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Anderson, Mark

Publications and source records attributed to Anderson, Mark.

At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Thermal Stratification Modeling for Sodium-Cooled Fast Reactors: A Status Update

Thermal hydraulic behavior in the upper plenum of pool- type sodium-cooled fast reactors (SFRs) is a major concern, as many design challenges are concentrated in this region. As SFR designs aim for licensing and commercialization, it is important to accurately analyze and predict the thermal-hydraulic behavior in this region during accident scenarios, specifically thermal stratification. Thermal stratification models are currently a major source of uncertainty in most system codes for all types of power plants. Most system codes, including SAS4A/SASSYS-1, a system level code developed by Argonne National Laboratory (Argonne), use very coarse meshes that cannot capture the complexities of the stratification phenomena. While the commonly employed lumped-volume based models for thermal stratification are able to run in a matter of seconds, they result in approximate results and can only handle simple cases. Other 2-D and 3-D methods, such as computational fluid dynamics (CFD) models, can analyze simple configurations with higher fidelity, but come with a relatively large computational expense. Finding a modeling solution that is both accurate and computationally efficient has proven difficult. This paper provides details of a review and gap analysis of the various modeling approaches proposed to date and explores a path forward for future thermal stratification modeling efforts, with a focus on developing new models for the SAS4A/SASSYS-1 system code.

Morgan, Sarah↗
NASA NTRS2018

Mitigation of Attentional Tunneling in the Flight Deck Using a Spatial Auditory Display

The role of attentional bias to a specific form of information, e.g., visual versus aural instructions, was evaluated to address the issue of attentional fixation (i.e., "cognitive tunneling") in the flight deck. A method of mitigating attentional tunneling using a spatial auditory display was evaluated in the presence and absence of a biasing factor. Data were gathered for participant recognition, accuracy and response times to aural and visual instructions under different conditions of distractor stimuli. Results showed a significant main effect for target acquisition with the use of spatial audio techniques. No significant main effects or interactions were found for target acquisition timing. Target acquisition was not affected by a "gaming score" biasing factor associated with the distractor tasks.

readback errors↗
DOE OSTI2018

Thermal Stratification Analysis for Sodium Fast Reactors

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool is desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.

Schneider, James↗
DOE OSTI2018

Thermal Stratification Analysis for Sodium Fast Reactors

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool is desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.

Schneider, James↗
DOE OSTI2018

Compact Heat Exchanger Design and Testing for Advanced Reactors and Advanced Power Cycles

The goal of the proposed research is to demonstrate the thermal hydraulic performance of innovative surface geometries in compact heat exchangers used as intermediate heat exchangers (IHXs) and recuperators for the supercritical carbon dioxide (s-CO 2 ) Brayton cycle. Printed-circuit heat exchangers (PCHEs) are the primary compact heat exchangers of interest. The overall objectives are: To develop optimized PCHE designs for different working fluid combinations including helium to s-CO 2 , liquid salt to s-CO 2 , sodium to s-CO 2 , and liquid salt to helium; To experimentally and numerically investigate thermal performance, thermal stress and failure mechanism of PCHEs under various transients; and To study diffusion bonding techniques for elevated-temperature alloys and examine post-test material integrity of the PCHEs. The project objectives were accomplished by defining and executing five different tasks corresponding to these specific objectives. The first task involved a thorough literature review and a selection of IHX candidates with different surface geometries as well as a summary of prototypic operational conditions. The second task involved optimization of PCHE design with numerical analyses of thermal-hydraulic performances and mechanical integrity. The subsequent task dealt with the development of testing facilities and engineering design of PCHE to be tested in s-CO 2 fluid conditions. The next task involved experimental investigation and validation of the thermal-hydraulic performances and thermal stress distribution of prototype PCHEs manufactured with particular surface geometries. The last task involved an investigation of diffusion bonding process and posttest destructive testing to validate mechanical design methods adopted in the design process. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed s-CO 2 test loop (STL) facility and s-CO 2 test facility at University of Wisconsin – Madison (UW).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Supercritical Brayton Power Conversion with a Direct Cooled Reactor for Space Power

