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DOE OSTI2023

Summary of INL Integrated Energy Systems Research for the Global National Laboratories Consortium on IES

The DOE Office of Nuclear Energy (DOE-NE) program on Integrated Energy Systems (IES) is led by researchers at Idaho National Laboratory (INL), and work is conducted in partnership with an array of other DOE laboratories, industry, and academia. IES research and development activities are additionally complimented by the DOE-NE Light Water Reactor Sustainability (LWRS) program, where work under the Flexible Plant Operations & Generation pathway supports analysis of opportunities for non-electric applications of current fleet nuclear plants and collaborates with multiple plants on near-term hydrogen production demonstration opportunities. The DOE-NE programs additionally partner with the Hydrogen and Fuel Cell Technologies Office under the DOE Office of Energy Efficiency and Renewable Energy to jointly fund the development of analysis tools, technologies, and nuclear-integrated hydrogen demonstration projects. This document provides a brief, high-level summary of IES work as contribution to the annual report for the Global National Laboratories Consortium on IES.

08 HYDROGEN↗

DOE OSTI2025

Direct Comparison of the Activity and Selectivity of Rh 1 Cu and Ni 1 Cu Single-Atom Alloy Sites for Ethanol Decomposition

Ethanol is an important source of clean hydrogen, acetaldehyde, acetic acid, acetate esters, and light hydrocarbons. Controlling the divergent reaction pathways to these products requires understanding how different active sites influence the elementary steps involved. Herein, we present a combined surface science, theory, and nanoparticle catalysis study demonstrating how two single-atom dopants (Rh and Ni) in a Cu host can distinctively alter the selectivity of alcohol conversion. Specifically, our model studies reveal that ethanol reacts on Ni 1 Cu single-atom alloys to selectively produce acetaldehyde, whereas methane and CO are also formed on Rh 1 Cu single-atom alloys. Interestingly, these different reactivities are in contrast to the behavior of the pure metals as Ni(111) and Rh(111) surfaces favor methane/CO and surface carbon/CO, respectively. DFT calculations of reaction pathways and simulated product desorption based on microkinetic analyses explain these reactivity differences, demonstrating that C–C cleavage leading to methane formation has a lower barrier on Rh single-atom sites. To test the catalytic relevance of these fundamental results we synthesized and characterized supported Ni 1 Cu and Rh 1 Cu single-atom alloy nanoparticles with dopant:Cu ratios of 1:200. Flow reactor results revealed that both Ni and Rh increased ethanol conversion over Cu and that Ni 1 Cu catalysts were >99.9% selective to acetaldehyde, while Rh 1 Cu also produced 0.6%–2.6% of equimolar methane and CO between 433 and 493 K, demonstrating that C–C bond cleavage is enabled by isolated Rh sites. Furthermore, these catalytic results bridge the pressure and materials gaps, and together, this study provides insights into how different isolated dopant sites promote different catalytic pathways.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2026

Ion Cyclotron Heating in a Levitated Dipole Fusion Reactor

OpenStar Technologies is pursuing the levitated dipole (LD) as a highly modular, loosely-coupled system that leverages their expertise in high temperature superconductor (HTS) technology. The next generation experiment at OpenStar, Tahi (Ma¯ori for “first”), will demonstrate the generation and confinement of fast ions in a levitated dipole for the first time. Ion cyclotron range of frequency (ICRF) heating is a leading candidate for energetic ion formation in Tahi. A frequency in the 10 MHz range will be used for H minority heating or D majority heating with waves launched from an antenna located above the floating coil. Unlike a tokamak, where the targeted cyclotron resonance is typically a vertical path through the center of the plasma, in a dipole the resonance location follows a C-shaped path from the separatrix to the center of the plasma. The value of B also varies significantly within the confined plasma resulting in a large number of cyclotron harmonics present in the low field region. Furthermore, levitated dipoles contain a “first closed flux surface” surrounding the floating coil, in addition to the traditional separatrix /last closed flux surface. Simulation e ff orts using full-wave ICRF codes show that ICRF heating of a levitated dipole reactor is feasible using a pair of toroidal current straps phased to launch the appropriate parallel ( i.e. poloidal) refractive index.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

