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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.

61 - RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2025

Platinum and Gold Supported on Transition Metal Nitrides for Hydrogen Evolution in an Alkaline Electrolyte

Here, as the urgency to reduce reliance on fossil fuels increases due to carbon dioxide emissions, hydrogen produced by renewably powered water electrolysis has emerged as a promising technology. Alkaline electrolyzers typically exhibit lower current densities than acidic electrolyzers due to the slow kinetics of the hydrogen evolution reaction (HER) under alkaline conditions. This work developed Pt- and Au-modified transition metal nitride (TMN) thin films for improving alkaline HER kinetics. One monolayer Pt–VN, Pt–Mo 2 N, and Pt–TiN were the most promising thin-film catalysts, with alkaline HER activity approaching that of a bulk Pt foil. Additionally, the Gibbs free energy of adsorbed hydrogen was identified as a useful descriptor for alkaline HER activity on TMN and TMN-supported catalysts and has the potential to guide future studies on TMN-based catalysts for enhancing alkaline HER. For practical applications, the thin-film catalysts were then extended to Pt- and Au-modified TMN powders for alkaline HER. Both 5 wt % Pt/TiN and 2 wt % Pt/TiN powders exhibited lower overpotentials at 5 mA/cm 2 when normalized by the Pt electrochemical surface area than the commercial 5 wt % Pt/C benchmark, suggesting a Pt–TiN synergy that creates opportunities for more cost-effective alkaline HER cathodes. Moreover, 20 wt % Au/Mo 2 N also displayed an enhancement in HER activity when compared to the commercial 20 wt % Au/C benchmark.

58 GEOSCIENCES↗

DOE OSTI2025

Conformal electrochemical deposition of intermetallic AuCu thin films for convergent C–N coupling

A compositionally tunable series of AuCu films is active for electrocatalytic urea generation from carbon dioxide and nitrate.

Gerke, Carter S. [Department of Chemistry, Johns H↗

DOE OSTI2024

Godiva IV central cavity neutron environment characterization with threshold neutron detectors

Godiva IV is a cylindrical fast burst reactor comprised of approximately 65 kg of highly enriched uranium that is operated by Los Alamos National Laboratory and sited at the National Criticality Experiments Research Center at the Nevada National Security Site in Nevada in the United States. Godiva IV is typically operated at delayed critical and in the regime spanning from sub-prompt to super-prompt bursts. Godiva IV is used for sample irradiations, criticality safety demonstrations, dosimetry studies, and for studying super-prompt behavior. In preparation for both an upcoming experiment to reduce uncertainties in the prompt fission spectrum for 235U using threshold neutron detectors, and for future research using Godiva IV, it was desired to exercise the process of the selection of threshold neutron detectors/activation foils, radiation metrology, and the subsequent adjustment of the neutron spectrum. For this exercise, nine high purity threshold neutron detectors/activation foils were irradiated in a Godiva IV burst. The foils were then analyzed using a high-purity germanium detector in the NCERC counting laboratory to determine end of irradiation specific activities for available IRDFF-II reactions. This work summarizes the Godiva IV foil irradiation, radiation metrology results, and adjusted neutron spectrum. The results of this exercise ultimately characterized the neutron environment inside the sample irradiation cavity inside Godiva IV to a higher degree than previously performed, informed decisions for the upcoming larger scale experiment, and will inform future neutron spectrum characterizations at NCERC.

Whitman, Nicholas H.↗

DOE OSTI2025

Godiva IV Thermal Neutron Dosimetry Modeling and Variance Reduction

The transfer of the Godiva IV experiment from the Los Alamos Critical Experiments Facility (LACEF) to the National Critical Experiments Research Center (NCERC) introduced a vastly different experiment room return to the neutron flux. The contribution of the background to the burst neutron energy spectrum is significant in the thermal and epithermal neutron energies. Target materials may be placed in various locations in the Godiva room, or outside of the room, for thermal neutron activation. Modeling of this dosimetry problem in Monte Carlo N-Particle (MCNP) presented a novel challenge compared to previous Godiva IV glory hole irradiation simulations. An advanced dosimetry modeling framework for high efficiency calculations in locations far from the Godiva IV fission source was desired. The mesh-based weight windows and point detector advanced variance reduction techniques in MCNP were implemented and tested using adaptations of the critical experiment benchmark model of the Godiva IV problem. The models were validated against measured activations of Nickel, Indium, Scandium, and Cobalt foils at locations 2 meters from the Godiva IV core. Dosimetry measurements were performed in collaboration with Sandia National Laboratory. The weight windows and point detector variance reduction coupled method resulted in the highest problem efficiency.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2026

