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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 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Physics-informed Estimation of the Covariance Matrix for Various Neutron Spectra

A method for estimating covariance matrices which capture the uncertainties in calculated reactor spectra has been developed. This method is based on perturbing the parameters of a physics-based analytic model fitted to a calculated spectrum. The covariance of the perturbed analytic spectra imposes energy-dependent correlations due to the physics of the neutron processes in the reactor, i.e., a fission component, a 1/E down-scatting component, and a thermal Maxwellian component. An analytic model is developed which is shown to produce good fits to several reactor environments. The covariance matrices produced via this method are then used as the prior spectrum in STAYSL least squares spectrum adjustment where it is combined with integral metrics, such as activation measurements, to produce a high-fidelity neutron spectrum characterization. It was concluded that the methodology showed agreeable results for the ACRR free-field spectrum adjustment in STAYSL resulting in a 𝜒 2 value of 2.21 (per degree of freedom), but further work is needed to describe scattering and interface regions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

A Workflow to Optimize Fast Neutron Irradiation in A Thermal Neutron Spectrum Test Reactor Leveraging Open-Source Tools

The Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) is one of the key nuclear engineering research and testing facilities within the US Department of Energy (DOE). The ATR is one of few high-power research reactors in the world with different application including accelerated testing of nuclear fuel, materials irradiation in a very high neutron flux environment, and medical radioisotope production [1]. Also, the ATR offers opportunities for testing fast spectrum fission and fusion reactor materials. The key challenges in this area are in further detailing and optimizing a fast spectrum environment within a thermal test reactor. This challenge involves researching, developing, and testing novel concepts for the multiplying of neutron populations into ever higher energy spectra in high flux test reactors like ATR. The main objective of this work is to investigate candidate materials for establishing a fast neutron experiment irradiation in thermal neutron spectrum test reactors which can be accomplished by filtering thermal and epithermal neutrons and boosting fast neutrons at designated irradiation positions. However, adding these filters will render the neutron spectrum and the criticality of the system. The selection of the thickness and material layers should be accomplished by developing an optimization design algorithm that is applicable for ATR to enhance the fast neutron spectrum irradiation utilizing high-fidelity Monte Carlo methods along with advanced machine learning capabilities. This paper presents workflow for design optimization to enhance fast neutron irradiation in the ATR. The workflow leverages open-source tools to develop an algorithm that is viable to ATR and can be leveraged in other reactors. The following sections discuss the development of the experiment design optimization workflow and its application to ATR irradiation positions.

42 - ENGINEERING↗
DOE OSTI2026

Leveraging the High Flux Isotope Reactor for nuclear fuel development: a review of experiments, facilities, and capabilities

Materials testing reactors (MTRs) have been used to develop in-core nuclear fuels and materials since the outset of the nuclear power industry. However, the closure of prominent MTRs worldwide and protracted construction timelines for new facilities have increased reliance on existing infrastructure for near-term irradiation testing needs. One facility that can support these needs is the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. HFIR boasts the highest steady-state neutron flux in the Western Hemisphere and, among other roles, has been used to rapidly administer high fluences on fuels and materials for fission and fusion reactor applications. This paper reviews HFIR facilities and infrastructure, fuel-bearing irradiation experiments conducted in HFIR, and select nonfueled experiments that demonstrate advanced techniques transferable to fuels experiments. Collectively, these examples underscore HFIR's potential role as a nuclear fuels testbed supporting both the existing reactor fleet and advanced reactor fuel development.

Fuel qualification↗
NASA NTRS1976

Plasma core reactor applications

Analytical and experimental investigations were conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration. Axial working fluid channels are located along a fraction of each cavity peripheral wall. Results of calculations for outward-directed radiant energy fluxes corresponding to radiating temperatures of 2000 to 5000 K indicate total operating pressures from 80 to 650 atm, centerline temperatures from 6900 to 30,000 K, and total radiated powers from 25 to 2500 MW, respectively. Applications are described for this type of reactor such as (1) high-thrust, high specific impulse space propulsion, (2) highly efficient systems for generation of electricity, and (3) hydrogen or synthetic fuel production systems using the intense radiant energy fluxes.