High-Temperature Sodium System Structural Health Monitoring (SHM) Technologies at the Mechanisms Engineering Test Loop (METL)

This report documents the deployment of non-destructive testing (NDT) technologies at the Mechanisms Engineering Test Loop (METL). METL became operational in September 2018 with the mission to provide an ecosystem for Advanced Reactor Development (ARD). METLs flagship facility’s primary purpose is conducting small to intermediate scale tests for Sodium Fast Reactors (SFR). Its resemblance to a commercial SFR’s intermediate heat transport system, prototypic operating conditions, and industrial construction practices/materials provides the overarching benefit of establishing a proving ground for devices of all technology readiness levels as well as commercial off-the-shelf equipment yet to be demonstrated on a sodium system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Continuum Damage Mechanics Modeling of High-Temperature Flaw Propagation: Application to Creep Crack Growth In 316H Standardized Specimens and Nuclear Reactor Components

Predicting creep crack growth (CCG) of flaws found during operation in high-temperature alloy components is essential for assessing the remaining lifetime of those components. While defect assessment procedures are available for this purpose in design codes, these are limited in their range of applicability. This study assesses the application of a local damage-based finite-element methodology as a more general technique for the prediction of CCG at high temperatures on a variety of structural configurations. Numerical results for stainless steel 316H, which are validated against experimental data, show the promise of this approach. This integration of continuum damage mechanics (CDM) based methodologies, together with adequate inelastic models, into assessment procedures can therefore inform the characterization of CCG under complex operating conditions, while avoiding excessive conservatism. This article shows that such modeling frameworks can be calibrated to experimental data and used to demonstrate that the degree of tri-axiality ahead of a growing creep crack affects its rate of growth. The framework is also successfully employed in characterizing CCG in a realistic reactor pressure vessel geometry under an arbitrary loading condition. These results are particularly relevant to the nuclear power industry for defect assessment and inspections as part of codified practices of structural components with flaws in high-temperature reactors.

42 ENGINEERING↗

DOE OSTI2021

Explicit Modeling of Pebble Temperature in the Porous-medium Framework for Pebble-bed Reactors Applications

In this study, an explicit model for pebble temperature calculations in the framework of a porous-medium approach has been developed and implemented in the SAM code. The solid-phase energy conservation equation, including the explicit modeling of pebble temperature, is a multi-scale model, which can predict the macroscopic (pebble-bed) and microscopic (pebble) temperature distributions under both steady-state and transient conditions. Extensive and successful code verifications and demonstrations have been performed for this newly developed model. By explicitly modeling pebble temperatures, this new model addresses a major deficiency of the previously implemented model, which assumed a local solid-phase thermal equilibrium that is not appropriate for pebble-bed reactor design and safety analysis applications.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

A percolating path to green iron

About 1.9 gigatons of steel is produced every year, emitting 8% (3.6 gigatons) of global CO 2 in the process. More than 50% of the CO 2 emissions come from a single step of steel production, known as ironmaking. Hydrogen- based direct reduction (HyDR) of iron oxide to iron has emerged as an emission-free ironmaking alternative. However, multiple physical and chemical phenomena ranging from nanometers to meters inside HyDR reactors alter the microstructure and pore networks in iron oxide pellets, in ways that resist gaseous transport of H 2 /H 2 O, slow reaction rates, and disrupt continuous reactor operation. Using synchrotron nano X-ray computed tomography and percolation theory, we quantify the evolution of pores in iron oxide pellets and demonstrate how nanoscale pore connectivity influences micro- and macroscale flow properties such as permeability, diffusivity, and tortuosity. Our modeling framework connects disparate scales and offers opportunities to accelerate HyDR.

hydrogen↗

DOE OSTI2021

Observation of fully detached divertor integrated with improved core confinement for tokamak fusion plasmas

