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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

AMMT Round Robin 316H HFIR Irradiation Test Matrix and Readiness for Insertion

The Round Robin 316H HFIR-2 irradiation campaign, conducted under the Advanced Materials and Manufacturing Technologies (AMMT) Program, is designed to evaluate the variability in irradiation response of laser powder bed fusion (LPBF) stainless steel (SS)-316H across multiple national laboratories, powder heats, and processing conditions, with comparison to wrought counterparts. A total of 11 materials are intended for irradiation in the High Flux Isotope Reactor (HFIR), targeting two irradiation temperatures (400°C and 600°C) and one level of irradiation damage (2 dpa). This irradiation campaign utilizes the standardized general tensile (GENTEN) capsule design to accommodate subsized tensile specimens of these materials. Thermal analyses were performed to ensure appropriate temperature control by design and uniformity within the capsules. All the capsules for this irradiation except one have been successfully assembled, welded, and tested and are ready for HFIR insertion. The last capsule requires re-build following a failed weld. All eight capsules are intended to be inserted in HFIR cycle 517, which is scheduled to start September 8, 2026. This report summarizes the design, material selection, and readiness for insertion of the capsules.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Using radiochemistry to diagnose fuel-ablator mix in inertial confinement fusion studies at the National Ignition Facility: measurement of Tc/Mo isotopic ratios for the Pushered Single Shell campaign

The development of radiochemical measurement techniques as a diagnostic for fusion experiments at the National Ignition Facility enables a new method for assessing fuel-ablator mix and the impact of this mix on capsule performance. Diagnosing capsule mix in internal confinement fusion studies is difficult due to the small spatial scales (10s of µm) and short-time frames (100s of ps) over which the mix typically evolves in these experiments. For the Pushered Single Shell campaign, radiochemical measurements on debris collected from fusion experiments can be used to determine isotopic ratios of activation products, particularly 96g Tc/ 99 Mo and 95g Tc/ 99 Mo, to provide vital information on nuclear reactions in the burning plasma that can inform simulations that seek to understand the degree of capsule-fuel mix and the impact on the capsule performance. These radiochemical measurements have been conducted regularly since November 2023 providing data on a range of capsule designs and neutron yields. Data from eight NIF experiments is presented, the measured 96g Tc/ 99 Mo and 95g Tc/ 99 Mo range from (0.5–5) × 10 –4 to (0.3–3) × 10 –4 , respectively. The development of radiochemical diagnostics aids in understanding and optimizing the design of fusion experiments, providing unique and valuable insights into capsule behavior and directly measuring fuel-ablator mix.

and nuclear chemistry↗
DOE OSTI2025

Development of an In Situ Fission Gas Release Instrument for Fuel Sample Irradiations in the High Flux Isotope Reactor

Experimental measurement of gaseous fission product release with respect to temperature and burnup is a critical aspect of understanding nuclear fuel performance, validating predictive models, and qualifying new fuels. To measure this phenomenon in real-time, Oak Ridge National Laboratory has developed an instrument for measuring in situ fission gas release from small-scale fuel samples irradiated in the High Flux Isotope Reactor (HFIR). The instrument uses a continuous flow of Heover the fuel samples to sweep gaseous fission products from a sealed capsule in the HFIR Be reflector to an instrument station adjacent to the reactor. The instrument station houses two high-purity germanium (HPGe) detectors that measure decay gamma rays from fission products passing through a room temperature dwell chamber placed over the detector crystal. The sealed capsules in the reactor are designed to modulate fuel sample temperatures between 700 and 1,100°C by changing the Ar/He gas mixture surrounding the capsules during irradiation. N-type thermocouples are incorporated into the capsule housing to record real-time fuel temperatures. The capsules are heated primarily by prompt gamma rays emitted from the HFIR core with minimal heat contributions from fission in the fuel samples to minimize temperature gradients in the specimens for separate-effects characterization of the material. This paper describes modeling of time-dependent nuclear heating and fission product formation in fuel samples, thermal characteristics of the in-core capsules, and expected gaseous fission product gamma spectra at the HPGe instrument station.