Latham, T. S.↗
DOE OSTI2024

SCALE Analyses of Scenarios in the Molten Salt Reactor Fuel Cycle

In support of the US Nuclear Regulatory Commission non-light-water reactor fuel cycle demonstration project, the capabilities of SCALE 6.3.1 for radionuclide characterization, criticality, and shielding were demonstrated through several scenarios in a molten salt reactor (MSR) nuclear fuel cycle. Three scenarios were selected for this study, with the molten salt breeder reactor (MSBR) serving as the reference design. The original thorium-based fuel salt used in MSBR was replaced with a 235 U-enriched fuel salt to better reflect anticipated future MSR concepts. The first scenario focused on the fuel salt preparation stage. Criticality analyses, using SCALE/CSAS-Shift and SCALE/SAMPLER, examined potential critical configurations in the fresh fuel salt container based on variables such as 235 U enrichment, UF 4 molar fraction, temperature, and container geometry. The second scenario explored the release of fission products during reactor operation in which the tritium buildup in the primary fuel salt was quantified through depletion calculations using SCALE/TRITON with continuous fuel makeup and fractional fission product removal. This scenario also assessed radioactivity levels in the off-gas system (OGS), quantifying radiotoxic nuclides that could be released into the environment. SCALE/SAMPLER was employed to perform uncertainty quantification to identify key operating parameters influencing the buildup of radiotoxic nuclides in the OGS. Additionally, radiation dose rates were calculated using SCALE/MAVRIC to estimate the maximum exposure levels within the reactor cell complex during operation. The final scenario evaluated the decay heat and radioactivity of the primary fuel salt collected in the drain tank, as well as the radiation dose rate outside the building from the presence of the drain tank. Overall, the application of SCALE’s various capabilities for radionuclide inventory generation, criticality analysis, and shielding in MSR fuel cycles was successfully demonstrated across these scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Modeling and Simulation of Fuel Dispersal During the Loss-of-Coolant Accident

This document is the compilation of the milestone portion to a larger end of project NEUP report. The executive summary of the modeling portion is provided below: In the event of cladding rupture during a postulated LOCA in a pressurized water reactor, fuel particles, along with fission gases, can be expelled into the reactor core from the fractured fuel rod, a phenomenon referred to as fuel dispersal. The initial stage of fuel dispersal is strongly influenced by the high-pressure ejection of fuel fragments, the size and geometry of the ruptured cladding, and the depressurization history of the fuel rod during the postulated LOCA transient. Depending on the location of the burst orifice relative to the quench front, the dispersal event represents an intricate three-phase flow and heat transfer phenomenon, where high-temperature fuel particles carried by the fission gases interact with the coolant within the narrow subchannels of the fuel assemblies, inducing localized phase change. Given the unique multiphysics nature of this phenomena, the current study develops a dedicated computational framework to predict the mass distribution and cooling of dispersing fuel particles, facilitating post-accident assessment and management of the fuel assemblies. Considering the scale of nuclear reactor applications, a continuum three-fluid model is proposed for simulating the transport of solids within the reactor core. With high-temperature fuel fragments within the liquid media, nucleation sites inducing phase changes are dispersed within the flow domain. Coupled with the fact that the transient dispersal event occurs on different time scales than other three-phase flow applications, this study derives a time-averaged three-fluid flow model without losing generality. The assumptions regarding the continuum treatment of the solid phase and the modeling of fuel dispersal behavior are incorporated to simplify the governing equations and derive applicable closure relations. The computational validation of the model was conducted using adiabatic experimental results obtained from ongoing research at Oregon State University, focusing on characterizing fuel dispersal behavior during simulated LOCA conditions. Settlement characteristics of the solids, quantified by the probability distribution of equivalent particles, closely matched the probability density functions reported in experimental studies. The transport of fuel particles within a scaled 5 × 5 lattice of a pressurized-water reactor rod bundle geometry was modeled through a two-fluid Eulerian framework. The required boundary conditions were evaluated from the fuel performance code BISON in a postulated large-break LOCA scenario. The modeling framework considered solid fuel particles as granular matter, interacting with the gaseous dry steam phase and fission gases through the governing interfacial momentum exchange between the participating fluids. The simulation results provided the volume fraction of the solids obtained at the bottom surface of the enclosing tank geometry. Postulated LOCA leading to fuel dispersal phenomena involves the strong coupling between fuel thermomechanics, cladding deformation, thermal-hydraulics, and fuel particle transport. Incorporation of such a strong coupling in numerical simulation is performed by coupling the multiphysics solvers. In the case of fuel dispersal, a strong coupled simulation can be performed by coupling the BISON code for fuel performance, the TRACE code for system-level thermal hydraulics, and fuel particle transport in Multiphysics Object-Oriented Simulation Environment (MOOSE). For such intricate infrastructure, the MOOSE Framework eases the data transfer between codes. The recent version of MOOSE has incorporated the Navier-Stokes module for the fluid flow. An exploratory exercise was done to gain familiarity with finite volume capabilities in the MOOSE framework to incorporate the Spalart-Allmaras (SA) turbulence model. New finite-volume and auxiliary kernels were introduced to assemble the SA transport equation, compute turbulent viscosity, and evaluate wall distance and diagnostic turbulence terms, fully integrated with existing Navier-Stokes modules. A turbulent lid-driven cavity at a Reynolds number of approximately 10,000 is used for verification. MOOSE shows the robust solver convergence and produces the turbulent features. But it underpredicts the velocity profile and turbulent quantities, emphasizing the need to develop improved SA near-wall treatments (e.g., low-Re corrections or wall functions) as a key direction for future work.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Influence of Surfactants on Noble Metal Liquid-Gas Interface Mass Transfer in Molten Salt Reactors