Integration of divertor detachment with a high-performance (β N ~ 3, β p > 2, H 98 ~ 1.5) core plasma has been demonstrated in DIII-D high-β p (poloidal beta) plasmas associated with a sustained core internal transport barrier (ITB) and an H-mode edge transport barrier (ETB). Such good core-edge integration has been achieved for both neon and nitrogen seeding, for both favorable and unfavorable B-field directions, independently from the impurity puffing locations, though these variations play important roles on divertor characteristics. Compared to the standard H-mode plasmas, the high-β p plasma exhibits a much wider window of detachment compatible with high confinement core. Fully detached divertor plasmas with low plasma temperature (Te < 5 eV), low particle flux, and low heat flux across the entire divertor target plate were obtained by using nitrogen seeding. This detached high-β p plasma is compatible with a newly developed detachment control system which can help optimize the nitrogen gas flow rate. Several features, i.e., the high edge safety factor in the high-β p scenario, impurity injection, closed divertor and reduced heating power requirement due to the high confinement, facilitate the achievement of full divertor detachment at lower density. Instead of degrading global performance, the divertor detachment facilitates the access to an even stronger ITB at large radius with a relatively weak ETB through self-organized synergy between ITB and ETB, leading to sustained high confinement. The strengthening of the large-radius ITB compensates for the ETB degradation associated with divertor detachment. In addition, a weak ETB naturally has smaller edge localized modes (ELMs). In particular, with neon injection, a long-period no-ELM H-mode phase has been achieved simultaneously with high-performance core and partially detached divertor plasmas. Furthermore, these results demonstrate the possibility of integrating excellent core plasma performance with an effective divertor solution, an essential step toward steady-state operation of reactor-grade plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

Mesh-based multiphysics coupling acceleration for fusion neutronics through clustering for fusion blanket applications

Accurate modeling of particle transport within fusion blankets is essential for predicting performance metrics such as heat deposition and the tritium breeding ratio (TBR). However, high-fidelity coupling of thermal fluids from computational fluid dynamics (CFD) to neutronics simulations often incurs significant computational costs due to the complexity of surface intersection calculations in Monte Carlo codes. This paper presents an accelerated multiphysics coupling method for neutronics that utilizes hierarchical agglomerative clustering to map complex material property distributions to a neutronics model. Implemented within the fusion reactor design and assessment (FREDA) framework, the method leverages existing Python packages to automate the creation of clustered geometries for OpenMC. The approach is demonstrated on a sector model of an ARC-class tokamak with an immersion molten salt blanket, and an simple geometry with varying isotopic concentrations. Results show that the clustering method significantly reduces computational burden without compromising fidelity, providing a foundation for agile iteration of neutronics simulations involving multiple coupled material properties.

