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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 1,711 records · Page 95
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Chemical Kinetic Models for Advanced Engine Combustion

The objectives for this project are as follows: Develop detailed chemical kinetic models for fuel components used in surrogate fuels for compression ignition (CI), homogeneous charge compression ignition (HCCI) and reactivity-controlled compression-ignition (RCCI) engines; and Combine component models into surrogate fuel models to represent real transportation fuels. Use them to model low-temperature combustion strategies in HCCI, RCCI, and CI engines that lead to low emissions and high efficiency.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2014

Dakota, a multilevel parallel object-oriented framework for design optimization, parameter estimation, uncertainty quantification, and sensitivity analysis version 6.0 theory manual

The Dakota (Design Analysis Kit for Optimization and Terascale Applications) toolkit provides a exible and extensible interface between simulation codes and iterative analysis methods. Dakota contains algorithms for optimization with gradient and nongradient-based methods; uncertainty quanti cation with sampling, reliability, and stochastic expansion methods; parameter estimation with nonlinear least squares methods; and sensitivity/variance analysis with design of experiments and parameter study methods. These capabilities may be used on their own or as components within advanced strategies such as surrogate-based optimization, mixed integer nonlinear programming, or optimization under uncertainty. By employing object-oriented design to implement abstractions of the key components required for iterative systems analyses, the Dakota toolkit provides a exible and extensible problem-solving environment for design and performance analysis of computational models on high performance computers. This report serves as a theoretical manual for selected algorithms implemented within the Dakota software. It is not intended as a comprehensive theoretical treatment, since a number of existing texts cover general optimization theory, statistical analysis, and other introductory topics. Rather, this manual is intended to summarize a set of Dakota-related research publications in the areas of surrogate-based optimization, uncertainty quanti cation, and optimization under uncertainty that provide the foundation for many of Dakota's iterative analysis capabilities.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Chemical Reactivity and Recommended Remediation Strategy for Los Alamos Remediated Nitrate Salt (RNS) Wastes

Nitrate salt wastes were generated during historical processing operations to recover and purify plutonium at Los Alamos National Laboratory (LANL). These nitrate salts have the potential to behave as oxidizers, and to contain entrained liquids, both of which are prohibited in the Waste Isolation Pilot Plant (WIPP). A process was developed to remediate nitrate salts by mixing salts with an inorganic zeolite pet litter to absorb free liquids, and remove the oxidizing characteristic of the wastes. Process changes introduced instead an organic wheat-­-based pet litter along with organic neutralizers, generating a new waste form referred to as Remediated Nitrate Salt (RNS) waste. The February 2014 radiological release at the WIPP site was traced to a LANL RNS waste drum number 68660. Experimental and modeling studies performed at LANL indicate that mixtures of metal nitrate salts (oxidizer) with Swheat organic kitty litter (fuel) create the potential for exothermic chemical reactions. Even in the presence of triethanol-­-ammonium nitrate (TEAN) created during neutralization, relatively high temperatures, greater than approximately 160 °C (320 °F), are required for runaway exothermic reactions. Evaluation of Drum 68660 characteristics coupled with extensive chemical testing indicate that, in addition to the nitrate salt/Swheat organic kitty litter mixture, an additional trigger mechanism (or mechanisms) is likely required to raise the internal drum temperature high enough to initiate the nitrate salt/Swheat organic kitty litter reaction. A combination of chemical conditions were identified that may lower the temperature for reaction, including initial high acid concentration of free liquids; significant quantities (> 1 gal) of neutralized, absorbed free liquids; the presence of reactive or catalytic metals like magnesium, iron, or lead; the presence of bismuth containing glovebox gloves; and the presence of natural biological activity. Complex surrogate nitrate salt mixtures prepared to simulate wastes, particularly those containing iron and magnesium, can generate NOx gases that partially nitrate the organic Swheat kitty litter and form a more energetic fuel. These complex surrogate salt mixtures display exothermic behavior at temperatures as low as 60 °C (140 °F) which is still well above the ambient temperature conditions experience by a drum. Neutralization of free liquids, and sorption onto Swheat establishes conditions (moisture) with near-neutral pH) that will support natural biological activity. Spontaneous self-heating generated by low-level chemical reactions and/or the respiration of bacteria, molds, and microorganisms is potentially important in the early stages and may be sufficient to raise the temperature as high as 60 °C (140 °F), where the other exothermic chemical reactions can take place. Additional studies are being conducted to evaluate the role biological activity may have played in initiating the event. The use of Swheat absorbent in the processing of nitrate salt wastes can be pinpointed as the critical processing decision that led to the failure of drum 68660, regardless of the ii details of the thermal processes that enabled the drum to achieve temperatures sufficient to initiate the chemical reactions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Technetium (VII) Co-precipitation with Framework Aluminosilicates