Noble metals, a group of insoluble fission products (FPs), can accumulate in molten salt reactors (MSRs) and influence system performance and safety through deposition. Upon generation in fission, noble metals are transported by circulating salt and may either deposit on structural surfaces or enter the circulating gas bubble interfaces. While sparging was primarily designed for xenon and krypton removal, the mass transfer of noble metals into bubble interfaces highlights a potential pathway for their transport and partial extraction. The properties of the liquid-gas interface therefore play an important role in noble metal deposition and removal behaviors. Observations from earlier studies suggest that noble metals accumulated at the bubble interface can act as surfactants, making the bubble interface partially immobile. This behavior reduces the mass transfer capacity of circulating bubbles and affects both the overall distribution of noble metals in the loop and their removal efficiency. To investigate the influence of surfactants on noble metal transport, deposition, and extraction behaviors in MSRs, a previously developed and validated species transport model incorporating noble metal interphase mass transfer was used to simulate these behaviors. Two designed cases were studied to predict the distribution of 132Te along the Molten Salt Reactor Experiment (MSRE) loop under different bubble interface conditions. The results show that the mass transfer coefficients at the liquid-gas interface, determined by bubble interface characteristics, significantly influence noble metal distribution within the reactor loop. These findings emphasize the importance of continuously introducing fresh circulating bubbles into the molten salt to improve the removal efficiency of insoluble FPs from the MSR primary loop.

MSR↗
DOE OSTI2025

A MOOSE-Based Model for Fission Product Transport and Source Term Estimation for High-Temperature Gas-Cooled Reactors

Thanks to fuel elements containing tristructural isotropic (TRISO) particles combined with a low core power density and passive feedback mechanisms leading to modest temperature rises in the event of accidental events, high-temperature gas-cooled reactors (HTGRs) offer a high degree of reliability in terms of fission product retention. While the anticipated source term for HTGRs is expected to be very low, it is important to provide a quantitative estimate of radiological releases during nominal and accidental conditions. Here, we propose a computationally efficient mechanistic source term methodology relying on the Multiphysics Object Oriented Simulation Environment (MOOSE) for tracking fission product transport from TRISO particles up to the coolant pressure boundary, as well as modeling the transport and potential deposition of these nuclides inside the reactor coolant loop. The proposed computational scheme is applied to estimate source term inventories for a representative 10-MW(thermal) prismatic high-temperature microreactor and is qualitatively compared against known release fractions. In addition to providing an alternate analysis tool, this MOOSE model can help reactor designers quantify the influence of key design parameters relevant for studies of radiological dose consequences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

ATOMIC AND MICROSTRUCTURAL ORIGINS OF STORED ENERGY RELEASE IN NEUTRON-IRRADIATED SILICON CARBIDE

Here, we employ a combination of advanced synchrotron-based scattering characterization techniques to understand and unravel the atomic origins of the colossal stored energy release in neutron irradiated silicon carbide. The quantification of the neutron irradiation-induced defects, and their impact on the structureproperty relationship is important for the design and safe operation of advanced fission and fusion reactors. Our experimental results show that the atomic structure in the as irradiated samples is significantly perturbed due to a large fraction of vacancy and interstitial type defects that lead to complex microstructures and additional components in the x-ray diffraction and pair distribution function results. We directly correlate the stored energy release to the recovery of the sub-lattices, with pair distribution function analysis highlighting that the carbon interstitial and vacancy-type defects contribute to stored energy more than those of silicon. We find these results to be striking and believe our discoveries to be timely and noteworthy given the technological importance of silicon carbide to the nuclear fission and fusion communities.