Bae, Jin Whan [ORNL] (ORCID:0000000326548907)↗

DOE OSTI2020

Assessment of Existing Transportation Packages for Use with HALEU

Commercial light water reactor operators and fuel vendors in the United States are pursuing changes to fuel that include increased 235 U enrichment. Economic studies generally anticipate maximum near-term fuel assembly designs with up to 8 wt.% 235 U. Many next-generation nuclear reactor designs require high-assay low-enriched uranium (HALEU) (19.75 wt.% > 235 U > 5 wt.%) fuel. One necessary element for the commercial-scale use of HALEU is the ability to safely transport large quantities of enriched fuel material in multiple forms. However, there is uncertainty as to whether subcriticality requirements can be satisfied with existing package designs and whether existing critical benchmark experiment data are sufficient to support criticality safety code validation for HALEU transportation applications. This study assesses the potential to use currently licensed transportation packages for the transportation of increased enrichment unirradiated U fuel forms. The assessment uses selected package designs that represent the five categories of fuel form-boiling water reactor pins and assemblies, pressurized water reactor pins and assemblies, UF 6 , U-metal and tristructural isotropic (TRISO) particles, and UO 2 pellets or powder-and focuses on demonstrating subcriticality and identifying benchmark critical experiments appropriate for use in criticality computer code validation. Key quantities of interest that relate to subcriticality are limiting conditions (e.g., optimum moderation), package or package array k eff , package capacity, and package transportation array size. The SCALE TSUNAMI-3D and TSUNAMI Indices and Parameters codes are used for sensitivity and uncertainty calculations and for identification of candidate critical benchmark experiments for code validation. The primary metric for identifying candidate benchmarks is the similarity coefficient, c k . For each fuel form category, a representative package is evaluated. Results provided for each package evaluation include enrichment and packaging limits (e.g., maximum transportation array size as a function of enrichment) and benchmark critical experiment similarity coefficients. Results indicate that there are viable means for increasing enrichments into the HALEU range across the spectrum of fuel forms with differing increase amounts available for different packages. Sources of subcriticality margin to offset increased enrichment reactivity include reduced transportation array size, reduced fissile mass, burnable absorber credit, and safety analysis margin harvesting. For all packages except the DN-30, numerous critical benchmark experiment candidates for validation were identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1999

Instellar Exploration: Propulsion Options for Precursors and Beyond

NASA is considering a mission to explore near-interstellar space early in the next decade as the first step toward a vigorous interstellar exploration program. A key enabling technology for such an ambitious science and exploration effort is the development of propulsion systems capable of providing fast trip times; mission duration should not exceed the professional lifetime of the investigative team. Advanced propulsion technologies that might support an interstellar precursor mission early in the next century include some combination of solar sails, nuclear electric propulsion systems, and aerogravity assists. Follow-on missions to far beyond the heliopause will require the development of propulsion technologies that are only at the conceptual stage today. These include 1) matter-antimatter annihilation, 2) beamed-energy sails, and 3) fusion systems. For years, the scientific community has been interested in the development of solar sail technology to support exploration of the inner and outer planets. Progress in thin-film technology and the development of technologies that may enable the remote assembly of large sails in space are only now maturing to the point where ambitious interstellar precursor missions can be considered. Electric propulsion is now being demonstrated for planetary exploration by the Deep Space 1 mission. The primary issues for it's adaptation to interstellar precursor applications include the nuclear reactor that would be required and the engine lifetime. For further term interstellar missions, matter-antimatter annihilation propulsion system concepts have the highest energy density of any propulsion systems using onboard propellants. However, there are numerous challenges to production and storage of antimatter that must be overcome before it can be seriously considered for interstellar flight. Off-board energy systems (laser sails) are candidates for long-distance interstellar flight but development of component technologies and necessary infrastructure have not begun.. Fusion propulsion has been studied extensively. However, fusion technology is still considered immature, even after many decades of well-funded research. Furthermore, fusion alone does not offer high enough energy density to make it a viable candidate for interstellar propulsion unless propellant can be collected in situ, as was considered by R. Bussard for his interstellar ramjet concept. The current research in investigating these propulsion systems will be described, and the range of application of each technology will be explored.

Johnson, Charles Les↗

NASA NTRS2000

Transportation System Options For The Interstellar Probe Mission

Johnson, Charles Les↗

NASA NTRS2002

Ternary Precursors for Depositing I-III-VI2 Thin Films for Solar Cells via Spray CVD

The development of thin-film solar cells on flexible, lightweight, space-qualified substrates provides an attractive cost solution to fabricating solar arrays with high specific power (W/kg). Thin-film fabrication studies demonstrate that ternary single source precursors (SSP's) can be used in either a hot or cold-wall spray chemical vapour deposition (CVD) reactor, for depositing CuInS2, CuGaS2, and CuGaInS2 at reduced temperatures (400 to 450 C), which display good electrical and optical properties suitable for photovoltaic (PV) devices. X-ray diffraction studies, energy dispersive spectroscopy (EDS), and scanning electron microscopy (SEM) confirmed the formation of the single phase CIS, CGS, CIGS thin-films on various substrates at reduced temperatures.

