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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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DOE OSTI2024

McGrath, Alaska Community Energy Plan [Slides]

The US Department of Energy's Energy Transitions Initiative Partnership Project (ETIPP) works alongside remote and island communities seeking to transform their energy systems and increase energy resilience. The City of McGrath took part in the ETIPP program in 2023-2024. As part of the project, community members formed the McGrath Energy Committee, made up of residents and local stakeholder organizations. The McGrath Energy Committee then worked with technical advisors from regional associations, university programs, and national labs to conduct a baseline energy assessment of the community, organize a community energy education series, identify key focus areas relevant to McGrath, explore funding opportunities, and create the McGrath Community Energy Plan. This plan serves as a foundational guide for future energy projects in the community, aligned with McGrath's long-term energy goal: "to be a catalyst to encourage energy resiliency in our community and the Upper Kuskokwim Region." - November 2024.

14 SOLAR ENERGY↗
DOE OSTI2021

Launching a saliva-based SARS-CoV-2 surveillance testing program on a university campus

Regular surveillance testing of asymptomatic individuals for SARS-CoV-2 has been center to SARS-CoV-2 outbreak prevention on college and university campuses. Here we describe the voluntary saliva testing program instituted at the University of California, Berkeley during an early period of the SARS-CoV-2 pandemic in 2020. The program was administered as a research study ahead of clinical implementation, enabling us to launch surveillance testing while continuing to optimize the assay. Results of both the testing protocol itself and the study participants’ experience show how the program succeeded in providing routine, robust testing capable of contributing to outbreak prevention within a campus community and offer strategies for encouraging participation and a sense of civic responsibility.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2020

RELAP-7 Application and Enhancement for FLEX Strategies and ATF Behavior under Extended Loss of AC Power Conditions

This report summarizes the results of a three-year research project sponsored by the U.S. Department of Energy (DOE) Nuclear Energy University Program (NEUP) to enhance and apply the RELAP-7 code by adding and improving several important components (e.g., a mechanistic Reactor Core Isolation Cooling (RCIC) system model) for thermal hydraulic studies of LWRs under ELAP conditions and evaluating the time available for transition to portable FLEX equipment. The project team included University of Massachusetts–Lowell, The Ohio State University, Texas A&M University, Idaho National Laboratory and Oak Ridge National Laboratory. In the Fukushima accident, it was found that the RCIC system played a crucial role in delaying core meltdown by almost three days in Fukushima Daiichi Unit 2, because of self-regulated operation of the steam driven RCIC turbine-pump injection system. Steam flow in the convergent-divergent nozzles of the RCIC Terry turbine is two-phase non-equilibrium transonic flow with homogenous nucleation condensation. To more accurately predict the dynamic process and behavior of the transonic compressible steam flow, a one-dimensional transient two-phase analytical model is presented. A simplified four-fluid model was employed in the present work with the consideration of four separate fluid fields: vapor, liquid film, entrained droplets and condensed droplets. The mass, momentum and energy interactions between the fluids were considered and modeled. An extended seven-equation non-equilibrium critical flow model was developed to obtain the critical pressure and velocities of each phase at the nozzle throat. To predict the wetness in the divergent section, a mechanistic nucleation condensation model was integrated in the nozzle analysis model, considering the generation and consequent growth of droplets. The governing differential equations on a staggered grid were discretized using the second-order Lax-Wendroff scheme with a flux limiter, and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE) algorithm was employed to solve the discrete linear system. To demonstrate the predictability and reliability of the physical models and the numerical method proposed in the present work, three representative nozzles were modeled and simulated. The results show good agreement with the available experimental data, even for condensation shock. Then, the 1D nozzle model was employed to obtain nozzle flow tables of the Terry turbine nozzle for different working pressures which can cover the operation pressure range of the RCIC system. A mechanistic RCIC turbine-pump system model was developed and implemented in the system code TRACE to simulate dynamic responses of the RCIC system under Beyond Design Basis Accident (BDBA) conditions. The turbine-pump governing equations are based on the control volume approach of the angular momentum balance. The physics based mechanistic RCIC model was developed using the TRACE control system components (i.e., signal variables, control blocks, and tables), and incorporated into a TRACE boiling water reactor (BWR) model. The TRACE model in this report has a detailed nodalization of the reactor pressure vessel (RPV), and all of the major flow paths and system components, including the safety relief valves (SRVs) and the containment suppression pool and drywell. Based on the nozzle flow tables generated from the 1D nozzle model developed, the turbine drive torque can be calculated from table lookup. Since the detailed specifications of the RCIC pump are unavailable, the homologous curves for a Bingham pump were used in the current pump component. A station black-out (SBO) accident test problem was selected to demonstrate the TRACE RCIC model. The short-term SBO simulations were performed for two cladding materials: Zircaloy and FeCrAl, to demonstrate the effect of the accident tolerant fuel cladding on fuel heat-up under BDBA accident conditions. The wetwell plays a vital safety role in SBO and other BWR accident scenarios in that it can reduce containment pressure and supply additional core make-up water. The suppression pool temperature distribution has a very large impact on both RPV and containment pressure. Thus, another novel contribution of the project comes mainly from an improved, systems-level wetwell model which can capture buoyancy-induced thermal stratification effects due to steam injection and condensation. A two-zone stratified wetwell model has been implemented in RELAP-7 and some results from that model are presented. This wetwell model is capable of simulating thermal stratification due to a low steam mass injection rate. With a low mass flow rate, the model assumes that all the steam condenses within the pipe and the resulting plume can be approximated with a purely buoyant, heat-source driven model. The wetwell model developed with these assumptions is adequate to simulate slow transients such as extended SBO transients.