Technetium-99 (99Tc), a long-lived radionuclide, is one of the most widespread contaminants within the Hanford subsurface. At some depths, it is only extractable with strong acids, suggesting incorporation into a solid phase. We hypothesized that Tc may have coprecipitated with feldspathoid aluminosilicates under waste tanks that had leaked caustic solutions into the vadose zone. Our objectives were to determine if Tc could be incorporated into the feldspathoids cancrinite and sodalite and under what conditions coprecipitation could occur. Our hypothesis was that sodalite was more likely to incorporate and retain Tc. Our approach was to use known methods of feldspathoid formation in solutions resembling those in Hanford waste tanks contacting sediments in terms of major ion (Na, NO3, OH, Al(OH)4, and Si(OH)4 concentrations. In some cases, Al and Si were supplied from zeolite. We used perrhenate (ReO4) as a surrogate for pertechnetate (TcO4) to avoid the radioactivity. The major findings of this study were 1) ReO4 could be incorporated into either sodalite or cancrinite but the concentration in the solid was < 1% of the competing ion Cl, NO3, or NO2. 2) The small amount of ReO4 incorporated was not exchangeable with NO3 or NO2. 3) In sodalite, NO3 was highly preferred over ReO4 but significant Re-sodalite was formed when the mole fraction in solution (Re/Re+N) exceeded 0.8. 4) A nonlinear relation between the unit cell parameter and amount of Re incorporated suggested that a separate Re-sodalite phase was formed rather than a solid solution. 5) We determined that sodalite preference for sodalite in the presence of different anions increased with the ionic size of the competing anion: Cl < CO3 < NO3 < SO4 < MnO4 < WO4 and significant incorporation did not occur unless the difference in anion radii was less than 12%. 6) Re(VII) was not significantly reduced to Re(IV) under the conditions of this experiment and Re appeared to be a good surrogate for Tc under oxidizing conditions as both oxyanions have the same size. In conclusion, incorporation of TcO4 is unlikely in the close proximity to the tanks where sodalite is likely to form in the presence of high NO3 and other anions much smaller than TcO4.

58 GEOSCIENCES↗
DOE OSTI2016

Numerical Estimation of the Spent Fuel Ratio

Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Summary Report of Laboratory Testing to Establish the Effectiveness of Proposed Treatment Methods for Unremediated and Remediated Nitrate Salt Waste Streams

