SEARCH · Engineering Papers
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Topics covered include: Our Science; Guiding Principle of Los Alamos; LANL the Design Laboratory; Weapons Mission; Science, Engineering and Computations; Future Weapons Program R&D; NNSA; Aging Weapons; A Final Thought; and, Questions and Answers.
Argonne National Laboratory’s Laboratory Directed Research and Development (LDRD) program encourages the development of novel technical concepts, enhances the Laboratory’s research and development (R&D) capabilities, and enables pursuit of strategic laboratory goals. Argonne’s LDRD projects are proposal based and peer reviewed, supporting ideas that require advanced exploration so they can be sufficiently developed to pursue support through normal programmatic channels. Among the aims of the projects supported by the LDRD program are the establishment of engineering proofs of principle, assessment of design feasibility for prospective facilities, development of instrumentation or computational methods or systems, and discoveries in fundamental science and exploratory development. All LDRD projects have demonstrable ties to one or more of the science, energy, environment, and national security missions of the U.S. Department of Energy (DOE) and its National Nuclear Security Administration (NNSA), and many are also relevant to the missions of other federal agencies that sponsor work at Argonne. A natural consequence of the more “applied” type projects is their concurrent relevance to industry. The LDRD program is managed in overarching portfolios, each containing multiple projects each fiscal year. The LDRD Prime portfolio is further divided into strategic focus areas aligned with Argonne’s strategic plan. The largest component of Argonne’s program is LDRD Prime, which emphasizes R&D explicitly aligned with Laboratory major initiatives in support of Argonne’s strategic plan. The choice of Focus Areas under the LDRD Prime component reflects the major initiatives; the state of development of relevant technical fields; the potential value of advancing those fields to DOE/NNSA and the nation; and the compatibility of the fields with existing facilities, capabilities, and staff expertise at Argonne. Focus Areas with projects that ended in FY18 are: Advanced Computing, Biological and Environmental Science Capability Development, Energy Manufacturing Science and Engineering, Hard X-ray Sciences, Materials and Chemistry, Securing Energy and Critical Resources, and The Universe as Our Laboratory (ULab).
This Corrective Action Plan (CAP) was developed by a team chartered by John Sarrao, Deputy Director for Science, Technology, and Engineering (DDSTE), to review and respond to the findings of the Joint National Nuclear Security Administration/Triad National Security (NNSA/Triad) Investigation Team (JIT) Joint Investigation Report for the “Sealed Source Recovery at the University of Washington Harborview Training and Research Facility Results in Release of Cesium-137 on May 2, 2019,” dated March 30, 2020. On May 2, 2019, International Isotopes, Inc. (INIS), a subcontractor to Triad National Security, LLC, inadvertently breached a sealed cesium-137 source at the University of Washington (UW), Harborview Medical Center Research and Training (HRT) building in downtown Seattle while attempting to recover the source for the National Nuclear Security Administration (NNSA) Off-Site Source Recovery Program (OSRP). The source breach resulted in contamination of personnel and the HRT building, as well as a release of radioactive material to the environment
Los Alamos National Laboratory (LANL or Laboratory) is one of several national laboratories that support the mission of the United States Department of Energy (DOE), National Nuclear Security Administration (NNSA), Los Alamos Field Office (Field Office). It is located in Los Alamos and Santa Fe Counties, New Mexico. LANL occupies approximately 40 square miles of land owned by the U.S. Government and under the administrative control of DOE/NNSA. Cultural resources management at LANL is the responsibility of the Field Office’s Cultural Resources Program Manager, with the assistance and facilitation of the LANL Environmental Stewardship Group (ES) Cultural Resources Programs (archaeology and historic buildings), which are part of the LANL Environmental Protection and Compliance (EPC) Division. LANL, which is primarily located on the Pajarito Plateau, is extremely rich in cultural resources. Nearly 2,000 archaeological sites have been recorded at LANL, based on the systematic archaeological survey of ~90% of LANL lands. Most of the known sites (~ 85%) date from the Ancestral Pueblo Late Coalition and Classic Periods of around AD 1150 to 1600. The remaining sites date to the Archaic Period (ca. 5500 BC to 600 AD), the Late Developmental and Early Coalition Periods (ca. 900 to 1200), and the Homestead Period (ca. 1890 to 1942).
