based data-driven-predictive-model applications (Apps) and real-time plant sensor measurements available through few existing sensors. Towards this overall goal, this report updates some of the ongoing work, such as on implementation of a NoSQL Database (such as MongoDB), FE based heat transfer analysis of a heat exchanger (e.g. of a PWR steam generator) for generating geometry-dependent virtual sensor data and evaluation of various AI/ML models such as based on multivariate linear regression, ensembled decision-tree based Random-Forest and Gradient-Boosting regression and high-dimensional-kernel-function-transformation based Support-Vector-Machine regression models. The AI/ML models were evaluated for predicting multi-time-series thermal states at thousands of 3D point-clouds

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Designing Damage-Resistant Multilayer Dielectric Gratings for Petawatt-Class Lasers

We have successfully developed high efficiency dielectric gratings for chirped pulse amplification (CPA) pulse compression with a focus on improving damage thresholds for high peak power. Specifically, we focused on developing first-of-kind designs that operate at TM polarization. Unpublished modeling within our group shows that the electric field enhancement in the solid material for gratings operating at TM can be significantly less than for TE. Subsequently, the laser damage threshold of TM gratings should be higher by potentially a factor of 2X.

42 ENGINEERING↗
DOE OSTI2022

Damage Accumulations Predictions for Boiler Components Via Microstructurally Informed Material Models

The goal of the project was to model material behavior and degradation during cyclic plasticity— with and without hold time—for nickel-based superalloys used in USC (ultra-super-critical) and A-USC (advanced-ultra-super-critical) boiler components. The study provided physically informed models, capturing the microstructural changes taking place in the industrial components under cyclic loading and long duration stress (up to 300,000 hours) and high temperature exposure (1100°F/593°C to 1400°F/760°C). The major developments were: 1) Qualitative and quantitative understanding of microstructure evolution (gamma prime precipitates), deformation (dislocation density), and damage mechanisms of Haynes ® 282 alloy. 2) Qualitative understanding of microstructural features generating local strain variations. 3) A continuum damage mechanics model (CDM) for Haynes ® 282 alloy at 1100°F to 1400°F capturing cyclic behavior with and without hold time. 4) Structural analysis for creep and LCF life predictions of an USC thick-wall Grade 91 superheater steel header and understanding life sensitivity to wall thickness of an AUSC Haynes ® 282 header.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2023

Control Carrier Board (CCB) Damage (ISS-95470.01)

CCBs installed in the first LRU32 preproduction units were found to not operate properly when received at LLNL from the integration vendor on 02/07/2023 during integrated testing of low voltage and high voltage as the first steps of validating a functional LRU for physics qualifications. CCB version dated 12/10/2021 was found to not accommodate a design change approved 09/28/2022 , which led to damage occurring during testing. The test stand holding pins would interfere with the location of a resistor due to CCB design modification and pressed on the resistor causing damage. Communication deficiency is identified as cause of issue.

42 ENGINEERING↗
DOE OSTI2023

MST-8 Colloquium: The Effect of Microstructure on the Spall Strength and Damage of 1045 Steel [Slides]

Three different microstructures of the same steel alloy were investigated to study the influence of microstructure on spall strength and damage. Microstructure influences spall behavior within the explored dynamic regime. Second phase morphology, interphase interface characteristics, and grain boundary character shown to influence spall behavior. Continued understanding of microstructure’s influence on spall and damage needed for targeted material design.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Phase-Field Modeling of Thermally-Grown Oxide and the Induced Damage Evolution in Environmental Barrier Coatings

The advent of next-generation hydrogen-based engines necessitates materials capable of withstanding temperatures beyond the reach of current superalloys. SiC-based ceramic matrix composites, augmented with environmental barrier coatings (EBCs), present a promising materials solution. Given the active search for effective and durable EBCs, there is a pressing need for modeling tools to understand and predict damage evolution in these materials to help accelerate their development. This study introduces a phase-field model (PFM) designed to simulate the critical role of thermally grown oxides (TGO) in the degradation and failure of EBCs. The model accounts for the severe volume expansion due to oxidation, alongside phase transformations and microstructural evolution during thermal cycling, offering a comprehensive view of the damage processes. Simulation results are validated against experimental findings reported in the literature, establishing the model's potential as a significant tool for understanding and improving the resilience of EBCs in cyclic oxidative environments.

Cheng, Tianle↗
DOE OSTI2025

Phase Field Modeling of Chemical Reaction Related Damage Evolution in Environmental Barrier Coatings

The advent of next-generation engines necessitates materials capable of withstanding temperatures beyond the reach of current superalloys. SiC-based ceramic matrix composites, augmented with environmental barrier coatings (EBCs), present a promising materials solution. Given the active search for effective and durable EBCs, there is a pressing need for modeling tools to understand and predict damage evolution in these materials to help accelerate their development. This study introduces a phase-field model (PFM) designed to simulate the thermally grown oxides (TGO) and phase transformation in the degradation and failure of EBCs. The model accounts for the severe volume expansion due to oxidation, alongside phase transformations and porosity evolution during thermal cycling, offering a comprehensive view of the damage processes. Simulation results are validated against experimental findings reported in the literature, establishing the model's potential as a significant tool for understanding and improving the resilience of EBCs in cyclic oxidative environments.

fast-diffusion path↗
DOE OSTI2025

Interpretation of Ion Irradiation and Neutron Irradiation Damage in Additively Manufactured 316 Stainless Steel using Multiscale Modeling

The accelerated adoption of nuclear energy necessitates advanced manufacturing technologies, such as additive manufacturing, to meet heightened supply chain requirements and support innovative reactor technologies. Due to the unique microstructural characteristics of additively manufactured materials under distinct solidification conditions, comprehensive evaluation of their performance in reactor environments is essential. The Advanced Materials and Manufacturing Technologies program under the Department of Energy's Office of Nuclear Energy focuses on understanding the irradiation performance and damage evolution of laser powder bed fusion 316 stainless steel, with an emphasis on integrating ion and neutron irradiation data to accelerate the development and qualification of materials for advanced nuclear reactor applications. While ion irradiation is a cost- and time-effective method, modeling and simulation are required to interpret the data for the broader range of irradiation conditions encountered in advanced reactors. In fiscal year 2025, integrated multiscale modeling and simulations were conducted to assess irradiation damage in additively manufactured 316 stainless steel. Key outcomes include predictions of chromium enrichment at grain boundaries, nickel enrichment at dislocation cell walls and void surfaces, and heterogeneous void evolution under ion and neutron irradiation conditions. Cluster dynamics simulations revealed the coarsening of voids at high irradiation temperatures and the suppression of void growth by high network dislocation density, while also demonstrating significant growth and coarsening of voids and self-interstitial atom loops at low dose rates. Machine learning-accelerated atomistic simulations highlighted the impact of the local environment and chromium concentration on vacancy diffusivity, providing key insights on the influence of composition on void swelling and radiation-induced segregation. Additionally, molecular dynamics simulations demonstrated the presence of defect production bias and a significant effect of carbon content on defect cluster behavior. These combined efforts aim to predict the performance of additively manufactured materials under various reactor conditions, supporting their qualification for nuclear reactor applications by interpreting ion irradiation data. This report underscores the potential of integrated multiscale modeling to analyze ion irradiation data in the effort to accelerate the qualification of additively manufactured materials for nuclear reactor components.

316 stainless steel↗