42 ENGINEERING↗
DOE OSTI2023

Compare predictions of transient fission gas release by empirical and mechanistic models to experiments in high burnup UO 2 fuel

Understanding and predicting fuel performance at high burnup require improving our understanding of transient fission gas release. High-burnup operations enable new mechanisms of fission gas release, which affect fuel performance. The Nuclear Regulatory Commission has recently published its interpretation of existing fuel fragmentation, relocation, and dispersal data in a research information letter. There, transient fission gas release was identified as one of the main factors that contributes to fuel fragmentation, relocation, and dispersal, and therefore limits fuel extension to high burnup. However, transient fission gas release is a complex phenomenon that cannot be fully described by simple empirical descriptions. This report summarizes the development of a mechanistic model for high-burnup transient fission gas release in the fuel performance code BISON. This research was supported by the Nuclear Energy Advanced Modeling and Simulation program during fiscal year 2023 to improve our understanding of high-burnup transient fission gas release and ability to predict it as a function of operation history. To support the development of a mechanistic transient fission gas release model, the existing Simple Integrated Fission Gas Release and Swelling (Sifgrs) model in BISON has been completely refactored to make it more modular and extensible. This effort supports the model's application to high-burnup conditions, its extension to other fuel forms, and the continuous improvement of its current features. Once refactoring was completed, models for high-burnup structure formation, fission gas transfer from non-restructured fuel to high-burnup structure, high-burnup structure intragranular and intergranular fission gas behavior, high-burnup structure bubble evolution, fuel pulverization, and the resulting transient fission gas release were tested and implemented in the Simple Integrated Fission Gas Release and Swelling (Sifgrs) model or tightly coupled to it. The new mechanistic model was then compared to an empirical model developed in parallel by a Nuclear Energy University Program project using a Studsvik high-burnup loss-of-coolant-accident assessment case. Finally, the report details the preliminary BISON results for a benchmark activity organized by the Nuclear Energy Agency to evaluate fuel performance codes' predictive capabilities for burst fission gas release. This work represents an important step toward a mechanistic understanding of fission gas release in high-burnup conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Phenomena identification and ranking table study for thermal hydraulics for Advanced High Temperature Reactor

The Advanced High-Temperature Reactor (AHTR) is a conceptual design of a Fluoride salt-cooled High-temperature Reactor (FHR) utilizing 7 LiF-BeF 2 (66-34 mol%) as its primary coolant. To identify key phenomena that impose potential challenges on thermal hydraulics modeling and simulation of such a reactor system, a thermal hydraulics phenomena identification and ranking table (TH-PIRT) study was performed for the AHTR in a Department of Energy Nuclear Energy University Program (NEUP) integrated research project led by Georgia Institute of Technology. A panel of experts from regulators, industries, national laboratories, and academia was assembled for the study. In this paper, the TH-PIRTs identified by the panel for two events, namely, station blackout and simultaneous withdrawal of all control rods, are summarized and discussed in detail. In addition, the key phenomena that warrant further study and research for AHTR analysis are identified to support the validation of thermal hydraulics system-level analysis codes and computational fluid dynamics simulation tools, as well as future FHR reactor licensing.

