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DOE OSTI2020

In-situ measurement of irradiation behavior in LiNbO 3

In-situ measurement of LiNbO 3 based surface acoustic wave (SAW) crystal resonator device under irradiation was demonstrated and used to characterize the impact of radiation on physical properties of this material. The resonant frequency of the SAW device was monitored as the output power of the reactor was varied. Upon step increase of the reactor power, a gradual shift in the device’s resonant frequency was observed. This frequency shift initially exhibits a linear growth and eventually reaches an equilibrium value proportional to the reactor power. The observed behavior can be attributed to two competing processes: increase of temperature due to gamma heating or accumulation of irradiation induced defects. In both cases, the response is attributed to changes in the physical properties of LiNbO 3 , particularly the elastic constants. This demonstrated ability to measure materials properties under irradiation is attractive for development of sensors and performing materials science under irradiation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2024

Considerations for Defining G-Values for Aluminum-Clad Spent Nuclear Fuel

Sealed-canister dry storage of aluminum-clad spent nuclear fuel (ASNF) generated by research reactors is an alternative to current storage and disposition pathways as directed by the U.S. Department of Energy. The major challenge faced for this storage approach is radiolytic H 2 generation, including from the aluminum (oxy)hydroxide layers on the surface of ASNF. Experimental and modeling activities have been carried out to characterize the radiolytic yield as part of a DOE-sponsored research program to develop the technical basis for ASNF dry storage. The G-value is a commonly way to report results of radiolysis testing and is defined as the radiolytic yield of a species (e.g. molecular hydrogen) per unit radiation energy deposited into the material system. An independent technical review of the ASNF dry storage technical basis performed by Pacific Northwest National Laboratory raised questions about differences in G-value definitions used for experiments on ASNF surrogates consisting of aluminum samples with adherent (oxy)hydroxides compared to G-values reported in prior literature and how the magnitudes compared between different studies. Material systems resembling ASNF pose complications for measuring/defining G-values to predict the evolution of H 2 in a sealed canister, including i) accounting for radiolytic yields potentially arising from multiple sources, i.e., residual free (vapor), physisorbed, and chemisorbed/chemically bound waters; ii) deciding what portions of the multi-material system to include in the absorbed energy (radiation dose) calculation, considering possible energy exchange between materials as well as measurement limitations, and iii) capturing variations in G-value associated with non-linear yield vs. dose curves and/or dependence on the cover gas. This report summarizes previous literature information on radiolytic H 2 generation and associated G-values from mixed-material systems (generally oxides in contact with water or organic compounds) and from (oxy)hydroxides/hydrates to compare with the definitions and values for ASNF surrogate samples containing adherent aluminum (oxy)hydroxides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Property Measurements of LiF-NaF-KF Molten Salts Doped with Surrogate Fission Products

Measurements of molten salt properties including transition temperatures, phase behavior, heat capacity, density, volumetric thermal expansion, surface tension, viscosity thermal diffusivity, thermal conductivity, and vapor pressure are performed at Argonne. The properties of several fuel and coolant salts including eutectic LiF-NaF-KF (FLiNaK) have been measured. These properties are suitable for use in evaluating reactor performance during startup and early operating conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Estimated Diffusion Out of a TRISO Particle Under Ambient Conditions

Many analyses performing diffusion calculations for tri-structural isotropic (TRISO) fuel focus on either reactor conditions or accident scenarios, both instances of elevated temperatures. However, it is worthwhile to understand the diffusion of different elements out of a TRISO particle at ambient atmospheric temperatures, i.e. atmospheric temperatures that are common on the Earth’s surface. This report will outline the thermal analysis and diffusion calculations and then present the results of a generic TRISO particle.

Ivanusa, Pavlo↗

DOE OSTI2021

Demonstrate the Out-of-Pile Performance of a Real-Time Measurement System to Measure Thermal Conductivity Based on Photo Thermal Radiometry

SUMMARY The goal of this project is to develop a fiber-based instrument to perform in-reactor thermal conductivity measurements of fuels and materials. This instrument is based on photothermal radiometry (PTR) and involves heating a sample locally and measuring the induced temperature gradient by collecting blackbody radiation [2]. Thermal conductivity of the sample is extracted by comparing experimental results with a continuum-based model [3,4]. As a laser-based technique, PTR is a non-contact measurement technique that can be performed remotely and non-destructively. In addition, it has several advantages over other photothermal techniques, making it an ideal approach for in-situ measurement of thermal conductivity of nuclear fuels. Because blackbody radiation increases with emissivity and temperature, the PTR technique works well with unprepared surfaces, and measurement accuracy increases with temperature. Moreover, this approach is capable of measuring samples with irregular or poorly defined boundary conditions, which is a common situation for friable spent fuels.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