Sodium Chemistry Assessment and Initiation of Sodium Exposure Experiment on A709 Commercial Heats and LPBF 316H SS

This report documents sodium chemistry characterization and the initiation of forced‑convection sodium exposure experiments for two structural materials for sodium‑cooled fast reactors (SFRs): Alloy 709 (solution‑annealed and with an additional 775 °C/10 h precipitation treatment) and laser powder bed fusion (LPBF) 316H stainless steel tensile specimens printed with an as‑built surface condition. The sodium exposure experiment was conducted in a forced‑convection sodium loop with two vessels operated at 550 °C and 600 °C, and a cold trap maintained at <130 °C. Carbon activity in sodium was quantified using a nickel‑foil equilibration method. Initial post‑exposure evaluation after ~1,000 h indicates negligible thickness change for Alloy 709 and a small but measurable mass decrease. To isolate thermal effects from sodium effects, parallel thermal‑aging experiments were performed on Alloy 709 at 550 °C and 600 °C. Baseline tensile tests at room temperature, 550 °C, and 600 °C show expected reduction in strength with increasing temperature and serration tensile curves at elevated temperatures, consistent with dynamic strain aging. After 3,000 h thermal aging, Alloy 709 exhibits modest strengthening and reduced ductility at 550 °C. Together, these results establish the loop chemistry conditions and provide the initial corrosion and mechanical property results for ongoing long‑duration sodium exposure of Alloy 709 and LPBF 316H stainless steel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Perovskite/Perovskite Tandem Photoelectrodes for Low-Cost Unassisted Photoelectrochemical Water Splitting

In this project, we aim to address the challenges of achieving efficient and cost-effective unassisted photoelectrochemical water splitting using perovskite/perovskite tandem photoelectrodes. We utilized the following unique approaches to advance our innovation. (1) We developed stable and efficient low-E g (1.2 – 1.4 eV) perovskites as the bottom electrodes. We demonstrated approaches to improve the photovoltaic performance and photothermal stability of low-E g Sn-Pb iodide perovskite solar cells. (2) We developed a low-temperature synthesis route to fabricate wide-E g (>1.8 eV) perovskite top electrodes. By combining theoretical and experimental investigation, we found methods to reduce the formation of defects and dislocations in the wide-E g mixed halide perovskites and suppress halide segregation. (3) We developed a robust metal oxide-based interconnecting layer to integrate two perovskite layers into tandem photoelectrodes. Our monolithically integrated tandem devices feature a high V OC of more than 2 V and a high J SC of more than 15 mA/cm 2 . We showed the new design of the tandem photoelectrodes that is critical to the stable operation of tandem photoelectrodes for unassisted PV/PEC water splitting. (4) We demonstrated a water-impermeable barrier of carbon paste/epoxy/metal foil composite, which can prevent photocorrosion and water ingress of perovskite active layers. This surface protection enabled the operation of perovskite photoelectrodes in water and significantly enhanced the long-term stability of our tandem devices. (5) We conducted standardized PEC characterization in collaboration with NREL and reported accurate determination of solar-to-hydrogen conversion efficiencies of perovskite/perovskite tandem photoelectrodes. At the end of the project, we demonstrated perovskite/perovskite tandem photoelectrodes with STH efficiencies of up to 18% and less than 20% efficiency loss after continuous operation for more than 500 hours in water. Our results demonstrate the potential to develop a low-cost, durable, and efficient water-splitting system that meets the DOE 2026 and 2031 cost targets for hydrogen production.