42 ENGINEERING↗
DOE OSTI2018

Microscopic Behaviors of Tri-n-Butyl Phosphate, n-Dodecane and Their Mixtures at Air/Liquid and Liquid/Liquid Interfaces: An AMBER Polarizable Force Field Study

In solvent extraction processes for recovering metal ions from used nuclear fuel, as well as other industrial applications, a better understanding of the metal complex phase transfer phenomenon would greatly aid ligand design and process optimization. We have approached this challenge by utilizing classical molecular dynamics simulations technique to gain visual appreciation of the vapor/liquid and liquid/liquid interface between tri-n-butyl phosphate (TBP) and n-dodecane with air and water. In this study, we successfully reparameterized polarizable force fields for TBP and n-dodecane that accurately reproduced several of their thermophysical properties such as: density, heat of vaporization, and dipole moment. Our models were able to predict the surface and interfacial tension of different systems when compared to experimental results also performed by us. Through this study, we gained numerous atomistic understanding of the behaviors of TBP and n-dodecane at the interface against air and water, useful in further computational studies of such systems. Lastly, our studies indicate that the initial configuration of a simulation may have a large effect on the final result.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Improvement of Design Codes to Account for Accident Thermal Effects on Seismic Performance

This report summarizes the results of experimental and numerical studies conducted to evaluate the effect of accident thermal loads on the seismic response of steel-plate composite (SC) walls and reinforced concrete (RC) walls and beams. The research was sponsored by U.S. Department of Energy through the Nuclear Energy University Program (NEUP), Award Number DE-NE0008256. Summary and conclusions of the research conducted are presented in this report. Based on the observations, analysis and design recommendations for SC and RC structures are also summarized

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Safeguards by Design Challenge (Final Report- FY19)

This University Engagement project challenged engineering students at universities, without nuclear engineering degrees but with research reactors, to incorporate Safeguards by Design concepts into their Senior Capstone Design Project. This University Engagement project was part of the U. S. Department of Energy’s (DOE) National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation, Office of International Nuclear Safeguards, Next Generation Safeguards Initiative, Human Capital Development: University Engagement Program. This program exposed university students with Mechanical Engineering major and Nuclear Engineering minors to the concepts of international nuclear safeguards. FY19, three teams at the University of Rhode Island and one team at the University of Texas - Austin participated in researching, designing, building, and testing projects to support international nuclear safeguards measurements or verification. The projects involved engaging in activities at the university’s research reactors. All the projects engaged students with prototyping a design and/or tool for application at the Universities’ Research Reactor. At the end of the course, most of the students expressed the experience was a positive and they learned more about international nuclear safeguards and applying requirements than they had previously encountered.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Safeguards by Design Projects (FY2020 Final Report)

This University Engagement project challenged engineering students at universities, that do not have Bachelor degrees in nuclear engineering but do have research reactors and some nuclear engineering coursework, to incorporate Safeguards by Design concepts into their Senior Capstone Design Project. This University Engagement project was part of the U. S. Department of Energy’s (DOE) National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation, Office of International Nuclear Safeguards, Next Generation Safeguards Initiative, Human Capital Development: University Engagement Program. This program exposed university students with Mechanical Engineering majors and Nuclear Engineering minors to the concepts of international nuclear safeguards. FY20, three teams at the University of Rhode Island and one team at the University of Texas - Austin participated in researching, designing, building, and testing projects to support international nuclear safeguards measurements or verification. The projects involved engaging in activities at the university’s research reactors. All the projects engaged students with prototyping a design and/or tool for application at the Universities’ reactor. At the end of the course, most of the students expressed the experience was a positive and they learned more about international nuclear safeguards and applying requirements than they had previously encountered. This school year the projects were further complicated by the COVID-19 pandemic. Both universities cancelled in class room classes, and limited direct student/professor interactions. Furthermore, Los Alamos National Laboratory (LANL) greatly restricted travel, therefore making it impossible to visit the students at the end of the semester for the review of their design projects. The final design and review meeting for the projects happened via the internet. Additionally, while the teams planned to build prototypes this did not happen since there was a social distancing ban on students meeting in person.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Safeguards by Design Projects (Final Report FY2021)