ACRRF High-Bay Dose Calculations using MCNP (Part A)

Analytical tools and models have been developed as a starting point for directly assessing dose in the Annular Core Research Reactor Facility (ACRRF) due to reactor operation. Key results include peak dose along the Central Cavity (CC) centerline (beamline) at the cavity level, dose throughout the High-Bay (HB), and dose on the facility roof for partially-shielded reactor operation where the 4” insert is removed from the CC Shield Plug (SP). Model results in the beamline are benchmarked against measured doses from passive dosimetry evaluations. Personnel total (neutron and gamma) dose in the ACRRF HB is calculated using Monte Carlo N-Particle (MCNP). Various CC and SP configurations are analyzed, including unshielded (no SP) and partially shielded (SP installed but 4” insert removed). Novel application of Variance Reduction (VR) techniques, namely the Surface Source Write (SSW) and Surface Source Read (SSR) capabilities in MCNP, enable impressive resolution (in a Monte Carlo modeling sense) of dose throughout much the facility. The VR techniques reduce stochastic error for challenging tallies, with more advanced techniques explored in the companion to this report (Part B) [1]. Supplementary studies (including a verification analysis) and pedagogic evaluations in Part B involve neutron spectra, angular distributions, and the dose impact of facility characteristics. With the SP 4” insert removed and the Lead-Boron (44”) Bucket (LB–44) in the reactor cavity, Total Effective Dose (TED) within the CC beamline is ≈140 rem per 300 MJ of reactor yield (or 3900 rem per hour at 100% Steady-State (SS) power). With no SP (unshielded) and a Free-Field (FF) cavity, TED within the beamline is ≈610 rem per 300 MJ (or 17000 rem per hour at 100% SS power). Due to the predicted collimation of radiation by the reactor pool (and partial SP, if present), beamline dose is much greater than the scattered radiation field surrounding the cavity and reactor tank. Comparisons are made to beamline dosimetry measurements to validate the model. Model predictions agree reasonably well (⪅10%) with measured quantities of neutron fluence, gamma fluence, and spectral metrics. Away from the beamline, comparisons made to previous dose measurements in the HB agreement within an order of magnitude.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2024

Production of Medically Desirable Radioisotopes in the EIRENE Molten Salt Reactor

Radioisotopes play a vital role in nuclear medicine, enabling the performance of diagnostic procedures such as Positron Emission Tomography (PET) and Single Photon Emission Computed Tomography (SPECT), in addition to innovative targeted therapies for selectively delivering cytotoxic radiation doses to tumor cells while minimizing damage to healthy tissue [1]. Beta emitting radionuclides such as 90Y and 131I find wide use in radio-immunotherapy, with the radiopharmaceuticals 90Y-ibritumomab tiuxetan (Zevalin®) and 131I-tositumomab (Bexxar®) having received FDA approval for treatment of non-Hodgkin’s lymphoma through targeting the CD20 surface receptor, which is commonly expressed in many B cell non-Hodgkin’s lymphoma subtypes [2, 3]. 90Y has also been successfully applied for treatment of brain tumors [4, 5] and liver cancer [6]. The favorable chelation chemistry of 90Y enables its use with a variety of ligands for the development of radiopharmaceuticals suitable for targeting different carcinoma types [1].

42 - ENGINEERING↗

DOE OSTI2023

The physics basis to integrate an MHD stable, high-power hybrid scenario to a cool divertor for steady-state reactor operation