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.
The National Security Research Center (NSRC) is a dynamic organization staffed with an expert, highly trained experts. The NSRC opened its doors in June 2019, making it one of the newest large organizations at Los Alamos National Laboratory (LANL). It supports a broad range of researchers within the LANL Weapons Program, customers across other National Nuclear Security Administration (NNSA) labs and sites, and partners in the Department of Defense (DoD). The maintenance and growth of the NSRC necessitate strategy and foresight. The NSRC leadership has developed a series of strategic goals through discussions with NSRC staff, customers, LANL leadership, and similar research institutions around the country to guide the NSRC leadership in moving the Center forward over the next five years.
Large pulsed power accelerators deliver multi-MJ pulses of electrical energy to a variety of high energy density (HED) physics experiments that support stockpile science programs. Understanding the plasma formation mechanisms and resulting electrical power transport (or "power flow") in the vacuum magnetically insulated transmission lines (MITLs) is an important area of ongoing research, and could provide a means to improve the performance of today's pulsed power accelerators while improving confidence in the design options for next-generation pulsed power concepts. Power flow science has been studied for decades, but these studies have not provided a predictive understanding of plasma formation and expansion in MITL systems. Several recent factors in pulsed power system design have generated a renewed (and urgent) interest in developing validated, multi-physics power flow engineering models with increased scrutiny and understanding. Examples of these factors include (i) the use of high inductance experimental configurations that could increase current "loss", (ii) interest in long-pulse applications that require predictable pulse shapes, and (iii) the desire to develop a deeper understanding of how current loss phenomena scale to larger accelerator configurations. This work is directed to support the validation of multi-physics power flow engineering models required to realize pulsed power systems for the NNSA mission.
As directed in the Consolidation Appropriations Act, 2016, the National Nuclear Security Administration (NNSA) initiated the preconceptual design and development of a Lifecycle Cost Estimate for the Surplus Plutonium Disposition (SPD) Dilute and Dispose Program. Based on August 2016 Program Requirements Document and subsequent supplemental guidance, LANL prepared the Lifecycle Cost Estimate under key assumptions that meet the program’s requirements. For Dilute and Dispose, the program would disposition surplus Pu by diluting oxide produced at LANL with inhibitor materials, packaging the materials in containers, and shipping the containers to a deep geologic repository for permanent disposal. The base assumption is that LANL would disassemble pits, convert the Pu metal to oxide, and characterize and package the material for shipment to SRS, where it would be diluted prior to geologic disposal at the WIPP site in New Mexico. Another major assumption for the Dilute and Dispose option is that LANL would increase the current oxide production rate (or throughput) to 1500 kg/year, 5 times higher than the maximum annual production of ~300 Kgs executed by the ARIES Oxide Production Program at LANL. Analysis based on the ARIES program’s throughput model revealed that 15 pieces of equipment would need to be installed within PF-4 and certain facility improvements would need to be accomplished in order to meet the desired throughput levels. The additional equipment would be essentially identical to equipment already used within PF-4 for existing operations.
Researchers at Sandia National Laboratories have integrated the GRANTA materials database with the MatCal calibration engine to calibrate material models from characterization data. GRANTA is gaining acceptance across the NNSA Tri-lab complex and is being populated with weapons-specific test data by Sandia experimentalists. To use that data to create material models for use by weapons systems analysts, MatCal has been enabled import calibration data and test conditions from GRANTA to quickly and reproducibly produce a calibrated set of parameters for a given constitutive model. The team is currently working to store the parameters characterizing material behavior in GRANTA to make them accessible by all weapons analysts.