Mulligan, Padhraic L [ORNL] (ORCID:000000025826540↗
DOE OSTI2025

Quantification of Changes in Fill Tube Curvature During Target Assembly

National Ignition Facility (NIF) targets used for inertial confinement fusion (ICF) experiments are precision-engineered assemblies composed of more than 100 components carefully and precisely assembled with accuracies no larger than a few micrometers in most cases. When individual components deviate from these strict specifications during the assembly process, they can induce target failure. An integral component essential to ICF targets is the capsule-fill-tube assembly. This assembly involves bonding the capsule to a fused silica tube less than 5 μm in thickness, through which the tritium and deuterium (T 2 + D 2 ) fuel mixture is injected into the capsule prior to the NIF shot. The filling tube is bonded to a larger fused silica capillary, about 130 μm in outer diameter, which is coated with a polymeric layer on the order of a 10-μm thickness or less. Only tubes exhibiting a deflection equivalent to less than 1% of their total length from a perfectly straight line are suitable for assembly. Slight curvatures on the order of 1 mm over 10 cm can induce unwanted stresses, potentially causing capsule misalignment and resulting in clogging or leaks during the filling process. Despite manually qualitatively sorting each tube for straightness prior assembly, it has been found that the treatments the tubes undergo once attached to the capsule can alter their curvature. Filling tubes that initially satisfy straightness tolerances can undergo geometric deformation, resulting in curvature deviations exceeding 1% of their total length and thus failing to meet target assembly requirements. Here, in this study, we propose a metric to assess the degree of curvature of the filling tubes and to gauge their changes in curvature caused by the standard thermal processing methods employed in target assembly. In addition, we suggest alternatives to mitigate tube curvature, ensuring they conform to specifications following assembly in the final targets.

Materials science↗
DOE OSTI2025

A methodology for corrosion testing ODS steels in liquid tin under reactor conditions in HFIR

Liquid metal–based divertor concepts have promising attributes for application in fusion reactors. Liquid tin is being considered as one of those liquid metals due to its excellent thermophysical properties. However, tin is extremely corrosive to steels at elevated temperatures, and the coupled effects of corrosion with neutron irradiation for application in fusion reactors have not been quantified. Researchers at Oak Ridge National Laboratory (ORNL) have designed and irradiated a series of capsules in the High Flux Isotope Reactor (HFIR) to help address gaps in the literature with respect to coupled corrosion and neutron irradiation effects. The capsules are filled with solid tin designed to melt due to the gamma heating in HFIR and allow interaction with the specimens. The capsules contain a SiC thermometer for post-irradiation temperature verification. Five capsules were irradiated in HFIR for 10.5 days, accumulating 0.93 dpa in the FeCrAl specimens. Post-irradiation temperature verification was performed on the SiC thermometers, and temperatures were observed to be higher than what was expected from thermal models. An explanation for the model underprediction is given herein. In conclusion, the capsules described within this article are the first to irradiate liquid metal within the HFIR flux trap and demonstrated the safety basis for continued liquid coolant studies in HFIR.

FeCrAl↗
DOE OSTI2024

Measurement of early time outer laser beam reflection inside a cylindrical hohlraum at the National Ignition Facility

In a National Ignition Facility (NIF) cylindrical hohlraum, any specular reflection (“glint”) off the wall from the outer cone (incident angle of 50° and 44° relative to the surface of the wall) laser beam will irradiate the capsule poles. If the glint power is sufficiently large during the picket (early time) of the laser pulse, it may seed high-mode perturbations on the capsule surface that can grow during the implosion. To quantify the glint power on the capsule during the picket by the outer beams, we performed dedicated experiments on NIF using a flat witness foil as a surrogate for the capsule in a half-hohlraum target. We found that the measured glint power is approximately comparable to simulations using a nominal electron conduction flux limiter of f = 0.15, and over an order of magnitude lower than that predicted using f = 0.03 in the wall. Based on our current understanding, we conclude that the glint from the outer beams plays an insignificant role in capsule drive asymmetry.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

NSUF BOILER Pre-Irradiation Characterization and High Flux Isotope Reactor Experiment Design