Abstract Coupling a high-performance core to a low heat flux divertor is a crucial step for ITER and a Fusion Pilot Plant or DEMO. Experiments in DIII-D recently expanded the steady-state hybrid scenario to high density and divertor impurity injection to study the feasibility of a radiating mantle solution. This work presents the physics basis for trade-offs between density, current drive and stability to tearing modes (TMs) at high β. EC power is crucial to tailor the plasma profiles into a passively stable state, and to eject impurities from the core. Off-axis EC depositions decrease the heating efficiency, but calculated electron heat transport coefficients show that this effect is partially mitigated by improved confinement inside the EC deposition. Additionally, the reduction in pressure is recovered by increasing the density. This favourable scaling of confinement with density was observed in high power plasmas for years, and this work provides a comprehensive explanation. ELITE predictions indicate that a path in peeling-ballooning stability opens up for certain conditions of density, power, q 95 and shaping, allowing the edge pressure to continue increasing without encountering a limit. In the core, calculated anomalous fast-ion diffusion coefficients are consistent with density fluctuation measurements in the toroidicity-induced Alfvén eigenmode range, showing that smaller fast-ion losses contribute to the enhanced confinement at high density. The edge integration study shows that divertor heat loads can be reduced with Ne and Ar injection, but this eventually triggers a cascade of n = 1, 2, 3 core TMs. We can now show that impurity radiation in the core is small and it is not the cause for the drop in confinement at high Ar and Ne injection rates. The overlap between the core TMs is consistent with the loss of pressure as estimated by the Belt model for the coupled rational surfaces. Optimization of these trade-offs has achieved plasmas with sustained H 98y2 = 1.7, f GW = 0.7 and ∼85% mantle radiation. The scenario and its variations at higher density and on- vs off-axis EC heating has been studied as a candidate for an integrated solution for several reactor designs, such as ITER, ARC, and the ARIES-ACT1 case, showing promising results in terms of fusion power and gain.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Atomic layer deposition for surface area determination of solid oxide electrodes

Surface area measurements are important for characterizing porous solids, but commonly used methods such as Brunauer–Emmett–Teller (BET) gas adsorption and computed nanotomography are lacking for certain applications such as the development of solid oxide fuel cell electrodes, which often have absolute surface areas <1 m 2 and contain nanoscale features. Presented is a novel method for the surface area determination of samples with total measurable areas of 1–1000 cm 2 with a standard deviation ± 1 cm 2 , utilizing the atomic layer deposition (ALD) of Al 2 O 3 over microstructurally complex internal porosity. The volume of alumina is then quantified using plasma spectroscopy and converted to an area using the known ALD layer thickness. Under the modest ALD reactor soak times used (8.5 s), the precursor penetration depth is found to be ~50 μm, exceeding the requirement for uniformly coating SOFC functional layers. A model system, (La 0.8 Sr 0.2 ) 0.98 MnO 3–δ /Ce 0.9 Gd 0.1 O 1.95 (LSM/GDC) scaffolds of ~2.7 m 2 g –1 , was measured using the technique and compared against the BET method, and comparable results were obtained but with 1000 times less material needed. The technique is demonstrated for measurements in two example areas of active SOFC research: exsolved (Ni, Fe) nanoparticles on an Sr(Ti, Fe, Ni)O 3–δ electrode and PrO x catalyst nanoparticles infiltrated onto LSM/GDC scaffolds. Finally, the technique may be broadly useful wherever the accurate surface area determination of small absolute quantities of porous ceramic structures on the order of 0.1–50 m 2 g –1 is sought.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Strategies to Recover Easily-Extractable Rare Earth Elements and Other Critical Metals from Coal Waste Streams and Adjacent Rock Strata Using Citric Acid

Maximum rare earth element (REE) extractability from a sample of Middle Kittanning coal seam underclay was previously demonstrated at ~30% of the total REE content with a citric acid solution (Montross et al., 2020). This report further refines the mechanisms of citrate application for leaching of coal seam underclays using an organic acid lixiviant and evaluates strategies to begin to scale this process to industrially relevant volumes. This study evaluates the suitability of citrate leaching solutions for downstream recovery and separation of the REEs via oxalic acid precipitation. The applicability of citric acid solutions to recover ion-adsorbed metals from clay surfaces was evaluated by leaching a prepared sample of kaolinite with known amounts of ion-adsorbed lanthanum. Citrate systematics was further defined through experiments varying the solution pH. Up-scaling investigations proceeded with percolation leaching columns and stirred tank reactors. Two embodiments were run in percolation leaching: a continuous flow-through process and a multi-stage saturated process. The results showed that in certain limited scenarios, low pH citrate solutions could be used to target REEs mineralized within secondary calcium phosphate minerals. Results from the upscaling attempts encountered difficulties in extraction efficiencies. For the percolation columns, hydrodynamic flow was insufficient to recover the extractable REE content fully. For the stirred barrels, mixing inefficiencies likely inhibited the extractability of the REEs. Oxalic acid precipitation experiments also highlighted difficulties in downstream operations. The REEs were removed from the citrate solution with high efficiency, but were accompanied by equally efficient removal of calcium in orders of magnitude of greater amounts. Overall, for citrate leaching to be effective, future studies will have to: 1) carefully consider the feedstock application (i.e., likely better suited for secondary mineralization of calcium phosphate minerals), and 2) overcome the technical barriers of large-scale processing, such as inefficient hydrodynamic regimes due to small crush sizes needed and difficulties in downstream separation and purification. Alternatively, citrate solutions may be better recommended as leaching amendments to other leaching solutions to enhance REE complexation in the solution.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2025

