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33 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Multi-scale modeling of wastage layer formation in metallic fuel cladding

Fuel-cladding chemical interaction (FCCI) is a major concern for U-Zr metallic fuels' performance, primarily due to the formation of a brittle layer (wastage) in the cladding. This brittle layer, resulting from intermetallic compounds between cladding constituents Fe, Cr, and lanthanide fission products, significantly impacts the cladding's mechanical integrity. Recent efforts focus on developing a mechanistic modeling framework to understand lanthanide production, transport to the fuel-cladding interface, and phase transformation to intermetallic phases. A multi-scale computational approach has been used to calculate lanthanide transport rates, with atomistic calculations determining Nd diffusivities through the solid fuel matrix and along pore surfaces. These diffusivities inform a mesoscale model to determine an effective diffusion coefficient, accounting for porosity and infiltration with bond sodium. This effective diffusivity is used in engineering-scale simulations via the BISON fuel performance code, which has been validated against EBR-II and FFTF reactor experiments.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

First-principles investigation of cerium and neodymium diffusion in BCC chromium and vanadium via vacancy-mediated transport

Lanthanide transport plays a crucial role in the performance and longevity of metallic nuclear fuels. This study examines the diffusion behavior of Ce and Nd—two major fission products—in body-centered cubic (BCC) Cr and V, which are potential liner or coating materials for mitigating fuel-cladding chemical interactions (FCCI). Using density functional theory (DFT) calculations and self-consistent mean-field (SCMF) analysis, the vacancy-mediated diffusion coefficients are evaluated. Our findings reveal that Ce and Nd act as oversized solutes and are strongly bound to vacancies in BCC Cr and V, with diffusivities in Cr significantly lower than in V and in hexagonal closed-packed (HCP) Zr, as investigated in our previous work. The activation energies for Ce and Nd diffusion are 3.39 and 3.32 eV, respectively, in BCC Cr, and 2.56 and 2.33 eV, respectively, in BCC V. Analysis of vacancy drag and partial diffusion coefficient ratios indicates a strong tendency for lanthanide enrichment at vacancy sinks in BCC Cr, and to a lesser extent in BCC V, with this effect persisting up to the melting point in Cr and remaining substantial for Nd in V at high temperatures. Under irradiation, the increase in vacancy concentration is expected to enhance lanthanide transport, potentially accelerating interactions at liner-cladding interfaces. Although BCC Cr exhibits relatively low lanthanide diffusivities under equilibrium conditions, the expected segregation tendencies under irradiation suggest that Zr liners may be a more favorable option. Further investigations using rate theory, cluster dynamics, and phase-field modeling are required to quantitatively assess the performance of these materials in reactor environments.

36 - MATERIALS SCIENCE↗
DOE OSTI2026

EDS Analysis of FCCI in AFC-FAST Fuel Pins

The Advanced Fuel Campaign’s Fission Accelerated Steady-state Testing (FAST) program uses metallic fuel pins with small diameters to reach a desired burnup more quickly. This enables accelerated testing of advanced fuel designs and decreases the time between idea conception and commercial usage. Post-irradiation examination is critical in this process, especially with respect to the fuel-cladding chemical interactions (FCCI). In this work, energy dispersive X-ray spectroscopy (EDS) is used to track how elements from the fuel and fission products have diffused through the cladding. Elemental redistribution along the fuel-cladding interface is mapped and FCCI region thicknesses are measured. The correlations between geometry, temperature, burnup, and FCCI thickness are presented.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multilayer Architecture Barriers for High Temperature Nuclear Applications