Alumina-forming austenitic (AFA) stainless steels have emerged as a candidate alloy because of their high-temperature strength, formability, cost, and compatibility with primary coolants for lead-cooled fast reactors (LFRs). This class of steels has exceptional high-temperature oxidation performance; however, a high concentration of Ni is required to stabilize the austenite phase and to provide sufficient creep strength. AFA stainless steels are also susceptible to liquid metal embrittlement (LME). Additionally, under neutron irradiation, Ni will enrich at grain boundaries due to radiation-induced segregation (RIS). Nickel RIS can increase the LME under these coupled effects. Oak Ridge National Laboratory (ORNL) and the NSUF program have leveraged its High Flux Isotope Reactor (HFIR) and experience with complex irradiation experiments to design experiment capsules that test the aforementioned coupled effects. These capsules are designed for insertion in the central flux trap, the highest flux region, of HFIR. The experiment capsules will be filled with Pb, designed to passively melt from the gamma heating in HFIR. The specimens were fabricated into miniature tensile specimens from two different alloys, GA05-25Ni and GA05-20Ni, varying Ni concentrations. The experiment capsules are designed to achieve target temperatures of 400 °C and 650 °C with accumulated dosage of 3 dpa. This report documents the specimen alloy characterization, experimental design, and expected performance of the capsules.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Design Report on the MiniFuel Instrumented Test Apparatus for Understanding Radiation Effects

Most nuclear fuels irradiations at Oak Ridge National Laboratory (ORNL) over the past decade have been conducted using MiniFuel—a static capsule design employing subscale fuel specimens to collect separate-effects irradiated fuel performance data. Irradiation conditions for MiniFuel experiments are predicted pre-test using reactor physics, and thermal models are verified post-irradiation via SiC dilatometry and various spectrometry methods. Relevant fuel performance parameters are also observed post-irradiation in a hot cell, thereby providing a single data point for each parameter representing the cumulative effects of the irradiation conditions. Substantially more data can be harvested from a single test and within a shorter duration by instrumenting irradiation vehicles and measuring desired quantities in situ. This report presents the design and analysis of the MiniFuel INstrumented Irradiation Test Apparatus for Understanding Radiation Effects (MINITAURE)—an instrumented test rig based on the separate-effects MiniFuel concept that aims to capture fission gas release (FGR) and thermal conductivity degradation of fuel specimens during irradiation in the High Flux Isotope Reactor (HFIR). MINITAURE will be integrated with the Materials Irradiation Facility (MIF) located in the HFIR building outside the reactor containment. The MIF will act as the instrumentation and control center for the experiment, enabling real-time feedback from in situ sensors and control of irradiation temperatures via a gas delivery system. Two unique capsule designs were developed to capture each phenomenon: the thermal conductivity capsule, which uses a thermopile method to estimate fuel specimen thermal conductivity, and the fission gas release capsule, which will have continuous flowing gas communication to high-purity germanium detectors that are housed in the MIF for monitoring FGR. This report details the reactor physics and heat transfer modeling activities that were used to inform the experiment design and predict capsule performance. It also describes out-of-pile activities conducted to stand up this new capability and verify the measurement techniques. Modeling efforts to date have demonstrated the feasibility of the in situ measurement techniques and supported the development of the MINITAURE assembly configuration. Out-of-pile testing of the thermal conductivity measurement shows promise in capturing relative changes in thermal conductivity. However, significant errors exist in the measured absolute value, posing a need for further refinement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of a Target Rod Rabbit Holder Facility in the Large Removable Beryllium Position

A new facility, the removable beryllium-target rod rabbit holder facility (RB*-TRRH), was designed and fielded in the large removable beryllium (RB*) position in the High Flux Isotope Reactor (HFIR). This facility supports three shortened target rod rabbit holders (sTRRHs), which allow for the irradiation of six axially stacked rabbit capsules per sTRRH. This will allow for the simultaneous irradiation of 18 rabbit capsules in the RB* position and will provide an alternative location for low-cost rabbit capsules to be irradiated in HFIR. The first capsules inserted into this facility are 18 perforated capsules that contain representative steel specimens of interest to the calandria of Canada deuterium uranium (CANDU) reactors as part of a lifetime extension program with Canadian Nuclear Laboratories (CNL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