FY25 Status Update on A709 Code Case Testing at ANL

This report provides an update on the status of the creep rupture testing on the precipitation-treated (PT) Alloy 709 samples fabricated from the first, the second and the third commercial heats, in support of the first, 100,000-hour Alloy 709 Code Case development under the American Society of Mechanical Engineers (ASME) Section III, Division 5 for high-temperature reactor construction. In Fiscal Year (FY) 25, no new tests were initiated, while 5 previously loaded tests were completed. This report presents an overview of the creep-rupture data gathered at Argonne National Laboratory (ANL) since FY21 that totaled 52 completed tests. Metallographic images showing crack morphologies in some representative samples ruptured in FY25 are also included. Key conclusions include: 1) the creep behavior of the three heats of Alloy 709 can be well described by the Larson-Miller relationship; 2) higher temperatures and larger stresses lead to shorter creep life and higher minimum creep rate; 3) Alloy 709 is overall very ductile under creep; 4) the crack morphologies near fracture surfaces are dependent on material’s grain size and test conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Characterization of boundary precipitation in a heavy ion irradiated tungsten heavy alloy under the simulated fusion environment

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as leading candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy’s unique balance of inherently high fracture toughness and low sputter yields; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. Further, to alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni + and He + ion irradiations to mimic the expected displacement damage and gas production expected after five years of service as a plasma facing material component. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and an a-W 2 C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation assisted and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material’s overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases.

36 MATERIALS SCIENCE↗

DOE OSTI2026

Integrated Neutronics Modeling for Inertial Fusion Energy Systems: Development and Application to LD-FIRST

Lawrence Livermore National Laboratory (LLNL) is proposing a new Laser Driven Fusion Integration Research and Science Test Facility (LD-FIRST) with the goal of providing an experimental testbed for future Inertial Fusion Energy (IFE) systems. However, IFE systems require detailed and accurate multiphysics modeling to quantify material damage, thermal loading, and tritium breeding within complex chamber environments. This article presents the first step in an integrated multiphysics framework that couples meshed CAD-based geometry within Monte Carlo neutronic simulations to enable high-fidelity analysis of IFE chamber concepts, with future coupling to external codes. The neutronics workflow utilizes OpenMC and its third-party capability to use CAD-based geometries through DAGMC and tally on unstructured meshes with Libmesh to evaluate neutron transport behavior, geometric fidelity, and material performance under reactor-relevant conditions. The use of tailored tallies on unstructured meshes in this framework allows direct transfer without interpolating to CFD simulation tools. Two IFE chambers were evaluated, both conceived by LLNL: HYLIFE-II and Laser IFE (LIFE). This work produced high-fidelity conformal surface and volumetric meshes of the HYLIFE-II and LIFE chambers with mapped spatial insight into material damage, thermal loading, and tritium breeding. The HYLIFE-II model was built utilizing available resources and used as a test case to verify that the neutronics framework can handle complex geometries. The LIFE chamber CAD was provided by LLNL and was the main focus of this work. This work analyzes multiple ternary alloy breeding materials for the LIFE chamber, across different 6 Li enrichments to produce data relevant to the LD-FIRST project. This work also investigates the level of model fidelity for the LIFE chamber, and results show that inclusion of detailed first wall and coolant structures increased the predicted tritium breeding ratio (TBR) by ~30%, highlighting the sensitivity of tritium breeding and the need for a high-fidelity simulation framework for IFE chambers. These developments provide a scalable toolset for the design and optimization of next-generation IFE chambers, forming a solid foundation for future coupled multiphysics analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

A multiscale packed-bed reactor model for sustainable ethylene production via chemical looping oxidative coupling of methane