Sprouster, D.J.↗
DOE OSTI2025

Assessing Stainless Steel Compatibility in Flowing FLiBe Salt

To address the considerable interest in LiF-BeF 2 (FLiBe) compatibility for fission and fusion reactor applications, static and flowing compatibility experiments were conducted to assess the compatibility with type 316H stainless steel. In static testing at 550° and 650 °C, small mass changes were measured and posttest characterization of the FLiBe showed increased levels of Fe, Cr, Ni and Mn in the salt. Adding Be in the static salt test reduced the dissolution of Fe and Ni. Further, an initial assessment of mass transfer in flowing FLiBe without a Be addition was conducted using a monometallic 316H thermal convection loop (TCL) operated for 1000 h with a peak temperature of 650 °C. Similar to prior results in flowing FLiNaK salt, the 316H specimens exhibited small mass losses in the hot leg. Posttest characterization of the 316H specimens suggested Cr surface depletion in the hot and cold legs and possibly Fe deposition in the cold leg. To further understand this behavior, Cr and Fe dissolution was measured in static FLiBe at 550–650 °C.

316H stainless steel↗
DOE OSTI2026

A dual dynamic shutter system for accelerating ion irradiation sample throughput via lateral gas implantation gradients

Ion irradiation for material performance testing is limited due to its serial nature, which allows for only one value of the implantation (appm) versus dose (dpa) parameter space to be explored for each ion and experiment at a time. While ion irradiation can accelerate the process by up to three orders of magnitude compared to neutron irradiation experiments, the sample throughput for ion irradiation remains relatively low. To address these limitations, a novel capability has been developed at the Michigan Ion Beam Laboratory (MIBL), enabling for the creation of single- and two-dimensional lateral ion implantation gradients using recently installed motorized-controlled ion-beam shutters. This advancement can generate a wide scope of the two-dimensional (H+, He2+) implantation parameter space within a single sample. Integration of this new capability now allows for dual- and triple-ion beam experiments to be performed with full user control over not only the ion implantation depth, but also laterally across the sample by imposing ion implantation concentration gradients, thus providing researchers with a high-throughput means for material testing under various irradiation conditions. Furthermore, recent improvements in MIBL's microbeam ion-beam analysis (IBA) target station now allow for probing these concentration gradients in irradiated alloys with exceptional spatial resolution, down to 10 µm. These two approaches promise to significantly improve ion irradiation capabilities and increase the sample throughput by several orders of magnitude. The application of the shutter technique plus the subsequent microbeam characterization of the imposed implantation gradients are showcased by two proof-of-principle ion-irradiated experiments, one performed on single-crystal Si and the other on the fusion-candidate alloy F82H-IEA. These advancements mark a substantial leap in ion-beam technology, offering researchers a robust, high-throughput method to efficiently investigate candidate alloys with high technological readiness for both advanced fission and fusion reactor applications, in a time- and cost-effective manner.

36 - MATERIALS SCIENCE↗
DOE OSTI2026

In-situ synchrotron X-ray investigation of phase evolution in gas atomized powder enabling manufacturing of nanoscale oxide-dispersion strengthened ferritic steels

Here, we present an investigation of Fe-14Cr-3W-0.4Y-0.4Zr-0.18Ti (wt.%) as a reduced-activation oxide-dispersion strengthened (ODS) ferritic steel, an alternative to the “14YWT” structural alloys designed for nuclear energy applications. Gas atomization reaction synthesis (GARS) was used to produce these Zr-modified powders with a non-equilibrium (metastable) phase that enable a heat treatment (delayed) route for producing nanoscale oxide-dispersion strengthening. In situ synchrotron X-ray diffraction and transmission electron microscopy were used to analyze phase evolution as a function of temperature and time, revealing formation of highly dispersed oxide nanoprecipitate phases at temperatures above those needed for powder consolidation and shaping of components. This would allow fabrication of net-shape parts using conventional powder-processing methods prior to thermal activation of a reaction producing nanocrystalline Y-(Ti, Zr)-O particles, with a size of 20 ± 7 nm. These results can inform processing of tubes, cladding, sheets, and plates for use in advanced fission and fusion reactors.