This University Engagement project challenged engineering students at universities, that do not have Bachelor degrees in nuclear engineering but do have research reactors and some nuclear engineering coursework, to incorporate Safeguards by Design concepts into their Senior Capstone Design Project. This University Engagement project was part of the U. S. Department of Energy’s (DOE) National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation, Office of International Nuclear Safeguards, Next Generation Safeguards Initiative, Human Capital Development: University Engagement Program. This program exposed university students with Mechanical Engineering majors and Nuclear Engineering minors to the concepts of international nuclear safeguards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

SiC cladding R&D execution plan

To advance the development of accident-tolerant fuel cladding for light-water reactors, the Advanced Fuels Campaign actively addresses the R&D gap between the specialized technological progress in SiC-based composite materials, led by industrial partners, and the fundamental scientific understanding of material behavior under reactor operating conditions. The laboratory activities are systematically organized based on the Development Strategy document, which outlines critical barriers to the use of SiC composites in nuclear energy applications for the coming decade and beyond. To complement the strategy, this execution plan specifies a detailed year-by-year work package and establishes key milestones for assessing progress in technological advancement. This plan has been formulated in collaboration with principal investigators at national laboratories, stakeholders, and industry partners with the goal of closing technology gaps and readying SiC-based cladding technologies for full-scale commercial testing. The research conducted at the national laboratories is categorized into three primary subjects: (1) addressing challenges associated with the performance of SiC cladding under normal operating environments, (2) identifying the failure limits of SiC cladding under accident conditions, and (3) modeling fuel performance for SiC cladding systems. Comprehensive laboratory activities integrate the development of advanced modeling tools for SiC-based cladding systems with experimental validation via integrated testing of multiphysics phenomena. This document lays out the laboratory plan to support the development of SiC-based technologies, which have the potential for significant operational and safety benefits for the U.S. nuclear industry. This plan aligns with industry developer needs, relevant Nuclear Energy University Programs, and Nuclear Energy Advanced Modeling and Simulation Programs to ensure SiC cladding technologies are matured as efficiently as possible and will be maintained and updated to accommodate the dynamically evolving progress of development, capability needs, and resource availability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Radiography of Dry Cask Storage for Used Nuclear Fuel

Since the U.S. currently only approves of storing used nuclear fuel in pools or dry casks, the demand for dry cask storage is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. With the rising demand comes additional pressure to ensure the integrity of dry cask systems. Visual inspection is costly and man-power intensive, so alternative nondestructive testing techniques are desired to insure the continued safe and effective storage of fuel. Several approaches being investigated by the University of Florida and the larger multi-university Department of Energy (DOE) Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) are gamma or neutron radiography, both transmission based and interior source emission. Additionally muon radiography is discussed. This work leverages the well-established techniques in medical image reconstruction and applies them to a significantly larger scale with additional considerations due to the dense nuclear materials which serves as effective shields for radiation. Emission source radiography relies on the radiation that is emitted from the used fuel. The main advantage of this technique is the passive nature of the measurement. The fuel is already emitting radiation and only detectors are needed. Transmission radiography uses an external source of radiation to penetrate the side of the cask. Challenges include selecting the external source spectrum to avoid overlap with the spectrum emitted from the fuel, and ensuring the radiation dose limits do not exceed those mandated by the Nuclear Regulatory Commission (NRC). Benefits of this technique include better understanding of the source location allowing for clear image reconstruction, and isolation of the signal from surrounding radiation sources. Muon radiography takes advantage of the natural background flux coming from the interaction of the atmosphere with cosmic radiation. The technique is particularly effective at interrogating large dense objects as the highly energetic muons have smaller cross sections than most typical sources and energies used in nuclear engineering. The various techniques are discussed individually then the conclusion compares them. Finally, future work is identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage

This final report is summarized four years of work of awarded project number DE-NE0008442\IRP-15-9318, title as: “Innovative Approach to SCC Inspection and Evaluation of Canister in Dry” supported by Nuclear Energy University Program\Integrated Research Program, NEUP\IRP, where last year is an extension of no cost project. This research work is focused on the possible degradation of dry storage canister holding used nuclear fuel (UNF), due to exposure of marine atmospheric environment. The canister usually designed from a welded austenitic stainless-steel of types 304 or 316, encased in a concrete shell known as an overpack. This canister designed and licensed for 20 years with an extension period need to be determined, in addition 20 years or more. These casks are located throughout the country in Independent Spent Fuel Storage Installations (IFSFIs) often on the site close to nuclear power plants, that are often located in coastal or lake-side regions where a marine atmospheric salt air environment exist. As a result of several recent decisions, such as not pursuing Yucca Mountain as a long-term geologic repository storage facility for UNF and the cancellation of the construction of a reprocessing facility in the 1980s due to proliferation risks, interim dry storage facilities for long-term storage are vital option for the immediate future of nuclear energy technology. One of the primary concerns with respect to the long-term performance of the storage casks is the potential for corrosion initiation due to deliquescence of salts deposited on the canister surface as aerosols; in regions of high residual weld stresses, this may lead to localized stress corrosion cracking (SCC). Dust and aerosols in the air being drawn though ventilation openings in the overpacks of passively ventilated dry canister storage systems may be deposited on the stainless-steel canister outer surfaces. Under these conditions, localized corrosion attack can occur. Chlorides ions deliquescence on the canister surface, a process by which the salts essentially pull moisture out of the air, creating a concentrated salt brine on the canister surface. Chloride-induced stress corrosion cracking (CISCC) of welded zones is of special concern, as it a well-documented mode of attack for austenitic stainless steels marine environments, and many independent spent fuel storage installations (ISFSIs) are located in coastal areas. Recent canister inspections have shown that chloride salts are present on the surface of in-service canisters at near-marine settings. However, canister surface inspections of satisfactory resolution to detect SCC have never been carried out in details, because access to the canister surfaces through vents in the overpacks is extremely limited, and high radiation fields make removal of the canisters from the overpacks undesirable. Although no significant corrosion damage has not been detected up to date, there is concern that this may lead to pitting corrosion, which could in turn develop into stress corrosion cracking, causing canister structural failure. The initiation and propagation of pitting corrosion has been widely studied in full immersion conditions, but limited studies have been done on marine atmospheric conditions, which is a focus of this work, these includes laboratory and field experimentally and propagation rate modeling of pitting and transition to crack, and crack propagation rate. It is known, however, that the rate of atmospheric corrosion can be affected by variations in relative humidity (RH), salt density, salt type, droplet area, temperature and stainless-steel microstructure and residual stress particularly in welded zones of canisters. The integrated research program has identified potential deterioration mechanisms for the stainless-steel canisters containing the UNF in dry storage that require detailed research and investigation and will have an impact on the performance of long-term interim storage under the normal and extreme environmental conditions experienced during the duration of this storage. The prediction and monitoring of canister corrosion processes while in storage can provide important data for the assessment of interim storage performance and the safety to the public. This report comprises of mainly the published results in different journals and covering a wide range spectrum from susceptible stainless-steel types 304\316 microstructural analysis through the implant test at atmospheric environment, residual stress analysis of mockup canister and simulation, inspection methodology includes nonlinear ultrasound and vibro-thermography and stochastic modeling and simulation of pits propagation rates and transition to cracks. Description of field specimens installed at ISFSI of Main Yankee is given in the following chapters and will be analyzed in the coming future and serves as an important data for a stochastic analysis as it was drawn from the publications and reports described in this report. The purpose of this report is to bring together the project efforts that includes improving in our understanding of structural aging performance at current dry storage casks fleet at ISFSIs in the context of remedial maintenance and inspection applications technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Effects of Calcite Impurities in Bentonite Buffer Systems on Uranium(VI) Solution Speciation and Sorption Behavior