To achieve future goals in space exploration a long life, compact power system will be necessary, and nuclear power sources are a promising option. The goal of this project is to achieve mass optimization with a supercritical Brayton Cycle coupled to a direct-cooled nuclear reactor. It is critical to minimize the total mass of this system because of space launch costs. The total mass involves both the components’ masses as well as the fuel mass which is related to system efficiency. This paper discusses a project in which a detailed reactor model is integrated with a carefully constructed cycle model in order to simultaneously optimize the design of these two aspects of the system. Component models are discussed as well as an initial optimization study. A robust, simple recuperated Brayton cycle model is complete and is in the process of integration with a developing reactor model. Preliminary observations of the cycle model have led to the conclusion that there will be an optimum point between a large recuperator and small reactor with high system efficiency, and a large reactor and small recuperator with low recuperator mass for any given radiater size. Once the reactor model is integrated into the cycle optimization, the tradeoffs between the sizes of the radiator, recuperator, and reactor will be further investigated and an optimum system mass will be apparent.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Supercritical Brayton Power Conversion with a Direct Cooled Reactor for Space Power

To achieve future goals in space exploration a long life, compact power system will be necessary, and nuclear power sources are a promising option. The goal of this project is to achieve mass optimization with a supercritical Brayton Cycle coupled to a direct-cooled nuclear reactor. It is critical to minimize the total mass of this system because of space launch costs. The total mass involves both the components’ masses as well as the fuel mass which is related to system efficiency. This paper discusses a project in which a detailed reactor model is integrated with a carefully constructed cycle model in order to simultaneously optimize the design of these two aspects of the system. Component models are discussed as well as an initial optimization study. A robust, simple recuperated Brayton cycle model is complete and is in the process of integration with a developing reactor model. Preliminary observations of the cycle model have led to the conclusion that there will be an optimum point between a large recuperator and small reactor with high system efficiency, and a large reactor and small recuperator with low recuperator mass for any given radiater size. Once the reactor model is integrated into the cycle optimization, the tradeoffs between the sizes of the radiator, recuperator, and reactor will be further investigated and an optimum system mass will be apparent.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Model validation using CFD-grade experimental database for NGNP Reactor Cavity Cooling Systems with water and air

This project has been focused on the experimental and numerical investigations of the water-cooled and air-cooled Reactor Cavity Cooling System (RCCS) designs. At this aim, we have leveraged an existing experimental facility at the University of Wisconsin-Madison (UW), and we have designed and built a separate effect test facility at the University of Michigan. The experimental facility at UW has underwent several upgrades, including the installation of advanced instrumentation (i.e. wire-mesh sensors) built at the University of Michigan. These provides highresolution time-resolved measurements of the void-fraction distribution in the risers of the water-cooled RCCS facility. A phenomenological model has been developed to assess the water cooled RCCS system stability and determine the root cause behind the oscillatory behavior that occurs under normal two-phase operation. Testing under various perturbations to the water-cooled RCCS facility have resulted in changes in the stability of the integral system. In particular, the effects on stability of inlet orifices, water tank volume have and system pressure been investigated. MELCOR was used as a predictive tool when performing inlet orificing tests and was able to capture the Density Wave Oscillations (DWOs) that occurred upon reaching saturation in the risers. The experimental and numerical results have then been used to provide RCCS design recommendations. The experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility has been equipped with state-of-theart high-resolution instrumentation to achieve so-called CFD grade experiments, that can be used for the validation of Computational Fluid Dynanmics (CFD) models, both RANS (Reynold-Averaged) and LES (Large Eddy Simulations). The effect of risers penetration in the upper plenum has been investigated as well.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Advanced Instrumentation for Transient Reactor Testing

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2017

Spacecraft Contamination Control Challenges for Space Missions with Organic Compound Detection Capabilities and for Potential Sample Return