In-situ synchrotron X-ray diffraction↗
DOE OSTI2024

Materials Assessment for FLiBe Fusion Blankets

This evaluation was prepared at the request of DOE-FES and expands on an earlier evaluation prepared by Wilson, Wiffen and Keiser in 2015 (attached as Appendices A and B). Additional useful information for materials in fission molten salt reactor systems that use FLiBe can be found in a prior NRC report [Busby 2019]. The purpose of this report is to review the considerations that will determine the fusion blanket structural materials that will be suitable for use in a FLiBe breeder-coolant fusion power plant (FPP). Identifying a fusion-relevant structural material that is sufficiently compatible with FLiBe is a key issue to be addressed. The constraints and trade-offs made in structural materials selection include the usual power generation requirements such as mechanical properties, fabricability and durability, which assumes reasonable compatibility with the working fluid. For an FPP, additional constraints include neutron economy (breeding ratio), resistance to irradiation damage and low activation (i.e. waste disposal) characteristics.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Technical report Letter: RAFM, ODS steels and MMLC for Nuclear energy application

The lifetime, thermodynamic efficiency, safety and economic viability of new generation fission and fusion reactor concepts can largely be tied to the mechanical performance and stability of structural alloys under extreme environments. In this context, engineered nano materials could have broad-reaching impact on the future of advanced nuclear fuel-cycle and reactors. These systems are characterized by a large number density of interfaces which are efficient sinks for point defects and moderately biased; therefore limiting the deleterious effects of irradiation. Broadly, nuclear nano-technology deals with the use of the latest engineered-nanomaterials for improving the nuclear power performances and safety in all areas of nuclear energy production to bring new generations of nuclear power units. New advanced fuel assembly designs also have implications for securities and safeguards. To support the readiness for potential future license applications, an understanding of the technologies that would enable new reactor designs in the areas of component performance and domestic safeguards is necessary. This technical report letter work explores the technical issues and potential regulatory considerations associated with developing and adopting fuel claddings made of advanced nano- materials. Specifically three classes of nanomaterials are considered: (i) reduced activation ferritic/martensitic (RAFM) steels, (ii)oxide dispersed steels (ODS) and (iii) multi-metallic layered composites (MMLC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Cost-Effective Thermomechanical Processing of Nanostructured Ferritic Alloys: Microstructure and Mechanical Properties Investigation

Nanostructured ferritic alloys (NFAs), such as oxide-dispersion strengthened (ODS) alloys, play a vital role in advanced fission and fusion reactors, offering superior properties when incorporating nanoparticles under irradiation. Despite their importance, the high cost of mass-producing NFAs through mechanical milling presents a challenge. This study delves into the microstructure-mechanical property correlations of three NFAs produced using a novel, cost-effective approach combining severe plastic deformation (SPD) with the continuous thermomechanical processing (CTMP) method. Analysis using scanning electron microscopy (SEM)-electron backscatter diffraction (EBSD) revealed nano-grain structures and phases, while scanning transmission electron microscopy (STEM)-energy dispersive X-ray spectroscopy (EDS) quantified the size and density of Ti-N, Y-O, and Cr-O fine particles. Atom probe tomography (APT) further confirmed the absence of finer Y-O particles and characterized the chemical composition of the particles, suggesting possible nitride dispersion strengthening. Correlation of microstructure and mechanical testing results revealed that CTMP alloys, despite having lower nanoparticle densities, exhibit strength and ductility comparable to mechanically milled ODS alloys, likely due to their fine grain structure. However, higher nanoparticle densities may be necessary to prevent cavity swelling under high-temperature irradiation and helium gas production. Further enhancements in uniform nanoparticle distribution and increased sink strength are recommended to mitigate cavity swelling, advancing their suitability for nuclear applications.

36 MATERIALS SCIENCE↗
NASA NTRS1973

Shield materials recommended for space power nuclear reactors

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

Kaszubinski, L. J.↗
NASA NTRS2008

NaK Plugging Meter Design for the Feasibility Test Loops

The design and predicted performance of a plugging meter for use in the measurement of NaK impurity levels are presented. The plugging meter is incorporated into a Feasibility Test Loop (FTL), which is a small pumped-NaK loop designed to enable the rapid, small-scale evaluation of techniques such as in situ purification methods and to permit the measurement of bulk material transport effects (not mechanisms) under flow conditions that are representative of a fission surface power reactor. The FTL operates at temperatures similar to those found in a reactor, with a maximum hot side temperature of 900 K and a corresponding cold side temperature of 860 K. In the plugging meter a low flow rate bypass loop is cooled until various impurities (primarily oxides) precipitate out of solution. The temperatures at which these impurities precipitate are indicative of the level of impurities in the NaK. The precipitates incrementally plug a small orifice in the bypass loop, which is detected by monitoring changes in the liquid metal flow rate.

Pearson, J. Boise↗