AGR-5/6/7 Irradiation Experiment Fission Product Mass Balance

This report presents the fission product mass balance for the AGR-5/6/7 TRISO fuel irradiation experiment. The fission product inventories deposited on capsule components outside of the fuel (e.g., stainless-steel shells, graphite holders, Grafoil disks, and associated hardware) were quantified as part of the post-irradiation examination (PIE) to assess the performance of this fuel. Comparisons were made between these inventories and depletion calculations, non-destructive measurements of fuel fission product inventories in the intact fuel compacts, prior AGR experiments, and results from among each of the five distinct AGR-5/6/7 capsules. The data served as estimates of the condensable fission products released from the fuel during irradiation. Excluding Capsule 1 (which experienced accidental damage during irradiation) and Capsule 3 (which was tested at very high irradiation temperatures of >1300°C), the results indicate that the AGR-5/6/7 fuel performed comparably to fuel from earlier AGR experiments. The mass balance results were also used to estimate the number of particles with in-pile SiC failures. Subject to the assumptions made in these estimates, and excluding Capsules 1 and 3, the in-pile SiC failure rates for AGR-5/6/7 are comparable to those observed in AGR-2.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Exploring scenarios for enhanced fuel compression and performance on the National Ignition Facility with machine-learning-aided design techniques

Recent fusion experiments on the National Ignition Facility (NIF) have achieved ignition, producing multi-MJ fusion yields for input laser energies of roughly 2 MJ [Abu-Shawareb et al., Phys. Rev. Lett. 132, 065102 (2024)]. Building on the success of the target designs that have achieved ignition, we explore new implosion scenarios predicted to generate significantly more compression of the dense DT ice layer and correspondingly higher yields while preserving many of the key physics characteristics of present-day ignition designs. Our main result is a novel 3-shock implosion scheme that effectively minimizes the shock-induced entropy in the dense, accelerating DT shell and maximizes the resulting fuel compression subject to a fixed leading shock strength consistent with present-day ignition experiments, which is necessary to melt the crystalline high-density carbon ablator. Compared to the first NIF experiment to fulfill Lawson's ignition criterion, shot N210808 [Abu-Shawareb et al., Phys. Rev. Lett. 129, 075001 (2022)], our design exhibits a 40% increase in simulated peak areal density (ρR) and a 5× increase in 1D fusion yield using a 4% lighter ablator and identical DT payloads. We also present a complete integrated 2D hohlraum design and laser pulse specifications capable of generating the desired 3-shock drive and maintaining control of the low-mode capsule implosion symmetry, where the increase in simulated 2D yield relative to N210808 is > 10×. This new implosion regime was discovered with help from a machine-learning-enabled capsule design optimization framework. We outline the workflow this automated tool uses to identify improved design candidates by running several rounds of capsule simulations, constructing a surrogate model mapping input variations to key physics output quantities, and querying the resulting statistical model to propose adjustments to the x-ray drive and capsule to reach a set of physics objectives prescribed by the designer.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Direct Experimental Proof of the Principal Role of Reduced High-Mode Hydrodynamic Mix in Recent Ignition Success on NIF

This Letter details the role of reduced hydrodynamic mix in achieving the first fusion plasma to satisfy the Lawson criterion for ignition. Using novel 3D measurements of the temperature and density of the burning deuterium-tritium plasma across a series of experiments, we demonstrate that hydrodynamic mixing from small-scale capsule defects was a key degradation mechanism inhibiting ignition. A series of five layered deuterium-tritium experiments using 1.9 MJ of laser energy demonstrates a 2 × reduction of the thermonuclear yield when a lower quality capsule, typical of preignition experiments, is used. Radiation hydrodynamic calculations consistent with the observed mix profile show that the mixed fuel’s reduced compression and increased radiative loss explain the yield reduction. Furthermore, a hydroscaling argument suggests that a ∼ 3 MJ laser would be needed to reach ignition with such a capsule, well beyond the current capabilities of high energy laser systems. As such, this Letter gives critical insight into the physics of mix in self-heated fusion plasmas and establishes a foundation to evaluate capsule quality needs of future inertial fusion designs that aim at ignition and high gain.