The inadvertent creation of transuranic waste carrying hazardous waste codes D001 and D002 requires the treatment of the material to eliminate the hazardous characteristics and allow its eventual shipment and disposal at the Waste Isolation Pilot Plant (WIPP). This report documents the effectiveness of two treatment methods proposed to stabilize both the unremediated and remediated nitrate salt waste streams (UNS and RNS, respectively). The two technologies include the addition of zeolite (with and without the addition of water as a processing aid) and cementation. Surrogates were developed to evaluate both the solid and liquid fractions expected from parent waste containers, and both the solid and liquid fractions were tested. Both technologies are shown to be effective at eliminating the characteristic of ignitability (D001), and the addition of zeolite was determined to be effective at eliminating corrosivity (D002), with the preferred option1 of zeolite addition currently planned for implementation at the Waste Characterization, Reduction, and Repackaging Facility. During the course of this work, we established the need to evaluate and demonstrate the effectiveness of the proposed remedy for debris material, if required. The evaluation determined that Wypalls absorbed with saturated nitrate salt solutions exhibit the ignitability characteristic (all other expected debris is not classified as ignitable). Follow-on studies will be developed to demonstrate the effectiveness of stabilization for ignitable Wypall debris. Finally, liquid surrogates containing saturated nitrate salts did not exhibit the characteristic of ignitability in their pure form (those neutralized with Kolorsafe and mixed with sWheat did exhibit D001). As a result, additional nitrate salt solutions (those exhibiting the oxidizer characteristic) will be tested to demonstrate the effectiveness of the remedy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Spent Fuel Ratio Estimates from Numerical Models in ALE3D

Potential threat of intentional sabotage of spent nuclear fuel storage facilities is of significant importance to national security. Paramount is the study of focused energy attacks on these materials and the potential release of aerosolized hazardous particulates into the environment. Depleted uranium oxide (DUO 2 ) is often chosen as a surrogate material for testing due to the unreasonable cost and safety demands for conducting full-scale tests with real spent nuclear fuel. To account for differences in mechanical response resulting in changes to particle distribution it is necessary to scale the DUO 2 results to get a proper measure for spent fuel. This is accomplished with the spent fuel ratio (SFR), the ratio of respirable aerosol mass released due to identical damage conditions between a spent fuel and a surrogate material like depleted uranium oxide (DUO 2 ). A very limited number of full-scale experiments have been carried out to capture this data, and the oft-questioned validity of the results typically leads to overly-conservative risk estimates. In the present work, the ALE3D hydrocode is used to simulate DUO 2 and spent nuclear fuel pellets impacted by metal jets. The results demonstrate an alternative approach to estimate the respirable release fraction of fragmented nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Density of Plutonium Turnings Generated from Machining Activities

The purpose of this project was to determine the density of plutonium (Pu) turnings generated from the range of machining activities, using both surrogate material and machined Pu turnings. Verify that 500 grams (g) of plutonium will fit in a one quart container using a surrogate equivalent volume and that 100 grams of Pu will fit in a one quart Savy container.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Unreviewed Disposal Question Evaluation: Waste Disposal in Engineered Trenches 3 and 4

Revision 0 of this UDQE addressed the proposal to place Engineered Trench #3 (ET#3) in the footprint designated for Slit Trench #12 (ST#12) and operate using ST#12 disposal limits. Similarly, Revision 1 evaluates whether ET#4 can be located in and operated to Slit Trench #13 (ST#13) disposal limits. Both evaluations conclude that the proposed operations result in an acceptably small risk of exceeding a SOF of 1.0 and approve these actions from a performance assessment (PA) perspective. Because ET#3 will be placed in the location previously designated for ST#12, Solid Waste Management (SWM) requested that the Savannah River National Laboratory (SRNL) determine if the ST#12 limits could be employed as surrogate disposal limits for ET#3 operations. SRNL documented in this Unreviewed Disposal Question Evaluation (UDQE) that the use of ST#12 limits as surrogates for the new ET#3 disposal unit will provide reasonable assurance that Department of Energy (DOE) 435.1 performance objectives and measures (USDOE, 1999) will be protected. Therefore, new ET#3 inventory limits as determined by a Special Analysis (SA) are not required.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Aluminum Clad Spent Nuclear Fuel Task 4: Performance of Aluminum SNF in Dry Storage Experiment Test Plan

