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High-throughput ion irradiation of additively manufactured compositionally complex alloys

Several advanced nuclear reactor designs promise efficiency and safety improvements over the current reactor fleet but are limited by the current set of ASME code-qualified materials. Novel alloys including high-entropy alloys (HEAs), and more broadly compositionally complex alloys (CCAs), have shown promising irradiation-tolerance. However, the vast range of alloy compositions adds to an already time-consuming alloy development process. In this study, to accelerate the development of novel alloys for nuclear applications, a high-throughput (HTP) methodology has been employed. Additive manufacturing has been used to produce a compositional array of unary, binary, ternary, and quaternary alloys, including several CCAs, which span the Cr-Fe-Mn-Ni composition space. Additionally, the compositional array was homogenized at 1000°C for 24 hours and each sample was irradiated using 4-MeV Ni2+ ions at room temperature to a peak damage of 50 dpa, as estimated using SRIM, at the University of Wisconsin Ion Beam Laboratory. A custom XY stage was built to accommodate the large compositional array and half of each sample was masked during irradiation enabling both the irradiated and unirradiated properties of each alloy to be characterized side-by-side. Each alloy was characterized using X-ray fluorescence (XRF), X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and nanoindentation. CALPHAD simulations spanning the entire Cr-Fe-Mn-Ni composition space at 1000°C were performed to compare predicted equilibrium phases with phases identified experimentally from the unirradiated regions of each alloy. Nanoindentation measurements indicate radiation-induced hardening ranging from ~1-1.5 GPa in each of Cr-Fe-Mn-Ni CCAs, which is relatively insensitive to modest changes in alloy composition and comparable to hardening observed in neutron irradiated Cr-Fe-Mn-Ni CCAs in the literature. Overall, a substantial time savings was realized by employing HTP synthesis, irradiation, and characterization in this study compared to conventional techniques, the implications of which are discussed.