The rising global warming concerns and shale gas discovery have prompted research in the direction of greenhouse gas (GHG), such as methane, reduction and conversion. Oxidative coupling of methane (OCM) offers a pathway to low carbon-intense valorization of methane while producing ethylene, a chemical regarded as central to the petrochemical industry. Even after decades of OCM discovery, researchers keep understanding the process and underlying chemical reactions in a pursuit to achieve industrial viability for OCM. Here, in general, OCM suffers from low C 2 selectivity, yield and reactor temperature runaways due to highly exothermic nature of its reactions. Computational Fluid Dynamics (CFD) tools help analyze spatial gradients within the reactor to deeply understand the diffusion of species, mass and heat transfer phenomena. Furthermore, challenges associated with scaling up such as hot spot formation and parametric sensitivity can be addressed without having to expend on costly experiments. The current paper presents a multiscale packed-bed reactor CFD model coupled with a chemical kinetic model for the chemical looping OCM. The CFD model includes two scales i.e., macroscale for catalyst bed and microscale for individual pellets. Moreover, a chemical kinetic model based on 10 gas-phase reactions is integrated with the CFD model. An additional surface reaction for the formation of gas-phase oxygen from catalyst surface is added to account for the absence of feed oxygen. The model is calibrated against experimental results. The calibrated model captures trends in CH 4 conversion, C 2 selectivity and C 2 yield within a ± 4.35 % range across a temperature range of 700-900 °C. Moreover, model fidelity is evaluated by varying key computational parameters such as mesh resolution and time step size. The model is also verified by varying the inlet methane concentration and the gas hourly space velocity (GHSV) and comparing the results with literature. A sensitivity analysis and scale-up of the current model is undergoing.

Chemical looping↗

DOE OSTI2025

Computational Modeling of Molten Salt Infiltration and Oxidation in Nuclear Graphite

Graphite is utilized as a moderator and reflector in advanced nuclear reactor designs due to its high thermal conductivity, neutron moderation properties, and resistance to radiation damage. However, its longterm performance and reliability are challenged by degradation mechanisms such as molten salt infiltration in molten salt reactors (MSRs) and oxidation in gas-cooled reactors (GCRs). These mechanisms can compromise the structural integrity and operational lifetime of graphite components, necessitating a more detailed assessment of their physical behavior. This report focuses on the development of computational models for molten salt infiltration and oxidation of graphite to aid the design and performance analysis of graphite components. For molten salt infiltration, a computational framework is developed that couples incompressible Navier-Stokes and phase-field model to simulate the penetration of molten salt into graphite?s interconnected pore structure. Initial model verification is performed using two-phase flows in two dimensions, demonstrating the models ability to capture fundamental physical behavior and agree with analytical solution. This framework is then applied to a realistic IG110 nuclear graphite , where a computed tomography extracted pore geometry is used to analyse the infiltration behavior of FLiNaK molten salt. This model provides insights into how the microstructure and other relevant parameters influence the transport pathways of molten salt into graphite, potentially offering a means to rapidly evaluate a graphite grade?s resistance to infiltration. For oxidation, the report details pore-scale mass and heat transport models, describing the diffusion of gases, reaction kinetics, and thermal effects. Additionally, this report highlights inconsistencies in the existing volume-averaged macroscopic model, particularly in upscaling of reaction kinetics and flux terms, and surface to volume transformations. These inconsistencies suggest that current formulations may not accurately capture the experimentally observed graphite oxidation process, highlighting the need for improved model development. This work advances the development of physics-based computational models for graphite degradation, contributing to improved predictive models for next-generation nuclear reactor designs. Future efforts will focus on refining the infiltration model to address non-physical behaviors and enhance its robustness. Additionally, for oxidation, further studies will employ the principles of volume averaging to rigorously derive the upscaled equations, potentially in collaboration with subject matter experts.