This document presents the activities to be performed in support of understanding the performance of aluminum spent nuclear fuel (ASNF) in dry storage. These activities will visually examine and acquire baseline samples of as-irradiated ATR fuel assembly components to determine fuel condition and the morphology of the aluminum oxyhydroxides present on the fuel external surfaces. Because the boehmite phase of oxyhydroxide is projected to be a significant contributor to radiolytic hydrogen production and general corrosion, being able to quantify its presence is of major importance to long term storage of aluminum spent nuclear fuel. Baseline samples will be taken from irradiated end adapters, removed from ATR SNF elements, stored in the ATR Canal. These will be transferred to Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) to perform non-destructive and destructive testing to determine the degree of hydration of the wet-stored aluminum components prior to drying and air storage. Examination of the dry stored SNF at Idaho Nuclear Technology and Engineering Center (INTEC) in CPP-603 Irradiated Fuel Storage Facility (IFSF) will allow the opportunity to identify gross degradation and as conditions permit, sample the surface oxide on fueled and un-fueled portions of the stored assemblies. The information resulting from analysis of the ATR Canal samples will be used to support validation of surrogate samples prepared under Task 6, sample surrogate preparation, and used in Task 1, oxyhydroxide layer behavior and chemistry, Task 2, oxide layer radiolytic gas generation, and Task 5, oxide layer response to drying. Surface oxide samples collected from SNF in IFSF dry storage will SNF dry storage prediction modeling from Task 3, sealed and vented system episodic breathing and gas generation prediction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Fundamentals of Pellet-Clad Debonding

This project focused on providing a fundamental mechanistic understanding of the complex degradation mechanisms associated with Pellet/Clad Debonding (PCD) through the use of a unique suite of novel synthesis of surrogate spent nuclear fuel, in-situ nanoscale experiments on surrogate interfaces, multi-modeling, and characterization of decommissioned commercial spent fuel. The understanding of a broad class of metal/ceramic interfaces degradation studied within this project provided the technical basis related to the safety of high burn-up fuel, a problem of interest to the DOE.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Survey of DAKOTA's V&V Capabilities in the Simulation of Residual Stresses in a Simple Composite Structure

Process-induced residual stresses occur in composite structures composed of dissimilar materials. As these residual stresses can result in fracture, their consideration when designing composite parts is necessary. However, the experimental determination of residual stresses in prototype parts can be time and cost prohibitive. Alternatively, it is possible for computational tools to predict potential residual stresses. Therefore, a process modeling methodology was developed and implemented into Sandia National Laboratories' SIERRA/SolidMechanics code. This method can be used to predict the process-induced stresses in any composite structure, regardless of material composition or geometric complexity. However, to develop confidence in these predictions, they must be rigorously validated. Specifically, sensitivity studies should be completed to define which model parameters are critical to the residual stress predictions. Then, the uncertainty associated with those critical parameters should be quantified and processed through the model to develop stress-state predictions encompassing the most important sources of physical variability. Numerous sensitivity analysis and uncertainty quantification methods exist, each offering specific strengths and weaknesses. Therefore, the objective of this study is to compare the performance of several accepted sensitivity analysis and uncertainty quantification methods during the manufacturing process simulation of a composite structure. The examined methods include simple sampling techniques as well as more sophisticated surrogate approaches. The computational costs are assessed for each of the examined methods, and the results of the study indicate that the surrogate approaches are the most computationally efficient validation methods and are ideal for future residual stress investigations.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Solidified Organics Testing