08 HYDROGEN↗

DOE OSTI2026

Measurement of the 75 As ⁢(𝑛, 2⁢𝑛) cross section at 14.1 MeV at the National Ignition Facility

Here, a new measurement of the 75 As (𝑛, 2⁢𝑛) measurement was performed using the 14.1 MeV neutron pulse produced by fusion experiments at the National Ignition Facility (NIF). The target material consisted of GaAs foils encapsulated in aluminum irradiation containers, along with Au monitor foils, and positioned in the NIF chamber during high neutron yield shots. After irradiation, the GaAs and Au foils were counted with high purity germanium detectors to assess the production of (𝑛, 2⁢𝑛) activation products to determine the flux and the cross section of interest. The measured cross section was 966 ± 71 mb at 14.1 ± 0.37 MeV. This work provides a proof of concept for a platform for performing “ride-along” cross section measurements at NIF to contribute to existing nuclear data as well as measure new cross sections in the future.

and nuclear chemistry↗

DOE OSTI2025

Direct writing of PVBVA/Ti 3 C 2 T x (MXene) triboelectric nanogenerators for energy harvesting and sensing applications

Triboelectric nanogenerators (TENGs) have gained recognition for their potential to convert mechanical energy into electrical energy, making them attractive for applications in healthcare, robotics, and human-device interfaces. However, many TENG devices rely on fluorinated polymers for high charge generation and involve complex fabrication processes, which limit their practicality and environmental sustainability. Here, we developed an eco-friendly composite of poly (vinyl butyral-co-vinyl alcohol-co-vinyl acetate) (PVBVA) and Ti 3 C 2 T x MXene for extrusion printing onto aluminum foil substrates, enabling the additive manufacturing of TENGs. Experimental results indicate that integrating 5.5 mg mL -1 of MXene (P-MX 5.5) into PVBVA resulted in a power density of 760 mW·m -2 , with simultaneous improvements in open-circuit voltage (129 %) and short-circuit current (250 %), demonstrating enhanced charge transfer efficiency. Beyond aluminum foil-based devices, we further explored the fully printed P-MX 5.5 TENG by utilizing silver ink electrodes, eliminating the need for aluminum foil. This fully printed, flexible TENG was successfully used for real-time human motion sensing, demonstrating its ability to detect activities such as walking, running, knee bending, and jumping. Collectively, our additively manufactured and sustainable PVBVA-MXene TENG composites, including both aluminum-based and fully printed versions, show promise for future energy harvesters, sensors, wearable electronics, healthcare, and robotic applications.

Additive manufacturing↗

DOE OSTI2025

Assessment of Thin Plastic Scintillation Detectors for Beta-Particle Measurements at the Advanced Test Reactor Critical Facility

The Fission Wire Measurement System is a custom measurement system designed in the 1960s to measure the beta-particle activity of irradiated uranium-aluminum fission wires. This measurement is conducted to determine the fission rate profile of the Advanced Reactor Test Critical facility. The Advanced Test Reactor Critical facility is an open-pool, low-power test reactor used to qualify experiment configurations and verify core models prior to full-power experiment irradiations in the Advanced Test Reactor. Power distribution measurements in ATR-C use uranium-aluminum wires that are distributed throughout the core to validate simulation and modeling results. These measurements require from 340 to 1500 wires to be irradiated and measured within a 12-hour window. The system consists of 4 measurement channels and one reference channel, each with a 2-pi proportional gas flow detector and the measurement channels each have an automated sample changer. The gas flow detectors are of a custom design for this detector system that use methane gas with a large anode wire compared to modern proportional counters. These detectors, which are nearly 60 years old are irreplaceable. The measurements from these gas detectors are affected by the gas flow rate, atmospheric and line pressure, and are very sensitive to the applied high voltage. Recent improvements have been made to the control and data acquisition system, but the detectors have remained the same. The nature of the measurement of the fission product decay activity is such that the energy spectrum of the signal is changing with time. Thin, 250-um thick, plastic scintillators were commercially obtained as a potential replacement for the gas flow detectors. The original calibration of the uranium-aluminum fission wires was conducted in 1965 using a series of irradiations of gold foils and the wires in a well-characterized thermal neutron field. These measurements provided a time-dependent fission rate conversion factor from the gold foil data to calibrate the fission wires based on the response from the 2-pi proportional gas detectors. Transitioning to the new detectors requires qualification and testing. The sensitivity of the scintillators to changes in the energy spectrum of the fission wires and translation of the calibration factor have been completed. These measurements indicated that the sensitivity of the scintillators over time changes at a different rate than the sensitivity of the gas flow detectors. However, the inverse activity of measurements of both detector types is linear with time. Initial results indicate that the scintillator detectors will be a sufficient replacement for the gas detectors with minor adjustments to the fission rate conversion factor. Replacement of the detectors will improve the fission wire measurements and provide a more stable and reliable measurement system.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI2025

