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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 253 records · Page 14

Evaluating Knowledge to Support Climate Action: A Framework for Sustained Assessment. Report of an Independent Advisory Committee on Applied Climate Assessment

As states, cities, tribes, and private interests cope with climate damages and seek to increase preparedness and resilience, they will need to navigate myriad choices and options available to them. Making these choices in ways that identify pathways for climate action that support their development objectives will require constructive public dialogue, community participation, and flexible and ongoing access to science- and experience-based knowledge. In 2016, a Federal Advisory Committee (FAC) was convened to recommend how to conduct a sustained National Climate Assessment (NCA) to increase the relevance and usability of assessments for informing action. The FAC was disbanded in 2017, but members and additional experts reconvened to complete the report that is presented here. A key recommendation is establishing a new nonfederal ‘‘climate assessment consortium’’ to increase the role of state/local/tribal government and civil society in assessments. The expanded process would 1) focus on applied problems faced by practitioners, 2) organize sustained partnerships for collaborative learning across similar projects and case studies to identify effective tested practices, and 3) assess and improve knowledge-based methods for project implementation. Specific recommendations include evaluating climate models and data using user-defined metrics; improving benefit–cost assessment and supporting decision-making under uncertainty; and accelerating application of tools and methods such as citizen science, artificial intelligence, indicators, and geospatial analysis. The recommendations are the result of broad consultation and present an ambitious agenda for federal agencies, state/local/tribal jurisdictions, universities and the research sector, professional associations, nongovernmental and community-based organizations, and private-sector firms.

54 ENVIRONMENTAL SCIENCES↗

Theoretical Treatment of Pressure Gradients in One-Dimensional, Inhomogeneous, Compressible Media

The purpose of the Generic Safety Issue (GSI) -191 research program sponsored by the U.S. Nuclear Regulatory Commission (NRC) was to determine if the formation, transport and accumulation of debris in a containment building following a loss-of-coolant accident (LOCA) could impede the operation of the emergency-core-cooling system (ECCS) in pressurized-water reactors (PWRs). The definition of “debris” was incrementally expanded over the course of the program to include not only thermal insulation damaged by direct jet impingement from a broken pipe, but also the latent dust and dirt that may be present in containment, and most recently, any adverse chemical products that might arise from the unique combination of spray additives, structural materials, and damaged insulation and coatings that will be present in the post-LOCA accident environment. The Integrated Chemical Effects Test (ICET) program was initiated by the NRC and conducted at the University of New Mexico (UNM) under the direction of Los Alamos National Laboratory (LANL) to study the possible formation of chemical products in prototypical containment conditions.

42 ENGINEERING↗

Experimental Study and Computational Simulations of Key Pebble Bed Thermo-mechanics Issues for Design and Safety

An experimental and computational study, consisting of modeling and simulation (M&S), of key thermal-mechanical issues affecting the design and safety of pebble-bed (PB) reactors was conducted. The objective was to broaden understanding and experimentally validate thermal-mechanic phenomena of nuclear grade graphite, specifically, spheres in frictional contact as anticipated in the bed under reactor relevant pressures and temperatures. The contact generates graphite dust particulates that can subsequently be transported into the flowing gaseous coolent. Under postulated depressurization transients and with the potential for leaked fission products to be adsorbed onto graphite 'dust', there is the potential for fission products to escape from the primary volume. This is a design safety concern. Furthermore, earlier safety assessment identified the distinct possibility for the dispersed dust to combust in contact with air if sufficient conditions are met. Both of these phenomena were noted as important to design review and containing uncertainty to warrant study. The team designed and conducted two separate effects tests to study and benchmark the potential dust-generation rate, as well as study the conditions under which a dust explosion may occure in a standardized, instrumented explosion chamber.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

A Historical Evaluation of the U16a Tunnel, Nevada National Security Site, Nye County, Nevada Volume 1