The Radiochemical Engineering Development Center currently has plutonium-containing liquid organic waste that must be solidified before final disposal as solid debris waste. The documented safety analysis for the Radiochemical Engineering Development Center limits the amount of plutonium that can be released from solid debris waste forms during a fire scenario, which is less than 10 -4 airborne release fraction. Experimental results for plutonium release fractions from burning various materials has been compiled by the Department of Energy [DOE, 1994]. Combustion of organic liquids gave an average airborne release fraction of 6 × 10 -3 and combustion of surface-contaminated solids (paper, plastic, etc.) gave an average airborne release fraction of 5 × 10 -4 . Burning of 30% tri- n -butyl phosphate (TBP) in kerosene gave cerium release fractions of 5.6–7.7 × 10 -3 [Mishima and Schwendiman, 1973]. Earlier testing had evaluated many different grout-based formulations to solidify plutonium-containing organic waste liquids, and the ones used in these tests were shown to produce monolithic, fairly homogeneous solids. Each of the waste forms tested would meet the waste acceptance criteria for disposal at the Waste Isolation Pilot Plant as transuranic waste, but there were questions about whether they would meet the plutonium release criteria during a fire. The purpose of the work described here was to quantify the airborne release fraction for cerium, as a surrogate for plutonium, in these proposed waste forms. Cerium is often used as a simulant for plutonium in bench-scale testing, and cerium nitrate hexahydrate [Ce (NO 3 ) 3 ·6H 2 O] was added to the organic solutions that were solidified to prepare the surrogate waste forms for these tests. The waste forms were subjected to the designated fire scenario of 600ºC for 1 hour, and any cerium in the off-gas was trapped using a quartz filter followed by a scrubber containing 0.1 M NaOH, which is the same as used by Mishima and Schwendiman, [1973].

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Radiation Effects on High Thermal Conductivity Fuels (Final Scientific Report)

High thermal conductivity nuclear fuels offer important potential advantages over traditional oxide-based fuels such as higher burnup, reduction in fission gas release, and better overall safety of the reactor system. One interesting approach to high thermal conductivity fuels utilizes high thermal conductivity nonfissile additives to UO 2 fuel to lower the fuel operating temperature and thereby take advantage of the highly favorable radiation resistance of UO 2 at lower operating temperatures. However, differential swelling between the matrix and high conductivity additive phases during high dose irradiation could lead to internal cracking and poor performance. In the current study, ceria (CeO 2 ) and zirconia (ZrO 2 ) surrogate matrices were used to model UO 2 behavior. Al 2 O 3 or SiC in the form of nanoscale powder or platelets, respectively, were used for the high conductivity second phase. The nuclear fuel surrogates were sintered to achieve densities greater than 93% of the ideal values. Scanning electron microscopy (SEM) imaging and X-ray diffraction confirmed the uniform distribution of the second phase and that no intermetallic second phase was formed during sintering. The thermal conductivity of the sintered samples was measured from 50°C to 900°C and revealed an important increase compared to pure CeO2 and ZrO2 pellets. Samples were irradiated with 20 MeV Ni6+ ions at midrange doses ranging from 1 to 15 displacements per atom (dpa) and temperatures from 300°C to 700°C. Post irradiation characterization revealed a good stability of the samples at low to medium doses (1 to 5 dpa) but showed a significant microstructural deterioration and decrease of the mechanical properties at 15 dpa. Although microcracking is not observed at the interface between the matrix and high conductivity phase for any investigated irradiation condition, the strength degradation at 15 dpa suggests that these high thermal conductivity fuel forms may not be suitable for high burnup applications in current or proposed fission reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Data Analysis of ENSA/DOE Rail Cask Tests

This report describes the results from a series of tests of surrogate pressurized water reactor (PWR) nuclear fuel assemblies in a rail cask during various modes of transportation and cask handling conducted between June and October 2017. The primary purpose of the tests was to measure strain and acceleration on surrogate fuel rods when the assemblies are subjected to normal conditions of transport (NCT) within the Equipos Nucleares, S.A. (ENSA) UNiversal (ENUN) 32P cask. Acceleration on the cask basket, the cask, the cask cradle, and the transport platforms were also measured. A summary of the test details, logistics and operations for performing the tests is included.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

LANL Phase 3 Test Plan For Container Loading to Support the Fire-Induced Pressure Response and Failure Characterization of PCV/SCV/3013 Containers

Fire testing of 3013 containers loaded with surrogate materials and water will be conducted to determine the response of 3013 containers stored in K-Area Complex (KAC) to a facility fire. Based on the proposal for the test configuration and loading, the objectives of the fire test are as follows: 1. “To increase understanding about the pressure and temperature conditions at which failure occurs; 2. To determine the allowable moisture for which failure will not occur under specified fire conditions, and 3. To increase understanding regarding the effects of salts on the pressurization, corrosion, and potential failure of the container.” Surrogate materials will be prepared at Los Alamos National Laboratory (LANL) and loaded in the 3013 containers. The containers will then be welded at LANL and shipped to Savannah River Site (SRS) for the installation of fittings and leak testing.