Investigating the effect of screen-printed structured graphite electrodes with low tortuosity for high-capacity and fast-charging lithium-ion batteries

A flexible screen-printed graphite electrode was fabricated as an anode for developing fast-charging lithium-ion batteries with low tortuosity. A homogenous anode ink was prepared by mixing graphite as the active material, carbon black (C45) as the conductive additive, and polyvinylidene fluoride (PVDF) as the binder in N-Methyl-2-pyrrolidone (NMP) solvent. The ink was deposited on a flexible copper foil via a stainless-steel screen consisting of an array of pores, that act as secondary pore networks (SPNs), using the screen-printing process. Lithium-ion battery half-cells were assembled using the printed graphite anode, lithium metal foil as the counter electrode, and 1.2 M lithium hexafluorophosphate (LiPF 6 ) in ethyl carbonate: ethyl methyl carbonate (EC: EMC = 3:7) as the electrolyte. The effect of SPNs on the cell performance was investigated by performing formation, rate and cycling tests on the assembled cells, at different C-rates. It was observed that the cells consisting of SPNs with a pore size of 100 μm and edge-to-edge distance of 100 μm between the pores exhibited significantly higher specific capacities of 168 and 129 mAh/g when compared to reference cells without SPNs, which had capacities of 120 and 85 mAh/g, at high C-rates of 4 C and 6 C, respectively. The cells with SPNs also demonstrated excellent cycling performance with ~ 95% capacity retention after 100 cycles at 2 C.

Fast charging lithium-ion battery↗

DOE OSTI2024

An electrochemical generator for the continual supply of 213 Bi from 225 Ac for use in targeted alpha therapy applications

Bismuth-213 is a radionuclide of interest for targeted alpha therapy and is supplied via a radiochemical generator system through the decay of 225 Ac. Radionuclide generators employ longer lived “parent” radionuclides to routinely supply shorter-lived “daughter” radionuclides. The traditional 225 Ac/ 213 Bi radiochemical generator relies on an organic cation exchange resin where 225 Ac binds to the resin and 213 Bi is routinely eluted. These resins degrade when they absorb large doses of ionizing radiation (>1 × 10 6 Gy/mg), which has been observed when the loading activity of 225 Ac exceeds 2.59*10 9 Bq (70 mCi). Herein we report the development of an electrochemical generator for the supply of 213Bi that has the potential to overcome this limitation. Bismuth-213 spontaneously electrodeposits onto nickel foils in 0.1 M hydrochloric acid at 70 °C. Using this method, we were able to plate an average of 73 ± 4 % of the 213 Bi in solution and obtain a final 213 Bi recovery of 65 ± 8 % in 0.1 M citrate pH 4.5 via reverse electrolysis using titanium as the cathode. The recovered 213Bi had an average radiochemical purity of >99.8 % and was successfully used to radiolabel DOTATATE with an average radiochemical yield of 85.1 % (not optimized).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2025