This report presents a historical evaluation of the U16a Tunnel on the Nevada National Security Site in southern Nevada. The work was conducted by the Desert Research Institute at the request of the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office and the U.S. Department of Defense, Defense Threat Reduction Agency. The U16a Tunnel was used for underground nuclear weapons effects tests in Shoshone Mountain in Area 16 of the Nevada National Security Site. Six nuclear tests were conducted in the U16a Tunnel from 1962 to 1971. These tests are Marshmallow, Gum Drop, Double Play, Ming Vase, Diamond Dust, and Diamond Mine. The U.S. Department of Defense Threat Reduction Agency, with participation from Lawrence Livermore National Laboratory and Las Alamos National Laboratory, sponsored the tests. Fifteen high explosives tests were also conducted at the tunnel. Two were calibration tests during nuclear testing and the remaining were U.S. Department of Defense, Defense Threat Reduction Agency tunnel defeat tests. The U16a Tunnel complex is on the top and slopes of Shoshone Mountain, encompassing an area of approximately 16.7 hectares (41.1 acres). Major modifications to the landscape are a result of three principal activities, road construction and maintenance, mining activities related to development of the tunnel complex, and site preparation for activities related to testing. Forty-seven cultural features were recorded at the portal and on the slopes of Shoshone Mountain. At the portal area, features relate to the mining, construction, testing, and general every day operational support activities within the tunnel. These include concrete foundations for buildings, equipment pads, and rail lines. Features on the slopes above the tunnel relate to tunnel ventilation, borehole drilling, and data recording. Feature types include soil-covered bunkers, concrete foundations, instrument cable holes, drill holes, and ventilation shafts. The U16a Tunnel complex is eligible to the National Register of Historic Places under criteria a and c, consideration g of 36 CFR Part 60.4 as a historic landscape. Scientific research conducted at the tunnel has made significant contributions to the broad patterns of our history, particularly in regard to the Cold War era that was characterized by competing social, economic, and political ideologies between the former Soviet Union and the United States. The tunnel also possesses distinctive construction and engineering methods for conducting underground nuclear tests. The Desert Research Institute recommends that the U16a Tunnel area be left in place in its current condition and that the U16a Tunnel historic landscape be included in the Nevada National Security Site monitoring program and monitored on a regular basis.

54 ENVIRONMENTAL SCIENCES↗

A Historical Evaluation of the U16a Tunnel, Nevada National Security Site, Nye County, Nevada Volume 2

This report presents a historical evaluation of the U16a Tunnel on the Nevada National Security Site in southern Nevada. The work was conducted by the Desert Research Institute at the request of the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office and the U.S. Department of Defense, Defense Threat Reduction Agency. The U16a Tunnel was used for underground nuclear weapons effects tests in Shoshone Mountain in Area 16 of the Nevada National Security Site. Six nuclear tests were conducted in the U16a Tunnel from 1962 to 1971. These tests are Marshmallow, Gum Drop, Double Play, Ming Vase, Diamond Dust, and Diamond Mine. The U.S. Department of Defense Threat Reduction Agency, with participation from Lawrence Livermore National Laboratory and Las Alamos National Laboratory, sponsored the tests. Fifteen high explosives tests were also conducted at the tunnel. Two were calibration tests during nuclear testing and the remaining were U.S. Department of Defense, Defense Threat Reduction Agency tunnel defeat tests. The U16a Tunnel complex is on the top and slopes of Shoshone Mountain, encompassing an area of approximately 16.7 hectares (41.1 acres). Major modifications to the landscape are a result of three principal activities, road construction and maintenance, mining activities related to development of the tunnel complex, and site preparation for activities related to testing. Forty-seven cultural features were recorded at the portal and on the slopes of Shoshone Mountain. At the portal area, features relate to the mining, construction, testing, and general every day operational support activities within the tunnel. These include concrete foundations for buildings, equipment pads, and rail lines. Features on the slopes above the tunnel relate to tunnel ventilation, borehole drilling, and data recording. Feature types include soil-covered bunkers, concrete foundations, instrument cable holes, drill holes, and ventilation shafts. The U16a Tunnel complex is eligible to the National Register of Historic Places under criteria a and c, consideration g of 36 CFR Part 60.4 as a historic landscape. Scientific research conducted at the tunnel has made significant contributions to the broad patterns of our history, particularly in regard to the Cold War era that was characterized by competing social, economic, and political ideologies between the former Soviet Union and the United States. The tunnel also possesses distinctive construction and engineering methods for conducting underground nuclear tests. The Desert Research Institute recommends that the U16a Tunnel area be left in place in its current condition and that the U16a Tunnel historic landscape be included in the Nevada National Security Site monitoring program and monitored on a regular basis.