42 ENGINEERING↗
DOE OSTI2019

Final Report on the Viability of Acoustic Techniques for Density and Mass Flow in Enrichment Plants

The primary purpose of this research is to study the viability of acoustic signatures and sensors that could support accurate, noninvasive, and unattended measurement of uranium hexafluoride (UF6) gas density and mass flow rate in situations representative of gaseous centrifuge enrichment plants (GCEP) under safeguards by the International Atomic Energy Agency. The measurement method under development can be readily retrofitted to the exterior of piping in existing systems, potentially facilitating acceptance by facility operators. The findings from the demonstration of a preliminary prototype design and first-generation analysis algorithms in this project will inform the safeguards community as to whether continued acoustic instrument and methods development are warranted. This project conducted two types of measurements, depending on the flow conditions of the gas. For static measurements, the focus is on understanding the acoustic properties at low pressure for non-flowing UF6. For flow measurements, the focus is on understanding the performance of the acoustic system in a simulation of a GCEP pipe using a surrogate gas under fully developed flow conditions. Testing with flowing UF6 was not considered because the cost of building a flowing UF6 system with flow similar to a GCEP was considerably greater than allocated project resources. This document summarizes the results of research conducted over the course of the project. Based on the results to date, the approach is seen to be promising with non-invasive acoustic measurements shown to be possible at the desired pressures. The project has demonstrated • 5% uncertainty of the UF6 gas density for static measurements at pressures relevant to GCEP operations over a 5-minute measurement window with a clear path forward to achieving considerably lower uncertainties over a two-hour period, • Measurements of flowing surrogate gases at pressures relevant to GCEP operations. However, additional research remains to be done to further develop the measurement approach and transition the technology for field use. These development needs include: • Understand the impact of flowing gas on gas density measurements, • Adding sensors and improving the measurement procedure, • Augmenting the data analysis methods by leveraging machine learning approaches, • Reevaluating uncertainty requirements for density and mass flow given observed Online Enrichment Monitor (OLEM) measurement uncertainties, • Designing a robust instrumentation package for field-testing, and • Evaluating measurement results to understand the potential impact on mass balance calculations within GCEP facilities.

42 ENGINEERING↗
DOE OSTI2020

Dakota A Multilevel Parallel Object-Oriented Framework for Design Optimization Parameter Estimation Uncertainty Quantification and Sensitivity Analysis: Version 6.12 Theory Manual

The Dakota toolkit provides a flexible and extensible interface between simulation codes and iterative analysis methods. Dakota contains algorithms for optimization with gradient and nongradient-based methods; uncertainty quantification with sampling, reliability, and stochastic expansion methods; parameter estimation with nonlinear least squares methods; and sensitivity/variance analysis with design of experiments and parameter study methods. These capabilities may be used on their own or as components within advanced strategies such as surrogate-based optimization, mixed integer nonlinear programming, or optimization under uncertainty. By employing object-oriented design to implement abstractions of the key components required for iterative systems analyses, the Dakota toolkit provides a flexible and extensible problem-solving environment for design and performance analysis of computational models on high performance computers. This report serves as a theoretical manual for selected algorithms implemented within the Dakota software. It is not intended as a comprehensive theoretical treatment, since a number of existing texts cover general optimization theory, statistical analysis, and other introductory topics. Rather, this manual is intended to summarize a set of Dakota-related research publications in the areas of surrogate-based optimization, uncertainty quantification, and optimization under uncertainty that provide the foundation for many of Dakota's iterative analysis capabilities.

97 MATHEMATICS AND COMPUTING↗