Software Validation Work With The ZPPR-15 Data

The analysis activities for fast reactors involve using many different pieces of software that are relied upon for their predictive capabilities. For this software to be considered reliable, documented proof that the predictions of the software are accurate is required. In this manuscript, the validation work that covers some of the Argonne software used in fast reactor design activities is discussed and displayed. This validation work includes neutron and gamma flux distributions, reaction rate distributions, and reactivity worth. In an ideal world, a reactor development program would have access to a comprehensive set of experimental facilities to help inform the design aspects of the reactor itself. While thermal-hydraulics experiments, and to a limited degree mechanical experiments, can be carried out today for validation needs, neutronics related experimental facilities are rather impractical because of the lack of experimental facilities. Given the desired time table for construction of new reactors, the reconstitution or creation of new neutronic experimental facilities is untenable and thus those reactor development programs must rely upon any available experimental measurements that are qualitatively similar to the design. While a methodology has been proposed to assess the similarity between the past experimental measurements and the reactor itself, that aspect is beyond the scope of this manuscript. In this manuscript, the focus is entirely placed on the analysis results for a series of experiments carried out at the ZPPR facility in Idaho in the mid-1980s. In this regard, this manuscript only shows the validation of the stated neutronics software for specific loadings of the ZPPR reactor. Because of the fuel form, its proposed enrichment, and the material content of the reactor core, the ZPPR-15 experiments were identified as potential validation data for the reactor. The ZPPR-15 experiments were intended as mockups of a 330 MWe Integral Fast Reactor program which was a follow on program to the Clinch River Breeder Reactor. In the ZPPR-15 series of experiments, measurements of the neutron spectrum, control rod worth, sodium void worth, foil reaction rate distributions, Doppler worth of heated samples, gamma dose, and axial expansion worth were all carried out and published. In many cases, these reactivity coefficients are good candidates to validate the reactivity coefficient calculation scheme used by the analysis software and included in the safety analysis activities of fast reactor development projects today. This manuscript discusses the modeling methodology and accuracy of the calculated experimental results using the LANL software MCNP and the ANL software package ARC (Argonne Reactor Codes). As will be shown, for many of the experimental measurements, the two software packages are found to be good predictive analysis tools for those experiments. In other cases, problems with the analysis methodology or underlying cross section data are exposed which indicates where predictive analysis is not as reliable. Finally, in some of the measurements the conclusion is reached that the experimental measurement cannot be reproduced with the analysis software as it is simply too difficult.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

Software Validation Work With The ZPPR-15 Data

Aliberti, Gerardo↗

DOE OSTI2025

Nuclear excitation functions for medical isotope production: Targeted radionuclide therapy via nat IR$(d, x)$ 193m Pt

193m Pt is an Auger emitting radionuclide which may have therapeutic potential, particularly when labeled to the chemotherapeutic drug cisplatin. One challenge to broader explorations of its clinical potential is the need for production routes with high specific activity. As part of a larger campaign to address gaps in reaction data for emerging medical radionuclides, this work seeks to characterize the nat Ir(d,x) reactions as a potential production pathway for 193m Pt. A stacked target irradiation, consisting of natural iridium, iron, nickel, and copper foils, was performed using a 33 MeV deuteron beam at the Lawrence Berkeley National Laboratory 88-Inch Cyclotron. This measurement, along with previous experimental data, suggests an energy window between 11 to 18 MeV to maximize the production and radiopurity of 193m Pt. This experiment has yielded cross sections for 43 channels of deuteron-induced reactions from threshold to 30 MeV, including the first experimental results of nat Ir(d,x) 188m1+g,190m1+g Ir (cumulative), nat Ni(d,x) 56,57,58 m,58g Co (independent), nat Cu(d,x) 61 Co (cumulative) and nat Fe(d,x) 53 Fe, 48 V (cumulative). The results were compared with literature data, the TENDL-2023 database, and default theoretical calculations from the TALYS-2.04, CoH-3.6.0, EMPIRE 3.2.3, and ALICE-2020 reaction modeling codes. Here, this work presents another example of the lack of predictive capabilities for this set of modern nuclear-reaction modeling codes, and highlights the unsatisfactory modeling of experimental cross sections. Experimental data are important to improve the codes in general, and new experimental results can be used to improve the models. Finally, this measurement has revealed the need for an updated evaluation of the nat Cu(d,x) 63 Zn deuteron monitor reaction.