High-energy-density plasmas↗
DOE OSTI2025

Role of glucuronoxylomannan and steryl glucosides in protecting against cryptococcosis

The development of vaccines for fungal diseases, including cryptococcosis, is an emergent line of research and development. In previous studies, we showed that a Cryptococcus mutant lacking the SGL1 gene (∆sgl1) accumulates certain glycolipids called steryl glucosides (SGs) on the fungal capsule, promoting an effective immunostimulation that totally protects the host from a secondary cryptococcal infection. However, this protection is lost when the cryptococcal capsule is absent in the ∆sgl1 background. The cryptococcal capsule is mainly composed of glucuronoxylomannan (GXM), a polysaccharide microfiber consisting of glucuronic acid, xylose, and mannose linked by glycosidic bonds forming specific triads. In this study, we engineered cells to lack each of the GXM components and tested the effect of these deletions on protection under the condition of SG accumulation. We found that glucuronic acid and xylose are required for protection, and their absence abrogates the production of IFNγ and IL-17A by γδ T cells, which are necessary stimulants for the protective phenotype of the ∆sgl1. We analyzed the structure of the GXM microfibers and found that although the deletion of SGL1 only slightly affects the size and distribution of these microfibers, it significantly changes the ratio of mannose to other components. In conclusion, this study identifies the structural modifications that the deletion of SGL1 and the consequent accumulation of SGs impart to the GXM structure of C. neoformans. This provides significant insights into the protective mechanisms mediated by SG accumulation on the capsule, with important implications for the future development of an efficacious cryptococcal vaccine.

Cryptococcus neoformans↗
DOE OSTI2026

Calculations Document

Figure 1 shows a simplified model schematic of the Tetracage target design. In this figure the green lines represent the Tetracage yarns, which consist of two parallel yarns on the top and two parallel yarns on the bottom with the pairs in a cross hatched pattern. For simplicity, as we are only considering vertical forces, we will perform calculations without consideration for the two pairs of yarns being perpendicular to each other. Also worth noting is the model shows the yarns extending directly to the capsule then closely hugging it – in reality, the yarns behave as you would expect where they hug the capsule only as much as the geometry of their deflection allows, similarly to if you held a tennis ball with a pair of parallel strings anchored at opposing ends. The dashed black lines show the path the yarns take straight across from slot glue point to slot glue point before the target is closed and the capsule deforms them. The variable r corresponds to the capsule outer radius. The variable L corresponds to the hohlraum inner radius. The variable a corresponds to the yarn offset from the center of the diagnostic band. The variable d corresponds to the wire deflection vertically.