The University of Missouri Research Reactor (MURR®) is one of six U.S. High Performance Research Reactors (USHPRR), including one critical facility, that is working with the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program to convert from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. The M3 Reactor Conversion USHPRR Project objectives include the development of LEU fuel element designs that will ensure safe reactor operations and to maintain the existing experimental performance of each facility. The work is being conducted through many inter-related activities being completed by four Project Pillars: Fuel Qualification (FQ), Fuel Fabrication (FF), Reactor Conversion (RC), and Cross Cutting (CC). A new type of LEU fuel based on an alloy of uranium-10 wt% molybdenum (U-10Mo) is expected to allow the conversion of those USHPRR, like MURR, requiring higher density fuels. The very-high-density LEU U-10Mo monolithic fuel is currently undergoing irradiation testing and post-irradiation examination under a planned and documented fuel qualification effort. The FQ Pillar will document fuel property and fuel performance data and qualify the fuel for use in these reactors. The FF Pillar is fabricating fuel for ongoing and future irradiation tests, as well as conducting fabrication demonstrations to validate or update preliminary fabrication assumptions. The FF Pillar is also working to develop and install commercial manufacturing capacity with the U-10Mo monolithic fuel to produce prototypic fuel. Working with the RC Pillar at Argonne, MURR has progressed through a preliminary fuel element design using preliminary data for the proposed monolithic alloy of U-10Mo. Analyses were completed in previous work that found for typical equilibrium operations with the preliminary LEU fuel element design, in conjunction with a power uprate to 12 MW and appropriate changes to the MURR Limiting safety system settings (LSSS), MURR will have adequate margins to safety for steady-state operations and postulated transient accidents and will have experimental performance in key locations that meets or exceeds current operations with HEU fuel. The purpose of this work is to develop a sequence of transition cycles that will enable MURR to transition from operation with the reactor core loaded with fresh LEU fuel elements only to typical equilibrium operations with mixed-burnup cores following conversion while meeting operational requirements on safety and experimental performance. It is expected that the use of fresh LEU fuel at conversion and subsequent low burnup of the LEU fuel elements that will initially be available for use following conversion will result in critical control blade positions that will substantially change the axial power distribution in the core and the neutron flux available in key experimental locations relative to equilibrium LEU operations. Given the constraints of MURR safety margins, operational practices, and production and research, a novel method has been developed to identify a transition sequence that minimizes the time MURR operates atypically compared to the current prototypic cycles using HEU fuel. The proposed transition sequence moves quickly to the same sort of equilibrium cycles for the LEU fuel that have already been evaluated in documented preliminary safety analyses. Although shifting the neutron flux peak to the lower half of the core during initial cycles with LEU at 12 MW reduces the experiment performance in some key locations relative to current HEU operations at 10 MW, all LEU cores provide an average performance that meets or exceeds that of HEU. An LEU cycle is reached that meets or exceeds the level of experimental performance predicted for current HEU and equilibrium LEU operations in more than 450 key locations identified by a reactor specialist at MURR by the 23rd cycle following conversion and that afterwards will enable MURR to consistently meet its experimental performance requirements. The proposed transition sequence only requires the fabrication of 34 fresh LEU elements in the first year of operation and does not exceed the anticipated availability of fresh elements that can be produced by the fuel fabricator. By the third year after conversion, 22 fresh LEU elements will be required each year, which is the same as expected for equilibrium LEU operations and the same as current operations with HEU fuel. The proposed transition sequence thus combines a relatively short time period before equilibrium burnup is achieved, a temporary increase of fuel elements needed annually relative to typical operations that are within the production capabilities of the fuel fabricator, and demonstrates comparable experimental performance of the LEU cores relative to current HEU operations. Further measures may be taken to reduce any initial experimental performance penalty even further, where possible, by repositioning certain experiments to leverage the increased performance in the lower axial experimental positions in the initial cycles following conversion or leaving the experiments in the irradiation facilities longer in order to achieve the required neutron fluence. This