193mPt↗

DOE OSTI2024

Proton beam power limits for stationary water-cooled tungsten target with different cladding materials

The proton beam power limit for a solid-tungsten spallation target is largely determined by beam induced thermomechanical structural loads and decay heat power deposition, while its lifetime is limited by radiation damage and fatigue life of the target materials. In this paper, we studied the power limits of a stationary water-cooled solid tungsten target concept. Tantalum clad tungsten was considered as a reference case. Being a low activation material, zircaloy 2 cladding option was studied and its decay heat driven power limit was compared with the reference case. Zirconium alloys have proven operations records in spallation target and nuclear fission environments, supported by materials data obtained from post irradiation examinations. Recent study also demonstrated feasibility of diffusion bonding zirconium to tungsten using vanadium foil inter layer. Particle transport simulations code FLUKA was used to calculate energy deposition and decay heat power deposition in the target, based on the beam parameters technically feasible at the Second Target Station of the Spallation Neutron Source at Oak Ridge National Laboratory. The energy deposition data were used for flow, thermal, and structural analyses to determine the beam intensity limit on the target concept studied. The decay heat deposition data were used to calculate the transient temperature evolution in the tungsten volumes in a loss of coolant accident (LOCA) scenario to determine its beam power limit. For a 1.3 GeV proton beam, the power limit on a stationary target was 400 kW for a tantalum clad target model and 800 kW for a zircaloy 2 clad target model.

Lee, Yong Joong↗

DOE OSTI2025

Development of an atomic layer deposition system for deposition of alumina as a hydrogen permeation barrier

Tritium permeation into and through materials poses a critical challenge for the development of nuclear fusion reactors. Minimizing tritium permeation is essential for the safe and efficient use of available fuel supplies. In this work, we present the design, construction, and validation of custom atomic layer deposition (ALD) and deuterium permeation measurement systems aimed at developing thin-film hydrogen permeation barriers. Using the ALD system, we deposited conformal $\mathrm{Al}_{2}\mathrm{O}_{3}$ films on copper foil substrates and characterized their growth behavior, morphology, and composition. ALD growth rates of ∼1.1 Å/cycle were achieved for temperatures between 100 ∘ C and 210 ∘ C. Permeation measurements on bare and alumina coated copper foils revealed a significant reduction in deuterium flux with the addition of a ∼10 nm $\mathrm{Al}_{2}\mathrm{O}_{3}$ layer. While bare copper followed diffusion-limited transport consistent with Sievert’s law, the alumina-coated samples exhibited surface-limited, pore-mediated transport with linear pressure dependence. Arrhenius analysis showed distinct differences in activation energy for the two transport regimes, and permeation reduction factors exceeding an order of magnitude were observed. These results demonstrate the potential of ALD-grown $\mathrm{Al}_{2}\mathrm{O}_{3}$ films as effective hydrogen isotope barriers and provide a foundation for future studies on film optimization and integration into fusion-relevant components.

atomic layer deposition↗

DOE OSTI2026

Impact of nitrogen on lithium-induced corrosion and cracking of RAFM steel

The emergence of private fusion enterprises, combined with advancements in U.S. and EU ITER-TBM as well as EU DEMO liquid breeder blanket concepts, has renewed focus on the materials challenges. This study examines the compatibility of reduced activation ferritic/martensitic (RAFM) steel in liquid Lithium (Li) containing various Nitrogen (N) concentrations, tested at 600 and 700 °C for 500 and 1000 h under static conditions. Microstructural characterization revealed the corrosion-induced dissolution and precipitation in Li containing 0.082 and 0.17 wt.% N at 600 °C, while limited dissolution and precipitation were observed in N-gettered Li using zirconium (Zr) and titanium (Ti) foils at 600 °C. Tensile testing at room temperature after Li exposure revealed a decrease in strength and ductility of the RAFM steel as a function of N content in Li. Severe embrittlement was observed after 500 h in Li containing 0.17 wt.% N at 600 °C and after 500 h in hot-gettered Li at 700 °C. Chromium (Cr) depletion from the steel matrix was observed, leading to the formation of M6C-type carbides on the surface identified by electron microscopy (SEM), X-ray diffraction (XRD), in agreement with thermodynamic calculations. These findings highlight the critical materials challenges posed by liquid Li breeder blanket concepts, emphasizing the need for improved strategies to address Li-induced dissolution, precipitation, and steel embrittlement due to N contamination of the coolant.

Romedenne, Marie [ORNL] (ORCID:0000000317936561)↗