54 ENVIRONMENTAL SCIENCES↗

Delayed Fission Gamma-ray Characteristics of 232 Th, 233 U, 235 U, 238 U, and 239 Pu

Delayed fission gamma-rays play an important role in determining the time dependent ionizing dose for experiments in the central irradiation cavity of the Annular Core Research Reactor (ACRR). Delayed gamma-rays are produced from both fission product decay and from activation of materials in the core, such as cladding and support structures. Knowing both the delayed gamma-ray emission rate and the time-dependent gamma-ray energy spectrum is necessary in order to properly determine the dose contributions from delayed fission gamma-rays. This information is especially important when attempting to deconvolute the time-dependent neutron, prompt gamma-ray, and delayed gamma-ray contribution to the response of a diamond photo-conducting diode (PCD) or fission chamber in time frames of milliseconds to seconds following a reactor pulse. This work focused on investigating delayed gamma-ray characteristics produced from fission products from thermal, fast, and high energy fission of Th-232, U-233, U-235, U-238, and Pu-239. This work uses a modified version of CINDER2008, a transmutation code developed at Los Alamos National Laboratory, to model time and energy dependent photon characteristics due to fission. This modified code adds the capability to track photon-induced transmutations, photo-fission, and the subsequent radiation caused by fission products due to photo-fission. The data is compared against previous work done with SNL- modified CINDER2008 and experimental data and other published literature, including ENDF/B-VII.1. The ability to produce a high-fidelity (7,428 group) energy-dependent photon fluence at various times post-fission can improve the delayed photon characterization for radiation effects tests at research reactors, as well as other applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Fabrication and Characterization of Surrogate TRISO Particles Using 800μm ZrO 2 Kernels

In support of fully ceramic microencapsulated (FCM) fuel development, coating development work is ongoing at Oak Ridge National Laboratory (ORNL) to produce tri-structural isotropic (TRISO) coated fuel particles with both UN kernels and surrogate (uranium-free) kernels. The nitride kernels are used to increase fissile density in these SiC-matrix fuel pellets with details described elsewhere. The surrogate TRISO particles are necessary for separate effects testing and for utilization in the consolidation process development. This report focuses on the fabrication and characterization of surrogate TRISO particles which use 800μm in diameter ZrO 2 microspheres as the kernel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Validation Data Acquisition in HTTF during PCC Events