analysis may require refinement depending on the experimental facilities in use at the time of conversion. Nonetheless, the results presented here, including the experimental performance, core burnup, and critical control blade positions throughout the transition cycles, show that the proposed transition cycle fuel management patterns are consistent with what is expected and desired for MURR operation with LEU U-10Mo fuel. Detailed core power distributions from the neutronics models were also used to evaluate safety margins during steady-state operations for the selected transition cycles and the equilibrium LEU core. It is shown that there are adequate safety margins for both steady-state operations and postulated accident scenarios. For the steady-state operations with the preliminary LEU fuel element design the analysis predicts at least 2.49 MW margin to the onset of flow instability at the LSSS power of 15 MW. Considering the LSSS power is 125% of full license power, the margin to OFI is sufficient. In addition, the critical heat flux ratio at LSSS power is well above the requirement of CHFR > 2.0 from NUREG-1537 for all considered cases. For postulated transient accidents, the minimum margin to the fuel temperature safety limit is at least 109 °C. In summary, the proposed sequence of core loadings for MURR operations following conversion to LEU fuel and a power uprate to 12 MW provides sufficient safety margins for both steady-state operations and postulated transient accidents during a proposed sequence of transition cycles to equilibrium operations. Analysis has shown that there are some local experimental performance penalties during the initial cycles. Although there are local shifts in the experimental performance, on average all LEU cores at 12 MW have equal or higher performance than HEU at 10 MW. Temporary adjustments are being planned that will produce suitable experimental performance during these cycles. The results indicate that for the equilibrium LEU core the experimental performance exceeds that of current HEU operations in all key locations while also demonstrating sufficient safety margins.
Sandia National Laboratories (SNL) personnel operate a low power research reactor (the Annular Core Research Reactor, or ACRR), and a zero-power critical experiment assembly referred to as CX. In accordance with 10 CFR 830, Subpart B, Appendix A, the acceptable methodology for developing a Documented Safety Analysis (DSA) for DOE nuclear reactors is the Nuclear Regulatory Commission’s (NRC’s) Regulatory Guide 1.70 (RG 1.70). RG 1.70 does not address certain areas required by 10 CFR 830 and expected by DOE (e.g., full facility hazard analysis).Thus, the current DSAs for SNL’s reactor nuclear facilities are based on RG 1.70, but also of necessity supplemented by DOE-STD-3009-94 methods. SNL personnel, in consultation with the National Nuclear Security Administration (NNSA) Sandia Field Office (SFO), have concluded that an alternate methodology is preferred to RG 1.70. The details of the proposal, and the reasons motivating its development, are discussed in the order described below. The proposed alternate methodology will be applicable to the ACRR and the CX (i.e., it will be applicable to nuclear facilities in which a reactor and/or a critical assembly will be operated).
The Sri Lanka Atomic Energy Regulatory Commission (AERC) and the United States Department of Energy (U.S. DOE) co-hosted the 6th Regional Review Meeting on Radiological Security involving representatives from over 20 countries, the International Atomic Energy Agency (IAEA), the International Criminal Police Organization (INTERPOL), and the World Institute for Nuclear Security. The purpose of the event was to discuss the implementation of, and plans for, high-activity radioactive source security (RSS). The U.S. DOE’s National Nuclear Security Administration (NNSA) Office of Radiological Security (ORS) fully sponsored this review meeting. Participants were welcomed to Colombo, Sri Lanka, and the meeting was formally opened by Nirmali Karunarathna of the Sri Lanka Atomic Energy Regulator Council who emphasized the strong partnerships among the participants. The opening Ceremony and Lamp Lighting included dignitaries from the sponsoring countries. Robert Hilton, Deputy Chief of Mission at the U.S. Embassy in Colombo, and Kristin Hirsch of the ORS gave other opening remarks that highlighted the social benefit from radiological sources in medicine, industry, and agriculture, while stressing the importance of addressing the risks associated with the malicious use of radiological sources. Emphasis was placed on the importance of partnerships among all the participants to help ensure the success of securing radiological materials throughout the world and the opportunity to share information and experiences. AERC was acknowledged with special thanks for hosting this event. A participant list is included as Attachment A, and the meeting agenda is provided as Attachment B. All presentations were made available to participants. The following sections summarize the meeting’s presentations, discussions, issues, suggestions, and recommendations.