Innovative thin film multilayer metal/ceramic barrier coatings have been developed by Argonne National Laboratory to protect fuel cladding against different phenomena that can be encountered under the harsh operating conditions of advanced nuclear reactors. One phenomenon is Fuel-Cladding Chemical Interaction (FCCI), which can represent a challenge to the performance of steel-clad metallic fuels in sodium-cooled fast reactors (SFRs), especially when operating to high burnups at elevated cladding temperatures, and long fuel residence time. Another phenomenon is the corrosion of zirconium-based cladding under prototypic boiling water reactor (BWR) conditions and associated cladding during off normal conditions, including loss of coolant accidents (LOCA). This work provides details of the development and testing of a thin film (few microns) multilayer metal/ceramic (CrY/Y 2 O 3 ) coatings on steel cladding and its performance against FCCI. The coating architecture evolved through systematic mechanical testing, radiation tolerance assessments, and high-temperature diffusion studies from a preliminary multilayer concept to an optimized high-performance design. The optimized structure leverages stoichiometric Y 2 O 3 for its exceptional radiation stability, chemical inertness, and ultra-low fission product diffusivity (e.g., Ce diffusivity ~10 -20 m 2 /s in irradiated Y 2 O 3 at 650 °C). This is paired with ductile chromium-rich interlayers (Cr 95 Y 5 ), which significantly enhance fracture toughness and mechanical resilience. Design refinements, including increased ceramic layer thickness and optimized Cr–Y interlayer composition, eliminated fracture propagation and preserved coating integrity under extreme mechanical stress (>10× typical reactor conditions) and prolonged radiation exposure (~300 dpa). Those conditions are well beyond expected conditions during normal operations in SFR, and cover some of the transient conditions as well. In addition, the coating deposition was demonstrated to be applied uniformly to the inner surface of steel tubes representing prototypic SFR cladding. Another phenomenon under consideration in this work is the corrosion of zirconium-based cladding under prototypic boiling water reactor (BWR) conditions and associated off normal conditions including loss of coolant accident (LOCA). Another multilayer thin film coating is considered for mitigation of this phenomenon, where a coating of yttrium aluminum and yttrium aluminum ceramic/metal thin film coating. Preliminary results on the coating performance during high temperature thermal cycling (up to 1000 C) have shown promising performance of the coating. On-going optimization of the coating will continue in FY26.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Fuel-Cladding Eutectic Study of Legacy Fast Flux Test Facility (FFTF) MFF HT9/U-10Zr Metallic Fuel

This report presents the first systematic investigation of fuel-cladding eutectic interaction (FCEI) in irradiated HT9/U-10Zr metallic fuel from the Fast Flux Test Facility (FFTF) Materials Fuels Form (MFF) program, using differential scanning calorimetry (DSC) coupled with scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). Two irradiated fuel cross-sections, MNT07H (9.5 at% burnup, x/L = 0.78) and MNT08H (7.0 at% burnup, x/L = 0.93), were subjected to three successive isothermal annealing rounds (R1–R3) at 820°C for 20 minutes each, yielding a cumulative transient duration of one hour. This study directly addresses a recognized gap in the existing FCEI database, which previously lacked irradiated HT9/U-10Zr data at burnup levels above 8 at%. Two principal findings emerge from the study. First, for both samples, FCEI remained spatially confined within the pre-existing fuel-cladding chemical interaction (FCCI) zone boundaries after R3, with no measurable eutectic penetration into unaffected cladding beyond the original FCCI layer. This self-limiting behavior is consistent with historical Fuel Behavior Test Apparatus (FBTA) results and is attributed to the near-eutectic phase composition of the FCCI zone, which rapidly absorb the available eutectic-forming constituents and then stall penetration once the FCCI zone is consumed. A comparison with unirradiated surrogate data further supports this mechanism: whereas a U–34 at.% Fe sample would be expected to show ~176 µm of iron penetration under comparable conditions, the irradiated samples exhibited only ~20 µm, a discrepancy attributed to irradiation-induced interfacial porosity and pre-existing FCCI composition gradients. Second, for MNT08H, FCEI was observed exclusively on the half of the cladding circumference where pre-existing steady-state FCCI was present, with no detectable FCEI on the opposite half. Three hypotheses are proposed to explain this asymmetry: the inhibiting role of a zirconium-rich rind at the fuel-cladding interface; the chemical sequestration of iron by redistributed zirconium within the fuel matrix; and the persistence of fuel-cladding gaps on the FCEI-free half that preclude direct contact. All three hypotheses require further experimental investigation. The results extend the empirical FCEI database into higher-burnup territory and demonstrate the viability of DSC-based testing as a substitute for the no-longer-available FBTA apparatus. Future work will include additional cross-section testing, compilation of the full FCEI dataset, model evaluation, and DSC testing of ternary fuel compositions to broaden the experimental basis for safety assessment of sodium-cooled fast reactor systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Multi-principal element alloys for fast reactor cladding applications