36 MATERIALS SCIENCE↗

Understanding the Space Weathering of Mercury Through Laboratory Experiments

Introduction: Airless surfaces across the solar system are continually modified by energetic particles from solar wind and micrometeoroid bombardment [1,2]. This process is known as space weathering, and it alters the chemical, microstructural, and optical properties of surface regoliths on airless bodies, including Mercury. On the Moon and S-type asteroids, the reflectance spectral signatures of space weathering include reddening (increasing reflectance with increasing wavelength), darkening (lowering of reflectance), and the attenuation of characteristic absorption bands [2]. Such spectral changes are driven by the production of Fe-bearing nanoparticles(npFe) through both solar wind irradiation and micrometeoroid bombardment. While our understanding of space weathering for the Moon and near-Earth S-types asteroids is advanced, insight into how these processes operate on other planetary bodies is limited. In particular, Mercury experiences a uniquely intense space weathering environment than planetary counterparts at 1 AU, including a moreintense solar wind flux and higher velocity micrometeoroid impacts [4]. Additionally, Mercury has a surface composition unique in the inner solar system, including regions of the surface with very low albedo known as the low reflectance material (LRM), which is enriched in carbon, likely graphite, up to 4wt.% [5].In addition, the concentration of Fe across Mercury’s surface islow (<2 wt.%) compared to the Moonor S-type asteroids asteroids[6]. Our understanding of the effects of space weathering on C-rich and Fe-poor phases is limited. Since Fe plays a critical role inthe development of space weathering characteristicson other airless surfaces(e.g., npFe), its limited availability may significantly affect the development of space weathering features in Mercury surface materials. We can simulate space weathering processes in the laboratory to explore their effects on the microstructural, chemical, and spectral characteristics of Mercury surface materials[7]. Here we used pulsed laser irradiation to simulate the short duration, high-temperature events associated with micrometeoroid impacts. We performed coordinated analyses including reflectance spectroscopy and electron microscopy to investigate the spectral, chemical, and microstructural changes in these mercurian analog samples. Methods: For these experiments, we usedforsteritic olivine with varying FeOcontents, a mineral phase proposed to be abundant on the surface of Mercury. We mixed each sample with graphite to simulate LRM regions of the surface. Wesynthesized the olivinesamplesat 1-bar at NASA’s Johnson Space Centerand prepared pressed powder pellets for laser irradiation [8].We prepared three samples, each with a base layer of olivineto maintain structural integrity and topped witha surface layer containing the graphite-olivine mixture: 1) Sample SC-001 San Carlos olivine(Fo90.91),2)Sample F-S-002 with0.05 wt.% FeO olivine, and 3) F-T-004 with 0.53 wt.% FeO olivine. Each sample was mixed with 5 wt.% powdered graphite and had grain sizes ranging from 45to 125μm. We irradiated each sample usinga pulsed Nd-YAG laser, (l=1064 nm, ~6 ns pulse duration, energy of 48 mJ/pulse) while undervacuumat Northern Arizona University. The laser was rastered1x and then 5x over the surfaceof each sample to simulate progressive space weathering. We collected in situreflectance spectra from the samples after each laser pulse witha Nicolet IS50 Fourier-Transform Infrared spectrometer (lfrom 0.65-2.5 μm). We used an FEI Nova NanoSEM200scanning electron microscope (SEM) and a Hitachi TM4000 Plus benchtop SEM at Purdue University to image the surface morphology and topography of the samples. We extracted thin sections for analysis in the transmission electron microscope(TEM)using the FEI Helios NanoLab 660 focused ion beam (FIB) SEM at the University of Arizona.We performed analysis of the microstructural and chemical characteristics of the samples using the 200 keV JEOL 2500 scanning TEM at Johnson Space Center. Reflectance Spectroscopy Results:Reflectance spectra for each sample are shown in Fig. 1.SC-001:The spectrum of the unirradiated sample exhibits a weak 1.0 μm absorption feature, associated with Fe2+in the olivine,and low overall reflectance (Fig. 1a). Thereflectance and the depth of the absorption band increases after 1x laser raster but are at their lowest after 5x laser rasters.F-T-004:The unirradiated sample has a blue-sloped spectrum with low reflectance without identifiable absorption features (Fig. 1b). With progressive laser irradiation, the sample reflectance increases and becomes strongly red-sloped.F-S-002:The unirradiated sample exhibits a dark, blue-slopedspectrum. The brightness of the sample increases significantly from <0.2average reflectance over >0.8 reflectance in the most irradiated sampleand thespectral slope also becomesslightly reddened (Fig. 1c). Microstructural and Chemical Analysis: Two primary alteration textures were observed in the samples exposed to simulated space weathering: 1) fluffy C-rich,and 2) vesiculated melt. The fluffy C-rich texture is composed oflow-densitydeposits distributed across the surface of the sample(Fig. 2A). Analysis of a FIB section extracted from a low-density C-rich region in sample SC-001 reveals multiple globule-type deposits, discrete from stacked graphite, likely produced via melting from the laser irradiation [9].The vesiculated melt textureis smooth and uniformly distributed across isolated regions of the sample surface. The vesicles measure up to 100s of nmin diameter. Analysis of a FIB section from this texture was extracted from sample F-T-004 reveals a layer of amorphous melt material, close to 100 nm thick and uniform across the FIB section (Fig. 2B). Isolated regions of this melt layer contain small nanoparticles, <5 nm in diameter. Chemical analysis through energy dispersive X-ray spectroscopy reveals the composition of this layer is enriched in Si and depleted in Mg and O compared to the underlying sample. Implications for Space Weathering on Mercury: Previous experiments simulating space weathering of Mercury have showndarkening and reddening of spectra[7,10].However, our use of low-Fe materials and graphite to create a sample set more analogous to the mercurian surface. Our results indicate that sample composition plays a significant and important role in the space weathering of Mercury. In particular, our spectral data demonstrates a strong correlation between spectral slope, Fe content, and simulated space weathering. While the variation in FeO content between samples F-S-002 and F-T-004 is <0.6 wt.%, the spectra deviate from flat to strongly red-sloped(F-T-004). This reddeningmay be linked to the presence of very small nanoparticles observed in the melt textures extracted from sample F-T-004. For the SC-001 sample, the fluffy C-rich textures may be developed by the amalgamation of small graphite particles into these unique morphologies. Such observations indicate that space weathering on Mercury may result in both familiar and new microstructural and chemical characteristics. References: [1]Hapke B. (2001) J. Geophys. Res.-Planet.,106,10039–10073. [2]Pieters C.M. and Noble S.K. (2016) J. Geophys. Res-Planet., 121, 1865–1884. [3] Lucey P.G., and Riner, M.A. (2011) Icarus,212, 451-462.[4]CintalaM.J.(1992)J. Geophys. Res.-Planet.,97,947–973.[5]Klima R.L.et al.(2018)Geophys.Res.Letters, 45, 2945–2953. [6]Nittler L.R., et al. (2011) Science 333, 1847-1850.[7]Sasaki S. and Kurahashi E. (2004) Space weathering on Mercury, Adv.Space Res., 33, 2152-2155.[8] Vander KaadenK.E., et al. (2018) LPSCXLIX, Abstract 1230. [9] McGlaun M.L. et al. (2019) LPSCL, Abstract 2019. [10] TrangD.et al. (2018)LPSCXLIX,Abstract2083