Computational Modeling of Molten Salt Infiltration↗

DOE OSTI2022

Catalytic and electrocatalytic performance of Sr(Ti 0.3 Fe 0.7 Ru 0.07 )O 3-δ for applications in solid oxide fuel cells supplied with ethanol steam reforming mixtures

We discuss how direct utilization of alcohols in Solid Oxide Fuel Cells (SOFCs) can improve the conversion efficiency by using excess heat for internal reforming and simplify the system balance of plant. Thanks to their scarce carbon- formation tendency, perovskite oxide anodes are promising alternatives to Ni-based cermets. Here we show that Sr(Ti 0.3 Fe 0.7 Ru 0.07 )O 3-δ (STF-Ru) exsolution anodes provide both catalytic reforming activity and electro- chemical activity with steam/ethanol (S/E) mixtures at 3:1 and 2:1 M ratio between 700 and 800 °C. Electrolyte- supported scandia stabilized zirconia SOFCs with STF-Ru anodes and Sr 0.95 Ti 0.3 Fe 0.7 O 3-δ (STF) cathodes are reasonably stable in the 3:1 S/E mixture over 20 h at 800 °C, with maximum power density of 195 mW/cm 2 and 140 mW/cm 2 at 0.7 V. Deactivation instead occurs with the 2:1 S/E mixture. Comparatively, cells with Ni-YSZ anodes rapidly coke and fail under the same conditions. Ethanol steam reforming experiments in a fixed bed powder reactor reveal that, above 650 °C, gas phase reactions convert ethanol to CO, hydrogen, acetaldehyde and ethylene, while STF-Ru catalyzes steam reforming. Thermogravimetric and Raman characterizations show that a moderate amount of carbon is deposited on the STF-Ru surface, which is more amorphous than the disruptive carbon which degrades Ni-YSZ.

09 BIOMASS FUELS↗

DOE OSTI2025

Machine-Learned Force Field for Molecular Dynamics Simulations of Nonequilibrium Ammonia Synthesis on Iron Catalysts

Ammonia (NH 3 ) is one of the most important industrial chemicals. The conventional NH 3 synthesis method-the Haber–Bosch process-converts atmospheric nitrogen (N 2 ) into NH 3 using H 2 with an iron (Fe) catalyst. However, this process requires high pressures (100–200 atm) and temperatures (700–800 K) near thermal equilibrium. Recently, Fe-based nanocatalysts have been reported to produce promising NH 3 yields under atmospheric pressures and temperature-modulated nonequilibrium conditions. Understanding the mechanism of nonequilibrium catalysis with programmed temperature variation could help to optimize this fully electrified and less energy-intensive process. Although reactive molecular dynamics (RMD) simulations can be a useful tool to model nonequilibrium catalytic processes, they require the development of accurate force fields (i.e., interatomic potentials). Here, we present a machine-learned (ML) force field within the Deep Potential MD (DPMD) framework, trained using periodic density functional theory (DFT) calculations, to model NH 3 synthesis on Fe catalysts with various surface adsorbates such as *N, *H, *N 2 , *H 2 , *NH, *NH 2 , and *NH 3 . Here, we generated the DFT data from static models of elementary reactions on the most stable (110) surface of body-centered cubic Fe, which then were augmented by data from constant number of particles–volume–temperature (NVT) DFT-MD trajectories at various temperatures. Finally, we utilized the fully optimized ML force field to investigate reaction dynamics at an Fe(110) surface at linearly increasing temperatures using NVT-DPMD simulations. Our simulations indicate that pulsed temperature ramping could prove favorable for NH3 synthesis. For example, we conducted ramping under multiple sets of conditions: (i) from 900 to 1200 K over periods of 0.1–0.3 ns for Fe surfaces precovered with N or NH along with H; and (ii) from 300 to 600 K over 0.1–0.3 ns for Fe surfaces precovered with NH 3 . While our simulations so far are limited to short time scales (very rapid heating), these observations shed light on the mechanism of the high NH 3 synthesis rate achieved in a novel temperature-modulated nonequilibrium catalytic reactor using pulsed heating and cooling.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

PORFLOW Implementation of Vadose Zone Conceptual Model for Naval Reactor Component Disposal Areas in the E-Area Low Level Waste Facility Performance Assessment

PORFLOW models have been developed to implement the proposed conceptual models for the two Naval Reactor Component Disposal Areas (NRCDAs), 643-7E and 643-26E, for the purpose of evaluating dose impacts and producing disposal limits for the E-Area Low-Level Facility (LLWF). New NRCDA conceptual models were proposed by Wohlwend and Butcher (2019) for two areas within/adjacent to the E-Area LLWF used as storage/disposal sites for reactor components from the U.S. Navy since 1987. Naval Reactor (NR) waste is comprised of highly radioactive components consisting of activated corrosion-resistant metal alloy contained within welded thick-walled steel casks, and auxiliary equipment primarily contaminated with lower levels of Activated Corrosion Products (ACPs) residing on the metal surfaces and contained within thinner-walled bolted containers.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