Given the extensive list of multi-principal element alloys (MPEAs) within literature and the overwhelming number of alloys that can be made from only a handful of elements, this article proposes a methodology for prioritizing alloys for use as cladding within advanced reactors. This paper applies neutronic, mechanical, and chemical assessments to a collection of MPEAs available within literature for use as advanced reactor cladding. The results are compared to a reference design of HT9, a ferritic/martensitic steel, for employed in sodium-cooled fast reactor. Mechanical assessments determined pressure limits for a thin-walled tube pressure boundary, thus relating the material's mechanical properties to a minimum acceptable wall thickness. Neutronic analyses reveal a maximum allowable wall thickness that a given material must meet to provide a level of neutronic economy that is equivalent to than that of HT9. Lastly, each alloying element is compared to typical fission products found in a fast reactor fuels to mitigate deleterious phenomena such as fuel-cladding chemical interactions (FCCI). These analyses indicate that Mo-Nb-Ti-V-based alloys are likely to be advantageous and that the inclusion of elements with a high neutronic penalty (e.g., Hf) could be considered with minimal consequences.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

A robust alloy design (RAD) strategy for next-generation (IV) nuclear fission reactors

Next-generation nuclear reactors demand structural materials capable of withstanding extreme conditions, including high temperatures, intense neutron flux, and corrosive environments. Multi-Principal Element Alloys (MPEAs) have emerged as promising candidates due to their exceptional radiation tolerance, thermal stability, and compositional flexibility. This study introduces a versatile and customizable Robust Alloy Design (RAD) strategy for systematically designing MPEAs for GEN-IV reactor fuel cladding. The RAD framework integrates nuclear-relevant selection criteria, empirical parameter assessments, and high-throughput CALPHAD simulations to efficiently narrow compositional space and identify stable alloys. A unified RAD score developed for the first time, combines key performance metrics, including fuel-clad chemical interaction (FCCI), neutron absorption cross-section (NAC), valence electron configuration (VEC), and melting point factor (MPF), into a flexible ranking system adaptable to reactor-specific priorities. Among 724 candidates, V555(5Al–5Cr–5Fe–85V) emerged as the top alloy, validated experimentally with a homogeneous single-phase BCC microstructure and superior mechanical properties (nano-indentation: 3.389 ± 0.258 GPa; Vickers hardness: 240 ± 6.7 HV), significantly outperforming Zircaloy-4 and V-4Cr-4Ti. Importantly, the RAD strategy is not limited to nuclear applications; its customizable weighting system enables scalability to other extreme environments. This adaptability positions RAD strategy as a versatile tool for advanced materials design across multiple industries.

Alloy design↗
DOE OSTI2025

AFC Metal Fuel Workshop Presentations

Compilation of PowerPoint slide decks from presentations given at the 3rd Annual AFC Metal Fuels Workshop. The purpose of this workshop is to bring the national metallic fuel stakeholders together to discuss status and needs of metallic fuel and review the DOE AFC Metal Fuel R&D. Desired Outcomes - Advance Metallic Fuel Technology: 1. Stakeholder Community Building 2. National Program R&D Accountability, Planning Support, and Increased Stakeholder Advocation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Radiation-reduced diffusion of Nd in bcc Fe

Here, this study combines ab initio calculations and kinetic Monte Carlo simulations to investigate the bulk diffusion of Nd, an exemplar lanthanide element, in bcc Fe. Contrary to the conventional belief that irradiation accelerates diffusion processes, our study reveals that irradiation can unexpectedly reduce the diffusion of Nd due to the strong binding between substitutional Nd atoms and vacancies in bcc Fe. Combined with the high concentration of nonequilibrium vacancies, this promotes the abundant formation of Nd+3Va defect complexes containing a single Nd substitutional atom and three vacancies, accompanied by the depletion of Nd-vacancy pairs that mediate the diffusion of Nd.