M S Thompson↗

Single-phase, natural circulation annular flow measurements for cartridge loop irradiation experiments

The nuclear industry is increasingly considering cartridge-style experiments for irradiation testing of advanced reactor fuels and materials under flowing conditions. Cartridge loops do not require the extensive support infrastructure that are necessary for external flow loops and minimize the possibility of coolant solidification over the long distance from the reactor to the external facilities. However, there is a general lack of quality flow data for internally heated fluids in an annular configuration representative of a cartridge-type irradiation experiment, particularly one with natural circulation. To address this data need, a series of experiments was conducted to measure the natural circulation flow rates of pressurized water in a sealed, internally heated vessel with annular flow conditions that represent a molten salt or sodium cartridge loop. Temperatures and flow rates were measured under steady-state and transient conditions. Here, this paper describes the facility, methods, and results of the experiments, including the determination of nondimensional parameters. A simple 1D model of the natural convection flow rates agrees well with the experimental results. Applying this model to simulate a liquid salt cartridge experiment predicts that natural circulation flow might be able to provide liquid salt Reynolds numbers similar to those of some molten salt reactor concepts at relevant power densities.

42 ENGINEERING↗

Radiation Improves Materials Bonding

Bonds with fluorocarbons made with surface preparation. Irradiating rubber-based adhesives produce strong adhesive bond with fluorocarbons without extensive surface preparation.

Bouquet, F. L.↗

Integrating HPC simulations and physical experiments to characterize the effects of gamma radiation on seismic protective devices

Seismic protective systems, composed of seismic isolators and dampers, can substantially reduce the effects of earthquake shaking on nuclear power plants and components therein. To enable the use of these devices to protect equipment inside a plant and close to a source of radiation, the U.S. Department of Energy (DOE) funded a project at the Idaho National Laboratory (INL) and the University at Buffalo to characterize the effects of absorbed gamma dose on their mechanical properties. An early task in the project was to determine the exposure time required in the INL Foss Therapy Services (FTS) 60 Co gamma irradiator to achieve a target absorbed dose in the materials used to construct isolators and dampers, including fluids, polymers, composites, and metals. This task required the novel integration of high-performance computing (HPC), Monte Carlo N-Particle (MCNP) simulations, and irradiation experiments using Fricke dosimetry. An MCNP model of the FTS irradiator at INL was developed and validated using Fricke dosimetry. Simulations of three experiments in the irradiator, two with Fricke vials only and one with Fricke vials and a large-size isolator, predicted the Fricke-measured absorbed dose rate to within 15% in all three cases, providing high confidence in the calculation of the gamma dose absorbed in the materials comprising the seismic protective devices. The simulations demonstrated that the effects of photon scattering on absorbed dose rate in the FTS irradiator are negligible for test articles installed close to the cobalt sources and near the rear of the irradiator. The validated MCNP model of the FTS irradiator is being used to support ongoing DOE-funded experiments on seismic protective devices and could be applied to future, non-seismic-related experiments. In conclusion, the novel validation process successfully deployed for the FTS irradiator at INL could be applied to other irradiators, requiring new MCNP models and simulations, and irradiation experiments using dosimeters.