A validation experimental campaign was conducted in an Integral Effect Test (IET) facility of High Temperature Gas Reactors (HTGR), the High-Temperature Test Facility (HTTF) at Oregon State University (OSU). The HTTF simulates Depressurized and Pressurized Conduction Cooldowns (DCC and PCC). This campaign required the development of a new laser spectroscopic diagnostic to measure velocity in the challenging conditions of the HTTF: low speed (~1 m/s) gas flows at HTGR prototypical temperature and 1/10th pressure. This was a collaborative effort between co-PIs at The George Washington University (GW), Oregon State University (OSU), and NASA Langley Research Center. The main accomplishments of this project include the record for dynamic range for velocimetry, highest accuracy obtained with this technique, successful deployment in an IET leading to new validation matrix for CFD. These are detailed below and in manuscript appended to this executive summary. For this project, we introduced a new class of laser spectroscopic diagnostics to Thermal-Hydraulics to measure velocity of gases; furthermore, the diagnostic was demonstrated in-situ in an IET during DCC events. In such scenarios, particles used in mainstream techniques, like Particle Image Velocimetry (PIV) are not appropriate as they settle down too rapidly and also contaminate the experimental facility. Molecular tracers stay mixed within the working gas and can seed the flow in a technique called Molecular Tagging Velocimetry (MTV). In MTV a molecular tracer is photo-dissociated by a first (write) laser into radicals or molecules. The pattern created by the write laser is then interrogated with planar laser-induced fluorescence (PLIF), the read pulse(s), which are recorded with a camera. The pattern is probed and matched at two times (interval or probe time, dt), resulting in a time-of-flight velocimetry technique. This project demonstrated a new application range for MTV in gases. MTV has been extensively used for high-speed gas dynamics: molecules do not suffer from inertia associated with particles in regions of high accelerations (such as shock waves). For high-speed flows, the probe time is on the order of 0.1-10 µs. In time-of-flight techniques, the velocity is reconstructed from the recorded displacement, dx, between time interval dt: U = dx/dt. To first order, the resolution of the pattern matching is typically a (fixed) fraction of a pixel; therefore the resolution of the velocity measurement increases with dx and, for a given flow velocity, dt. For the low-speed flows of interest here, this requires long probe time, and therefore finding tracers that are stable for a long time. After an extensive literature survey we identified 6 tracers that would be suitable. We also worked with Spectra Physics to design a custom laser system that is flexible enough to probe these various tracers. The final instrument is made of 4 lasers that are integrated into a compact and movable cart to facilitate the deployment of the instrument off-site. One unique aspect of the laser system is the tunable dye laser (needed to do PLIF of the radicals). By being pumped by two Nd:YAG lasers, this laser is able to provide two pulses with any probe time (dt) for wavelengths between 200 and 900 nm with a linewidth of 3 pm. Accommodating the dual-pulse capability required customizing the pump lasers and the dye laser, making this system unique. Additionally, a custom UV-intensified CCD camera capable of frame-straddling and exposure time as short as 10 ns was acquired. More than two-third of the instrument cost (~$200k) was covered by a cost sharing from the PI’s startup funds. Of the 6 potential tracers identified, two were tested extensively (H2O and N2O) at high-temperature and elevated pressure at GW prior to conducting the tests at OSU. For both tracers, an Ar-F laser (193 nm, 10 ns, 10 mJ/pulse) photo-dissociates trace amounts (0.1 – 5% molar fraction) of the gases into OH and NO. OH is probed at 281.905 nm and NO at 226.186 nm (4-30mJ, 10 ns pulse). NO is very stable in inert environment and we could reconstruct velocities precisely with probe delay times as long at 40 ms, compared to 2 ms for OH, resulting in increased precision. Therefore, for the tests in the HTTF, N2O was selected as the final tracer and its optimal concentration was determined to be 0.5% molar. In the GW laboratory, we obtained a velocity measurement precision of 4 mm/s at conditions mimicking the HTTF tests. We also demonstrated the benefit of molecular tracers over particles and the flexibility of our instrument by continuously measuring velocity from 1000 to 0.1 m/s in a blow down test. This represents a dynamic range on velocity of 80 dB, which is unprecedented for velocimetry! Once the diagnostic was optimized, two two-week campaigns were conducted in the HTTF. This experimental campaign required numerous practical solutions to successfully and rapidly deploy this very sensitive table-top diagnostic in-situ. This was accomplished thanks to very extensive planning from all parties involved in the project. Due to constraints on availability of the HTTF, only adiabatic DCC tests were conducted. Specifically, the velocity was measured at the exit of the hot leg into the Reactor Cavity Simulation Tank (RCST). We were able to record the gas velocity associated with the air ingress for long times (up to one hour). Depending on the initial conditions (density of the gases in the Reactor Pressure Vessel and the RCST) the lock exchange lasted 30-60 s past the valve openings. Large vortices were identified within the shear layer at the interface of the cross flow. These vortices are an indication of shear-driven instability between the counter streams of different density. After the lock-exchange, a small residual flow was still present and its strength decayed exponentially. It finds its source in a buoyancy difference between the gases in both vessels due to the mixing and diffusion of the gases in the cross-over duct and the RPV lower plenum. We did not expect to capture this part of the transient, which would have been less significant (if not negligible) in the presence of a heated core; the precision of our measurement in the HTTF was 6 mm/s, which exceeded our expectations for such a complex test! Using commercial software (Fluent) we simulated the experiment with Computational Fluid Dynamic (CFD) using Reynolds Averaged Navier-Stokes (RANS) turbulence models. We were not able to accurately reproduce the shear layer instability, mixing, and diffusion captured in our experimental data, even with large 3D meshes (>10 M points). We attributed this to the inadequacy of the RANS models for the flow of interest: our experimental Reynolds number was too low and was outside the applicability range of the RANS models. Instead, we should have employed CFD with less modeling, such as Large Eddy (LES) or even Direct Numerical (DNS) simulations. Unfortunately, we did not have the computational resources to do so. Finally, it should be noted that this experimental technique holds great promises for measuring velocity in thermalhydraulics. MTV has an application domain (highlighted in the narrative) that is complimentary to more established (and simpler) techniques, such as PIV. It has attracted the attention of several thermalhydraulics researchers and sponsors to measure velocity in harsh environments. The spectroscopic nature of this technique also enables to expand it to measure temperature and concentration simultaneously to velocity. In lieu of a narrative, 4 detailed journal publications are appended to this executive summary to describe the various development and accomplishments during this project. They include 2 published manuscripts (Experiments in Fluids and Measurement Science and Technology), and two invited contributions following the NURETH-17 conference (Nuclear Science and Technology). Our current Nuclear Engineering and Design draft is not included at this time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Experimental Investigation and CFD Analysis of Steam Ingress Accidents in HTGRs