Spacecraft contamination control is critical to space exploration missions with organic compound detection capabilities, and also for missions targeting acquisition of samples for potential return to Earth. Significant challenges are being addressed and resolved in the design of current flight projects and conceptual mission studies at JPL. These challenges extend to both orbiting spacecraft, as well as landed missions, for future missions to Mars and Europa, and potential missions to Titan and Enceladus. Contamination control during all phases of a mission, from preliminary design through operation, is fundamental to ensure that organic compounds of terrestrial origin are controlled to ensure successful completion of science objectives. This paper examines contamination control challenges specific to landed missions (which include sample acquisition, encapsulation, caching, potential sample return, and UV instruments), and orbiting missions (modeling interactions between the spacecraft and local exospheres and plumes).

Sylvia, Margarite↗
DOE OSTI2017

Advanced Simulation and Computing FY18 Implementation Plan, Rev. 0

The ASC Program continues to improve its unique tools to solve stockpile problems— with a focus on sufficient resolution, dimensionality, and scientific detail—to enable Quantification of Margins and Uncertainties (QMU) and to resolve the increasingly difficult analyses needed for stockpile stewardship. The needs of the Directed Stockpile Work (DSW) and major modernization programs also drive the requirements for simulation and computational resources. These requirements include planned LEPs, stockpile support activities, and mitigation efforts against the potential for technical surprise. All of the weapons within the current stockpile are in some stage of the life extension process. The simulation and computational capabilities are crucial for successful execution of these lifetime extensions and for ensuring NNSA can certify these life-extended weapons without conducting a UGT. Specific work activities and scope contained in this Implementation Plan (IP) represent the full-year annual operating plan for FY18. The Initial IP effective October 1, 2017, should be consistent with: 1) the Department’s Base Table when operating under a Continuing Resolution (CR); and 2) the final signed appropriation or full-year CR once enacted.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Component and Technology Development for Advanced Liquid Metal Reactors

The following report details the significant developments to Sodium Fast Reactor (SFR) technologies made throughout the course of this funding. This report will begin with an overview of the sodium loop and the improvements made over the course of this research to make it a more advanced and capable facility. These improvements have much to do with oxygen control and diagnostics. Thus a detailed report of advancements with respect to the cold trap, plugging meter, vanadium equilibration loop, and electrochemical oxygen sensor is included. Further analysis of the university’s moving magnet pump was performed and included in a section of this report. A continuous electrical resistance based level sensor was built and tested in the sodium with favorable results. Materials testing was done on diffusion bonded samples of metal and the results are presented here as well. A significant portion of this work went into the development of optical fiber temperature sensors which could be deployed in an SFR environment. Thus, a section of this report presents the work done to develop an encapsulation method for these fibers inside of a stainless steel capillary tube. High temperature testing was then done on the optical fiber ex situ in a furnace. Thermal response time was also explored with the optical fiber temperature sensors. Finally these optical fibers were deployed successfully in a sodium environment for data acquisition. As a test of the sodium deployable optical fiber temperature sensors they were installed in a sub-loop of the sodium facility which was constructed to promote the thermal striping effect in sodium. The optical fibers performed exceptionally well, yielding unprecedented 2 dimensional temperature profiles with good temporal resolution. Finally, this thermal striping loop was used to perform cross correlation velocimetry successfully over a wide range of flow rates.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Multi-objective optimization of a PCHE-type intermediate heat exchanger using genetic algorithms

For the advanced nuclear reactor systems and power conversion units, it is important to optimize the design of intermediate heat exchangers (IHXs) which transfer heat from the primary loop to the secondary or intermediate loop. For a multi-objective optimization problem, a printed circuit heat exchanger-type IHX with zigzag and S-shaped fin channels is selected to be optimized in terms of both the total cost and the heat exchanger thermal effectiveness using genetic algorithms. A computer code is developed to search a set of optimized designs of the zigzag channel geometry. It is found that among the Pareto-optimal solutions, the total cost is approximately linearly related to the thermal effectiveness between 88 and 95% while it rises rapidly after the effectiveness exceeds 95%. An optimal flow channel diameter is also determined through the optimization. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