42 - ENGINEERING↗

Actinide targets for the synthesis of superheavy nuclei

The use of heavy actinide targets, including 243 Am, 240,242,244 Pu, 245,248 Cm, 249 Bk, and 249 Cf, irradiated by intense heavy ion beams of 48 Ca has resulted in a significant expansion of the periodic table since 2000, including the discovery of five new heaviest elements and more than 50 new isotopes. These actinide materials can only be produced by intense neutron irradiation in very high flux reactors followed by chemical processing and purification in specialized hot cell facilities available in only a few locations worldwide. This paper reviews the reactor production of heavy actinides, the recovery and chemical separation of actinide materials, and the preparation of actinide targets for superheavy element experiments. The focus is on 248 Cm, 249 Bk, mixed 249–251 Cf, and 254 Es, including current availabilities and new production processes. The impacts of new facilities, including the Superheavy Element Factory at Dubna, accelerator and separator upgrades at RIKEN, and proposed upgrades to the High Flux Isotope Reactor at Oak Ridge are also described. Examples of recent superheavy element research are discussed as well as future opportunities for superheavy research using actinide targets.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Nuclear fuel irradiation testbed for nuclear security applications

The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO 2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO 2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Material Control & Accountancy for Molten Salt Reactors (FY2021 Report)

There is significant domestic and international interest, investment, and research and development momentum to pursue advanced nuclear reactor technologies. Molten salt reactor (MSR) concepts display the largest variability in fuel type and design features among the current advanced concepts. MSRs have been proposed with various core designs, sizes (power), and fuel cycles. Salt-fueled molten salt systems represent the only advanced reactor type with fuel that is not in a solid form during operation. These “liquid-fueled” MSRs are unique from perspectives of fuel fabrication, spent irradiated fuel and waste components, licensing, and material control and accountability (MC&A) including the potential of fissile material holdup. The liquid fuel salt is the defining distinction in comparison to other advanced reactors that propose TRI-structural ISOtropic particle fuel pebbles, various coolant options (e.g., molten salts or metals, high temperature gas), or small modular alternatives using solid fuel variants including both light water reactors and non-light water reactors. MSRs are appealing to the nuclear energy industry because of the diverse reactor characteristics they can support including various neutron energy spectra, fueling requirements, fuel cycles, and/or fuel utilization. However, because of the significant deviation and diversity of a salt-fueled system compared to traditional solid fuel light water-cooled reactors (LWRs), the history, regulatory licensing framework, modeling capabilities, and supporting engineering technology are either lacking or, in some cases, nonexistent. Therefore, the research community is actively supporting advanced MSR development on many of these fronts in particular to assist MSR vendors with licensing requirements. ORNL is leading the research and development of respective MC&A approaches for salt-fueled MSRs. This report summarizes the research performed at Oak Ridge National Laboratory (ORNL) under the US Department of Energy, Office of Nuclear Energy, Advanced Reactor Safeguards (ARS) program to investigate safeguards and security by design concepts, licensing and regulatory considerations, and dynamic system-level modeling to understand radioisotope concentrations for salt-fueled MSRs. The report builds upon the previous research and literature, identifies the MC&A challenges inherent to a salt-fueled MSR, reviews current regulatory frameworks for LWRs and their applicability towards salt-fueled MSRs, summarizes the status and progress of an MSR dynamic modeling tool, and discusses a prospective MC&A approach based on the Molten Salt Demonstration Reactor (MSDR) model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Accelerated Irradiation Testing and Post-Irradiation Characterization: U.S.-Based Capabilities for Advanced Nuclear Systems and Radioisotope Production