This study focuses on the short-term phenomena due to steam ingress following an SG tube rupture accident in HTGRs, instead of covering a long-term scenario induced by a small amount of steam ingress during plant lifetime. Our ultimate goal is to better understand water/steam ingress accidents in HTGRs, where the Rankine cycle is adopted as the power conversion cycle. General Atomics’ 350 MWth Modular High Temperature Gas-cooled Reactor (MHTGR) was chosen as the prototypic reactor design. The models of steam nuclear graphite oxidation proposed in the literature are surveyed, and will be examined, validated or improved if needed. Furthermore, this study aims to develop and validate a predictive tool for steam ingress accidents. To measure the steam concentration evolution in an integral effect test (IET) facility, a tunable diode laser absorption spectroscopy (TDLAS) measurement system was developed. The TDLAS measuring system is a non-intrusive and non-perturbative diagnostic technique that can be used to measure the steam temperature and concentration at various conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Nuclear Metal Fuel Casting Modeling (Annual Report 2012)

In fiscal year 2012, FLOW3D was used to develop an understanding of the metallic nuclear fuel casting processes. There were two major accomplishments for FY12, 1) a rank-order list of the most important material properties to measure, 2) simulation of alternative mold-fill methods. We performed a design of experiments study to quantify several material properties and their impact on effective mold fill, for both circular and annular fuel rod geometries. The most important fuel material property to be measured is the wetting angle of the molten metal on the coated mold material (coated graphite). The second most important property to measure is the surface tension of the molten metal in argon gas at the design mold pressure. The magnitude of these material properties had the largest effect on whether the mold casting processes produced large voids in the fuel rods. The nominal mold fill method used in the above study was the gravity assisted top-fill method of the Idaho National Laboratory Bench Scale Casting facility. Alternative mold fill methods were investigated and using a bottom-fill method shows promise. In the bottom-fill method, the molten metal is injected into the bottom of the mold and the argon cover gas is allowed to escape through vents at the top of the mold. This allows for a smooth, uniform fill of the mold, where as the top-fill geometry has a counter-flow condition of the molten metal and argon cover gas. The mold is initially filled with argon, and using a top fill method, the molten metal must flow downward into the mold and the argon is displaced by the metal and must flow upward out of the mold. This counter-flow condition creates large voids in the cast because bubbles of argon get trapped as the metal flows into the mold. We also investigated various venting schemes for the top fill casting process and this also shows some promise to mitigate the void formation. In fiscal year 2013, we will continue to investigate mold fill approaches. We will use the results of the FY2012 work and begin investigating the segregation of binary alloys for these casting approaches and recommend separate effects testing requirements. If data becomes available from the BSC, we will calibrate the computational models and identify optimum casting parameters. Finally, we will study the feasibility of extending the calibrated model to ternary flow problem by developing appropriate scaled models.