In many nuclear waste disposal scenarios, bentonite is the proposed buffer material surrounding nuclear waste containers. Its main mineralogical component, Na-montmorillonite clay, is expected to limit radionuclide mobility to slow diffusion processes due to the clay’s low porosity and permeability, and the high sorption affinity of clay surface complexation and cation exchange sites for many radionuclides. Given its mineral structure, montmorillonite provides two types of porosities that can be accessible to radioactive contaminants: macropores between clay particles and interlayer spaces within clay particles, which become dominant at increasing degrees of bentonite compaction. Despite the fact that bentonite naturally contains a variety of mineral impurities (e.g., quartz, feldspars and calcite), most previous studies have only focused on radionuclide interactions with pure Na-montmorillonite phases. Hence, in this study, we will investigate the influence of calcite impurities on uranium(VI) (U(VI)) solution speciation and sorption behavior in bentonite systems. In the presence of water, calcite impurities are expected to lead to milli-molar concentrations of dissolved calcium over a wide range of pH conditions. Consequently, Ca 2+ could – at least partially – replace Na + on clay cation exchange sites, with further implications for exchange reactions of cationic U(VI) species. In addition, ternary U(VI)-Ca-carbonate solution species (CaUO 2 (CO 3 ) 3 2- and Ca 2 UO 2 (CO 3 ) 3 ) will largely replace U(VI)-carbonate species (UO 2 (CO 3 ) 3 4- and UO 2 (CO 3 ) 2 2- ) at alkaline pH. These larger, but lower-charged U(VI) species, may show different sorption and diffusion characteristics, e.g. if their access to negatively-charged clay interlayer spaces is different from binary U(VI)-carbonate complexes. In this presentation, we will first demonstrate the impacts of calcite impurities on U(VI) solution speciation based on PHREEQC simulations. We will then discuss first results from U(VI) sorption studies in these systems. This research is being performed using funding received from the DOE Office of Nuclear Energy's Nuclear Energy University Program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Sandia National Laboratories Early Career University Faculty Mentoring Program in International Safeguards

An international safeguards mentoring program was established by Sandia National Laboratories (SNL) for early career university faculty. The inaugural year of the program focused on course material development and connecting faculty to experts at national laboratories. Two faculty members were selected for participation; one developed a safeguards-by-design course, and the other created lecture material related to unmanned robotic systems in safeguards for integration in existing courses. Faculty members were paired with SNL subject matter experts based on the topic of their individual projects. The program also included a two week visit to SNL. The structure of this program not only supported the development of new course material, but also provided junior faculty members with an opportunity to make connections and build collaborations in the field of international safeguards. Programs like this are important for professional development of faculty members and to help strengthen connections between universities and the national laboratories.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Sandia National Laboratories Early Career University Faculty Mentoring Program in International Safeguards

Recent years have seen a significantly increased focus in the areas of knowledge retention and mentoring of junior staff within the U.S. national laboratory complex. In order to involve the university community in this process, as well, an international safeguards mentoring program was established by Sandia National Laboratories (SNL) for early career university faculty. After a successful experience during 2019, the program continued into 2020 to include two new faculty members who were paired with SNL subject matter experts based on the topic of their individual projects: one to work on advanced laboratory work for physics, technology, and policy of nuclear safeguards and nonproliferation, and the other to look at machine learning applied to international safeguards and nonproliferation. There is a two-pronged purpose to the program: fostering the development of educational resources available for international safeguards and exploring new research topics stemming from the exchange of mentor and mentee. Further, the program as a whole allows for junior faculty members to establish and expand a relationship network within international safeguards. In addition, programs such as this build stronger connections between the academic and the national laboratory community. Thanks to the junior faculty members that now have new connections into the laboratory community and potential for collaboration projects with the laboratories in the future, safeguards knowledge can actually increase far beyond just individually engaging students using this new and efficient avenue.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2021

Sandia Academic Alliance Program Collaboration Report: 2020-2021 Accomplishments

University partnerships play an essential role in sustaining Sandia’s vitality as a national laboratory. The SAA is an element of Sandia’s broader University Partnerships program, which facilitates recruiting and research collaborations with dozens of universities annually. The SAA program has two three-year goals. SAA aims to realize a step increase in hiring results, by growing the total annual inexperienced hires from each out-of-state SAA university. SAA also strives to establish and sustain strategic research partnerships by establishing several federally sponsored collaborations and multi-institutional consortiums in science & technology (S&T) priorities such as autonomy, advanced computing, hypersonics, quantum information science, and data science. The SAA program facilitates access to talent, ideas, and Research & Development facilities through strong university partnerships. Earlier this year, the SAA program and campus executives hosted John Myers, Sandia’s former Senior Director of Human Resources (HR) and Communications, and senior-level staff at Georgia Tech, U of Illinois, Purdue, UNM, and UT Austin. These campus visits provided an opportunity to share the history of the partnerships from the university leadership, tours of research facilities, and discussions of ongoing technical work and potential recruiting opportunities. These visits also provided valuable feedback to HR management that will help Sandia realize a step increase in hiring from SAA schools. The 2020-2021 Collaboration Report is a compilation of accomplishments in 2020 and 2021 from SAA and Sandia’s valued SAA university partners.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