Banger, K. K.↗

NASA NTRS2011

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

Garber, Anne E.↗

NASA NTRS2011

Experimental Studies of NaK in a Simulated Space Environment

Space fission power systems are being developed at the National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) with a short term goal of building a full scale, non-nuclear, Technology Demonstration Unit (TDU) test at NASA's Glenn Research Center. Due to the geometric constraints, mass restrictions, and fairly high temperatures associated with space reactors, liquid metals are typically used as the primary coolant. A eutectic mixture of sodium (22 percent) and potassium (78 percent), or NaK, has been chosen as the coolant for the TDU with a total system capacity of approximately 55 L. NaK, like all alkali metals, is very reactive, and warrants certain safety considerations. To adequately examine the risk associated with the personnel, facility, and test hardware during a potential NaK leak in the large scale TDU test, a small scale experiment was performed in which NaK was released in a thermal vacuum chamber under controlled conditions. The study focused on detecting NaK leaks in the vacuum environment as well as the molecular flow of the NaK vapor. This paper reflects the work completed during the NaK experiment and provides results and discussion relative to the findings.

Gibons, Marc↗

NASA NTRSDate not supplied

Analysis of CDC Bioreactor Internal Thermal Measurements and Sample Coupon Temperatures

The Center for Disease Control and Prevention (CDC) bioreactor is an integral laboratory tool for the Environmental Control and Life Support Systems (ECLSS) biofilm formation and growth research program. Critical to this research is the need to adjust and maintain various surface temperatures of the coupons housed within the CDC Bioreactor. The purpose of this study was to provide quantitative temperature gradient information when the CDC Bioreactor was operating according to several process scenarios. Two primary process parameters were evaluated. For the first set of test parameters, the liquid level was maintained at 350 mL, with an inlet flowrate of 0.1 mL/min, 1 mL/min, 10 mL/min. The liquid was allowed to gravity drain out of the outlet spout. For the second set of test parameters, the liquid level within the reactor was maintained at 550 mL, with an inlet flow of 0.1 mL/min, 1 mL/min, 10 mL/min and draining intermittently controlled to 0.8 mL/min to maintain the 550 mL level. Due to the placement of the thermocouple in the reactor, a difference in temperature occurred between the coupon surfaces and target Bioreactor temperature when operated according to the first set of test parameters. When the reactor was operated according to the second set of parameters, which resulted in the thermocouple being submerged, the temperature gradient was eliminated. The results demonstrated minimal temperature gradient between the top and bottom coupon surfaces for coupons placed in a single rod within the CDC Bioreactor for both sets of test parameters evaluated. The collection of this information helped to explain previous ECLSS biofilm formation test runs, along with providing guidance on best operating practices for future ECLSS experiments. The placement of the thermocouple also helps to explain the challenge of achieving and maintaining bulk liquid temperatures when biofilm is grown according to the standardized test methods.

CDC Bioreactor↗

DOE OSTI2023

An integrated coupling model for solving multiscale fluid-fluid coupling problems in SAM code

In this study, an integrated coupling method has been developed for solving multiscale fluid-fluid coupling problems in plant-scale safety analysis models in SAM (System Analysis Module) code. In this method, a higher-fidelity multi-dimensional (3D) flow module is used for reactor components of complex flow features (e.g., reactor core) and a lumped parameter one-dimensional (1D) flow module for plant-scale flow loops (e.g., primary loop pipe network), respectively. In this method, the 3D fluid equation/domain and 1D fluid equation/domain are tightly coupled at the residual level and solved simultaneously using the Newton’s method to overcome the convergence issues typically seen in existing approaches like separate domain approach, where the 3D fluid equation and 1D fluid equation are solved separately. Extensive and successful code verifications and demonstrations have been performed for this newly developed method. This new modeling approach significantly simplify the work flow in developing high-fidelity plant-scale safety analysis model, e.g. for pool-type reactors and pebble-bed reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Preliminary analysis of TREAT free-field experiments using openmc

This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