42 ENGINEERING↗
DOE OSTI2026

HDG-1 Experiment Irradiation Monitoring Data Qualification Final Report

SUMMARY The U.S. Department of Energy (DOE) Advanced Reactor Technologies (ART) Graphite Research and Development (GRD) Program is conducting a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. This report documents the qualification of irradiation monitoring data for the fifth experiment, High Dose Graphite-1 (HDG-1). Qualified monitoring data are required by the ART program to support the design and licensing of the first high-temperature reactor (HTR) nuclear plant. Data are classified as Qualified if they meet the usage requirements described in the experiment planning and quality assurance (QA) documents, Failed if they do not meet those requirements and provide no usable information, or Trend if they do not fully meet all requirements but still provide useful information subject to an assessment of how any deficiencies may affect a particular use of the data. HDG-1 irradiation began with Advanced Test Reactor (ATR) Cycle 168B on August 24, 2020, and concluded after Cycle 173C on January 27, 2025. The HDG-1 capsule was removed from the reactor core twice—during core internal change (CIC) Cycle 170A and powered axial locator mechanism (PALM) Cycle 172A—to prevent overheating of the graphite specimens during high-power PALM cycles. The capsule was therefore irradiated during a total of seven normal ATR cycles: 168B, 169A, 171A, 171B, 173A, 173B, and 173C. Irradiation monitoring data evaluated in this report include thermocouple (TC) temperature, gas flow rate, gas moisture, gas pressure, specimen load, and graphite stack displacement. Temperature. A total of 14,508,065 TC temperature records were captured. Of these, 13,901,785 (95.8%) are Qualified and 606,280 (4.2%) are Failed. The principal source of failed temperature data was the instrument failure of TC-9 (Zone 2) on June 24, 2024, and TC-10 (Zone 1) on July 5, 2024, near the end of Cycle 173A, which resulted in 595,554 Failed readings. An additional 379 missing values and 10,347 slightly negative values from TC-13 during ATR outages are also Failed. Neither TC-9 nor TC-10 was used as a temperature-control TC, and their failures did not compromise capsule condition monitoring. Correlation analysis of all 13 TCs found no evidence of virtual junction formation. Control chart analysis revealed clear downward drift of approximately 80°C for TC-6 (Zone 3) relative to other stable TCs, and possible downward drift of approximately 60°C for TC-13 relative to the Zone 5 control TC (TC-1), though TC-13 remained consistent with the Zone 2 control TC (TC-12). Gas flow. A total of 20,088,090 gas flow rate records were captured. Of these, 19,941,463 (99.3%) are Qualified and 146,627 (0.7%) are Failed due to missing values. All argon, helium, and total gas flow data were within expected ranges throughout the irradiation. Gas moisture. A total of 1,116,005 outlet gas moisture values were captured. Of these, 1,101,421 (98.7%) are Qualified and 14,584 (1.3%) are Failed, comprising 14,556 out-of-range values and 28 missing values. The out-of-range moisture values exceeded 22,000 ppmv for approximately 1 week at the beginning of Cycle 173A, when accumulated moisture evaporated after the capsule was retrieved from water storage during PALM Cycle 172A and reinserted into the east flux trap. Moisture levels returned to below 25 ppmv for the remaining three cycles, and the transient high-moisture event did not affect the integrity of specimen irradiation. Gas pressure. A total of 7,812,035 gas pressure values were captured. Of these, 6,642,048 (85.0%) are Qualified and 1,169,987 (15.0%) outlet pressure values are Failed, comprising 718,537 zero outlet pressure values due to sensor failure from Cycle 168B through Cycle 171B, 54,550 missing values, and 396,900 too-low outlet pressure values, ranging from 1.1 to 1.6 psia after sensor replacement during Cycle 173A. Load. A total of 6,696,030 load values were captured. Of these, 6,694,580 (99.98%) are Qualified and 1,450 (0.02%) are Failed due to missing values. Applied loads to the six specimen stacks were stable throughout the irradiation. Stack displacement. A total of 6,696,030 displacement values were captured. Of these, 5,713,297 (85.32%) are Qualified and 3,781 (0.06%) are Failed due to missing values. Stack displacement increased consistently throughout the irradiation, reaching approximately 3.08 in. for Channels 5 and 6 by the end of irradiation. 978,952 (14.62%) substantially elevated displacements observed for Channel 6 beginning in Cycle 171A and for Channel 5 beginning in Cycle 173A are assigned Trend status. Raising pressure. A total of 1,115,999 raising pressure values were captured. Of these, 1,115,430 (99.95%) are Qualified and 569 (0.05%) are Failed due to missing values. Ram pressure. A total of 6,696,030 ram pressure values were captured. Of these, 6,692,249 (99.95%) are Qualified and 3,484 (0.05%) are Failed due to missing values. Stack raising was perf

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Bayesian batch optimization for molybdenum versus tungsten inertial confinement fusion double shell target design

Access to reliable, clean energy sources is a major concern for national security. Much research is focused on the “grand challenge” of producing energy via controlled fusion reactions in a laboratory setting. For fusion experiments, specifically inertial confinement fusion (ICF), to produce sufficient energy, the fusion reactions in the ICF fuel need to become self-sustaining and burn deuterium-tritium (DT) fuel efficiently. The recent record-breaking NIF ignition shot was able to achieve this goal as well as produce more energy than used to drive the experiment. This achievement brings self-sustaining fusion-based power systems closer than ever before, capable of providing humans with access to secure, renewable energy. In order to further progress toward the actualization of such power systems, more ICF experiments need to be conducted at large laser facilities such as the United States's National Ignition Facility (NIF) or France's Laser Mega-Joule. The high cost per shot and limited number of shots that are possible per year make it prohibitive to perform large numbers of experiments. As such, experimental design relies heavily on complex predictive physics simulations for high-fidelity “preshot” analysis. These multidimensional, multi-physics, high-fidelity simulations have to account for a variety of input parameters as well as modeling the extreme conditions (pressures and densities) present at ignition. Such simulations (especially in 3D) can become computationally prohibitive to turn around for each ICF experiment. In this work, we explore using Bayesian optimization with Gaussian processes (GPs) to find optimal designs for ICF double shell targets, while keeping computational costs to manageable levels. These double shell targets have an inner shell that grades from beryllium on the outer surface to the higher Z material molybdenum, as opposed to the nominally used tungsten, on the inside in order to trade off between the high performance associated with high density inner shells and capsule stability. We describe our results for “capsule-only” xRAGE simulations to study the physics between different capsule designs, inner shell materials, and potential for future experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