1.02.02.02↗

Post-Irradiation Examination of Low Neutron Dose Metal Fuel Parallelepipeds

Metal fuel samples were subjected to series of low-dose irradiations at the High Flux Isotope Reactor (HFIR) Hydraulic Tube (HT) Facility in FY 2012. The objective of these irradiations was to further fundamental scientific understanding of phenomena involved in the evolution of metal fuels microstructures under irradiation based on separate effects testing of various factors including elemental composition, temperature, and neutron fluence. These data would then be used to inform theoretical models for in-service performance of metallic fuels. Because metallic fuels are generally considered for fast spectrum reactor designs, specialized irradiation capsules, referred to as “rabbits,” were designed with gadolinium thermal neutron shielding surrounding the samples. This design resulted in a fast-neutron spectrum comparable with the nominal in-reactor conditions for metal fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Utility Combined Heat and Power (CHP) Programs

This issue brief provides an overview of energy efficiency portfolios offered by electric and gas utilities that include CHP incentive programs, with examples of program drivers, structures, and eligibility criteria. The document discusses best practices for designing and administering a utility CHP incentive program, including cost-effectiveness tests and market outreach strategies.

24 POWER TRANSMISSION AND DISTRIBUTION↗

Ring Model Development and Validation for Prismatic HTGR Core Thermal-hydraulics and Safety Analysis

Because of the complex core geometry, prismatic high temperature gas-cooled reactors (prismatic HTGRs) often exhibit complex thermal fluid behaviors during both normal operating and transient conditions. Most HTGR designs rely on passive safety system for decay heat removal, such as the reactor cavity cooling system (RCCS). During postulated accidents like Pressurized Conduction Cooldown (PCC) event, the decay heat is first radially transferred from the core region to the reactor vessel outer surface, then to the RCCS cooling panels. The peak fuel temperature is controlled by heat transfer mechanisms with two distinctive characteristic length scales, i.e., the core-wise effective heat conduction and local heat conduction in the fuel pellet scale. As both length scales are essential to determine the fuel temperature, from the modeling perspective, computer codes must be able to capture heat transfer in both scales. This is challenging for both computational fluid dynamics (CFD) tools because of extremely large amount of computation resources required, and for system analysis codes because of the challenge to model the complex core geometry. Under the support of DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, efforts have been pursued to support HTGR technology development and its modeling and simulation needs. There is a particular need for advanced modeling and simulation tools to predict thermal-fluid behavior during safety-related transients. In our previous studies, the ring model was adopted in SAM and further developed to simulate the normal operating condition and a PCC event using the MHTGR-350 design of General Atomics as the reference design. This current work represents a continuation of these previous efforts, and the focus is to critically review and examine simplifications and assumptions made to develop the ring model, and to perform code validation using experimental data from an integral-effect test facility, the High Temperature Test Facility (HTTF) at the Oregon State University. In this study, the test PG-27 from the HTTF test suite was selected for code validation purpose. The test PG-27 is a transient test designed to simulate the PCC event of the MHTGR design. Very good agreements between SAM prediction and experimental measurements were found in both coolant and solid structure temperatures during the transient.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

TEAMER: DAISY Flow-noise Testing (Abstract)

Evaluating impacts of marine energy devices is generally difficult given the dynamic environment where these devices need to be placed. The University of Washington’s (UW) DAISY (The Drifting Acoustic Instrumentation SYstem) is designed to measure radiated noise around marine energy converters operating in energetic waves and currents. In currents, a primary limitation for measurement fidelity at low frequencies (< 100 Hz) is the potential for non-propagating “flow-noise” to mask propagating sound and inflate estimates of the radiated noise from marine energy converters at frequencies that overlap with hearing sensitivities of fish and some marine mammals. While free-drifting measurements help to minimize the relative velocity that produces flow-noise, significant levels were still observed during initial DAISY tests. This motivated the development of a fabric “flow shield” around the hydrophone that disrupts both flow-noise generation mechanisms proposed by the initial tests. First, the flow shield is a source of substantial drag which keeps the hydrophone package moving with the approximate velocity of the surrounding water, compensating for differential wind or current forcing on the surface expression. By minimizing relative velocity around the hydrophone, turbulence shed by the hydrophone is also minimized. Second, the flow shield creates a largely quiescent pocket around the hydrophone, minimizing advection of free stream turbulence over the hydrophone element. Field data collected in these experiments will test effectiveness of these flow shields and provide quantitative data use and deployment.