Density functional theory (DFT)↗
DOE OSTI2025

Chemical interaction and compatibility of uranium mononitride and alumina forming austenitic stainless steel

Uranium mononitride (UN) and alumina forming austenitic (AFA) stainless steel are a potential fuel-cladding combination for the lead-cooled fast reactor (LFR). Chemical compatibility between UN and AFA steel needs to be verified before implementation in a nuclear reactor. Diffusion couple experiments at 823 K and 1023 K were conducted for nonirradiated UN samples in contact with as-cast (no thermally grown Al 2 O 3 ) and preoxidized (with thermally grown Al 2 O 3 ) AFA for 500 and 1000 h in an inert environment. Preoxidized AFA exhibited little to no interaction with all UN samples tested at both 823 K and 1023 K, displaying the stability and capability of the Al 2 O 3 layer to prevent chemical interaction and inter-diffusion with UN. Chemical interaction occurs between UN and as-cast AFA. At 1023 K, an aluminum and nitrogen rich phase (likely AlN) formed along the interface of as-cast AFA and UN samples. At 823 K the AlN phase was not prominently observed due to the reduced diffusivity of aluminum through AFA. The aluminum and nitrogen-enriched phase was also observed in a high temperature pressure-assisted test sample of UN and as-cast AFA thermally treated at 1373 K. Finally, in UN samples doped with a low weight percent of UO 2 (< 3 wt%), AlN was not detected along the interface at either temperatures, and an Al 2 O 3 layer likely formed along the interface and prevented further chemical interaction between UN and as-cast AFA.

36 MATERIALS SCIENCE↗
DOE OSTI2025

INL_intern_poster_RashedA

Advanced nuclear reactors often operate at elevated temperatures, novel fuel with different chemical and thermal properties that may interact differently with cladding, new coolants (e.g., molten salt, liquid metal, or sodium) that are chemically aggressive, requires cladding materials with superior oxidation, corrosion, and creep resistance. Developing fuel cladding for advanced reactors is essential, as the cladding plays a critical role in reactor safety, performance, and longevity. In this study, innovative Non-Concentrated Alloys (NCAs) were developed as novel fuel cladding materials for advanced nuclear reactors. Multiple elements were chosen for the alloys. NCAs were manufactured and tested mechanically, showing promise with the arc-melting method, and setting a clear path for the next step of NCAs' development as a fuel cladding for advanced nuclear reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Formation of uranium nitride nanoparticles via mechanical alloying of uranium-molybdenum alloy fuels in gaseous nitrogen

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and fuel loading capability. However, U-Mo systems are susceptible to irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interactions. To address these shortcomings, this research investigated the formation of uranium mononitride (UN) nanoparticles within a 90 wt% U/10 wt% Mo (U-10Mo) matrix to act as a prospective defect sink for fission products at nanometric hetero-interfaces. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen atmosphere. Variations of the milling process investigated included media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. Characterization of the fuel microstructure was completed using light element analysis, X-ray diffraction, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. UN nanoparticles measuring 1–5 nm in radius were observed in the U-Mo matrix as early as 1 h into the mechanical alloying process. Milling time in excess of 10 h was found to lead to deleterious effects induced by the stainless-steel milling media.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1991

Fuels irradiation testing for the SP-100 program

An SP-100 fuel pin irradiation testing program is well on the way to providing data for performance correlations and demonstrating the lifetime and safety of the fuel system of the compact lithium-cooled reactor. Key SP-100 fuel performance issues addressed are the need for low fuel swelling and low fission gas release to minimize cladding strain, and the need for barrier integrity to prevent fuel/cladding chemical interaction. This paper provides a description of the irradiation test program that addresses these key issues and summarizes recent results of posttest examinations including data obtained at 6 atom percent goal burnup.

Makenas, Bruce J.↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1973

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Collins, J. F.↗