FLiBe compatibility of fusion-relevant materials

There is current interest in a 2LiF-BeF2 (FLiBe) molten salt blanket for fusion applications, however, very little compatibility data are available for fusion relevant materials. Initial static capsule testing results for 500 h at 550°-750 °C are presented for V-4Cr-4Ti, Fe-8Cr-2W (alloy F82H) compared to type 316H stainless steel and monolithic SiC as a surrogate for SiC/SiC composites. In addition, sintered Be12Ti specimens with varying additions of Be (0–20%) and a Be2C specimen were exposed with F82H coupons at 650 °C in individual Mo capsules. The V-4Cr-4Ti and SiC specimens exposed at 650 °C had small mass changes. Steel specimens showed increasing mass losses with increasing exposure temperature. However, when Be12Ti-Be specimens were included in a capsule with an F82H coupon, increasing F82H mass gains were observed with increasing Be content due to the formation of a Fe-Be surface layer. No layer and no mass gain were observed for the F82H specimen tested with Be2C.

Pint, Bruce [Oak Ridge National Laboratory (ORNL) ↗
DOE OSTI2025

Stabilization of Preternatural Barium Oxidation States as an Unexpected Byproduct of β-Decay: Discovery of a New Halide Semiconductor Alloy

137 Cs has a wide range of roles in the nuclear industry. The solid material, safely encapsulated in CsCl as 137 CsCl, is stored as fission product waste from nuclear power production and legacy waste from nuclear weapons production; it has also served as a radiation source in food and sewage irradiators as well as medical devices. However, because of the solubility of the chloride salt and the relatively high specific activity of 137 Cs, damaged or broken capsules can lead to severe radiological accidents. Safe capsule design and material recycling are complicated by the unclear structural evolution during β-decay, which remains ambiguous due to the differing oxidation states of Cs (1+) and Ba (2+). Here, in this study, we use first-principles calculations to investigate the evolving structure–property relationships of Cs 1–x Ba x Cl during β-decay. Despite the well-established 2+ formal oxidation state of alkali-earth metals, we find that Ba 1+ can be stabilized in the form of a mixed-valence alloy at low concentrations. Specifically, we identify three regimes for the β-decay of 137 Cs into CsCl: Ba-doped CsCl (Ba ≤ 14%), wherein Ba has the expected 2+ oxidation state; Cs–Ba–Cl alloys, where Ba has a mix of the usual Ba 2+ and highly unusual Ba 1+ oxidation state in the form of a quasi-disordered mixed-valence alloy (Ba = 25%); and phase separation into a CsCl + BaCl 2 + Ba (m) mechanical mixture, where Ba reverts to its expected 2+ oxidation state (Ba > 25%). Surprisingly, the Cs 0.75 Ba 0.25 Cl mixed-valence alloy is a narrow indirect band gap semiconductor (1.05 eV) despite the insulating nature of both CsCl and BaCl 2 . It also exhibits strongly excitonic polarized optical properties, has glass-like ultralow thermal conductivity (directional average of 0.21 W/mK at 300 K), and shows greater resistance to deformation under both tensile and volumetric strengths compared with the original CsCl structure (e.g., shear and Young’s modulus of 9.04 and 31.62 GPa, respectively). These findings imply that transmutation of 137 Cs leads to highly unusual chemical bonding that stabilizes Ba 1+ in local regions of the quasi-disordered Cs 0.25 Ba 0.75 Cl, resulting in anomalous physical properties. Moreover, this discovery provides valuable insight for safe nuclear waste capsule design, which can aid in preventing environmental or human exposure to radioactive materials.

Fuhr, Addis S. [Oak Ridge National Laboratory (ORN↗