16 TIDAL AND WAVE POWER↗

Evaluation of Reduced Order Model for HT-9 Creep and Modifications to Current HT-9 Creep Model in BISON

In order to leverage existing reduced order models (ROM) for the modeling of LANL-developed HT-9 cladding constitutive behavior, mechanistic-based models have been implemented into BISON. It is posited that the increased fidelity of the ROM will show improvement over engineering-scale models in terms of comparison against experimental measurements. INL will assist LANL in implementing an HT9 ROM in BISON. Once the HT-9 mechanistic constitutive model can be leveraged in nuclear performance simulations, the numerical results of fast reactor models will be analyzed and compared against the Fuels Irradiation & Physics Database (FIPD) and separate effects tests to attempt a validation process. INL and LANL will work together to perform any necessary improvement that is identified during the implementation and use processes. Successful completion of the milestone will enable the general use of validated mechanistic ROMs for HT-9 cladding in BISON. This report provides a brief introduction to the Los Alamos Reduced Order Model Applied to Nonlinear Constitutive Equations (LAROMANCE) code, mechanical testing used to illustrate its creep and plastic deformation predictions for HT-9 cladding, and a comparison with existing models implemented into BISON. Current models have been modified to more appropriately account for primary thermal creep and results for corresponding testing are included in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Common Approach to Molten Salt Reactor Safety Evaluation [Slides]

The presentation describes the necessary elements of developing a common approach to molten salt reactor (MSR) safety adequacy evaluation. Developing a common approach to establishing adequate safety for MSRs could substantially decrease the developer risks. However, safety adequacy is intimately associated with reactor licensing which is an element with national sovereignty. No reactor class has a common, portable international license. The advantageous performance characteristics of MSRs can be leveraged to decrease the cost and time necessary to develop reasonable confidence of adequate safety. Adequate safety can be demonstrated though either accident prevention or mitigation. MSR have characteristics that may make accident mitigation a preferable safety adequacy evaluation approach. Adequate quality as well as a thorough understanding of accident progression validated through separate and integral effects testing remains central to any safety adequacy demonstration method.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Complete Development of Critical Capabilities for TRISO Fission Product Source Term Calculations and Quantify Mechanisms for Pd Penetration of SiC

Overall fission product (FP) release will be an important consideration for the licensing and deployment of advanced reactors utilizing tristructural isotropic (TRISO) fuels. This work focuses on enhancing and applying the BISON models needed to predict FP transport within TRISO particles and particle failure probability, both of which factor directly into release predictions. Specifically, this report details (1) the development of the models needed to predict palladium (Pd) conservation at the engineering scale and the application of those models to characterize Pd fluxes for input into a mechanistic multiscale model for Pd penetration; (2) the refinement of sorption mass transfer models and the development of models for trapping in porous layers, which were applied and compared to particle scans from AGR-2 to provide proof of concept for a method of particle-scale validation that may reduce uncertainties compared to compact-scale validation using data from integral effects tests; (3) the development of a failure-statistics-informed, mesh-independent methodology for applying smeared cracking, enabling further study of the localized multiphysics behaviors associated with cascading particle failure mechanisms; and (4) the preliminary characterization of those coupled multiphysics particle failure behaviors using smeared, nonretentive diffusivities to provide a baseline for future study and to guide ongoing engineering applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Measuring & Verifying Monopole Gauge Factors Into 50 Ohms

Student poster for HEART 2025. Abstract: Experimental data from multiple methods of measurement as compared to simulation and model data show that free-field antenna characterization methods can be appropriately applied to short, surface mount monopole antennas used in small cavities for shielding effectiveness testing.

Ferreri, Joseph Anthony [Sandia National Laborator↗