The goal of this work was to develop unique LLNL capabilities for the computational prediction of electrical, optical and thermal transport properties from quantum simulations. Using a recently developed theory of microscopic quantum transport, we were able to model compressed metallic hydrogen and it’s non-linear and out-of-equilibrium contributions in electrical and thermal transport. This theory was implemented in the open-source, real-space time-dependent density functional theory code Octopus. This model is general and widely applicable to high energy density systems that are important to LLNL and NNSA mission areas.
The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.
The U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA), Los Alamos Field Office (Field Office), has prepared final documentation for the resolution of adverse effects to Building 306 in Technical Area (TA) 16 at Los Alamos National Laboratory (LANL or the Laboratory). This documentation is being submitted to the New Mexico State Historic Preservation Officer (SHPO). TA-16-306 was determined eligible for listing in the National Register of Historic Places (Register) in 1995 in the report, TA-16 Heating System Replacement (LA-CP-95-0180). TA-16- 306 was identified as an excess property to be decontaminated, decommissioned, and demolished by the LANL Footprint Reduction Program in fiscal year 2021, an adverse effect to Register eligible building 16-306 requiring mitigation. To bring a resolution to the adverse effects to Building TA-16-306, the Field Office initiated consultation with the SHPO on December 4, 2017. The New Mexico Historic Preservation Division concurred with the mitigation actions outlined in the Programmatic Agreement among the U.S. Department of Energy, National Nuclear Security Administration, Los Alamos Field Office, the New Mexico State Historic Preservation Office, and the Advisory Council on Historic Preservation Concerning Management of Historic Properties of Los Alamos National Laboratory, Los Alamos, New Mexico (PA) in correspondence dated January 30, 2018. The PA states in Appendix D.2.A that adverse effects to Register-eligible buildings and structures will be resolved according to the procedures defined in A Plan for the Management of the Cultural Heritage at Los Alamos National Laboratory, New Mexico (CRMP) (LA-UR-19-21590, formerly LA-UR-15-27624) and within the PA itself. Volume 2 contains a collection of photos of building TA-16-306 with index.
The ASC Advanced Technology Development and Mitigation (ATDM) sub-program was established in 2014 to develop new simulation tools operating on exascale-class computers to serve NNSA (see Appendix B). Over the course of ATDM, LANL management have set a strategy for exascale-class application codes that follows two supportive and mutually risk-mitigating paths: evolution for established production integrated design codes (IDCs) – with a strong pedigree within the user community – based upon existing programming paradigms(MPI+X); and a new start ATDM project, Ristra, a high-risk/high-reward push for a next-generation multi-physics, multi-scale simulation toolkit based on emerging advanced programming systems(with an initial focus on data-flow task-based models exemplified by Legion). The role of Ristra as the high-risk/high-reward path for LANL’s codes was fully consistent with the goals of ATDM as described in Appendix B, in particular its emphasis on evolving ASC capabilities through novel computing programming models and computing technologies.
The Y-12 National Security Complex (Y-12) participates in the Fuel Fabrication Pillar of the National Nuclear Security Administration’s (NNSA’s) Office of Material Minimization and Management (M3) Office of Conversion Pillar. Y-12’s primary responsibility is to establish the fabrication process for the low-enriched uranium-molybdenum (U-10Mo) feedstock.
Slides to be presented during the NNSA 2021 Stewardship Science Academic Programs (SSAP) Symposium (https://web.cvent.com/event/1893b742-2d98-49d5-87cc-f07b1c3cb66c/summary?environment=P2). The symposium will be held virtually February 16–18, 2021.