Irradiation experiments and post-irradiation examinations, together referred to as irradiation testing (IRT), are prerequisites for nuclear fuel and material qualification for the deployment of new and advanced reactors, as well as radioisotope production, thereby ensuring regulatory compliance. Qualified research and test reactors (RTRs) and testing facilities are essential to enable IRT to verify performance and safety under prototypical reactor conditions. In the past, qualification of new fuels or structural materials required about 20 years. Synergist strategies, advanced tools, and qualified methods are needed to greatly reduce this timeframe of IRT and radioisotope production. This study, termed accelerated-IRT, focuses on identifying gaps and leveraging U.S.-based RTRs and material testing capabilities, leveraging the preliminary evaluation and qualification of selected RTRs to provide a generic as well as specific-case solution paths forward, ensuring adherence to stringent regulatory standards. Furthermore, IRT and radioisotope production utilizing qualified RTRs necessarily includes modeling and simulation to support the design (i.e. neutronics, thermal, and structural aspects) and manufacturing of irradiation test specimens, vehicles, capsules, apparatuses, and flow loops. In addition, IRT can be improved by applying advanced manufacturing techniques and in-pile sensors and instrumentation, as discussed in this study.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Sirius Irradiation Experiments and Post-Irradiation Examinations for Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high temperature gas-cooled fission reactor to provide thermal energy to heat the coolant which also serves as the propellant. The fuel systems of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, has been irradiated in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) enabling evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak temperatures (~2527 K). These tests named Sirius-1 (UN-W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced cracks in the fuel specimen and spalling of surface material. Uranium soot was found on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and edge features producing a laminar microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Adapting Nuclear Forensics from Light Water to Molten Salt Reactors: A Survey of Emerging Needs

Rising interest in molten salt reactors for commercial power production presents an opportunity to evaluate the techniques used to characterize materials of nuclear forensic interest. Since extensive research has been performed to identify and develop signatures of light water reactor (LWR) materials (e.g. uranium ore concentrates and uranium dioxide fuel pellets), we use this as a basis to explore possibilities for molten salt signature development. Through this comparative method, nuclear forensic signatures used today to identify the provenance of nuclear materials found out of regulatory control are adapted to molten salt reactor (MSR) fuel cycle materials. Radiological, elemental composition, isotopic composition, and model age signatures will likely not need large adaptations before being applied to MSR materials but may need to expand to be applicable to both thorium- and uranium-fueled systems. The liquid nature of molten salt fuel may erase signatures related to production and irradiation history that are informative for typical LWR materials. However, it may also lead to opportunities for new signatures, such as cooling rate–controlled morphology. Targets identified for further research include radiological attributes of fuel salts; elemental, chemical, and isotopic analysis of salts with differing production routes; morphological effects of various thermodynamic environments; and relevant fuel cycle radiochronometers. Additionally, the comparative nature of the proposed signatures implies a need for MSR-relevant databases and the production of salt standard reference materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Interpretation of Ion Irradiation and Neutron Irradiation Damage in Additively Manufactured 316 Stainless Steel using Multiscale Modeling

The accelerated adoption of nuclear energy necessitates advanced manufacturing technologies, such as additive manufacturing, to meet heightened supply chain requirements and support innovative reactor technologies. Due to the unique microstructural characteristics of additively manufactured materials under distinct solidification conditions, comprehensive evaluation of their performance in reactor environments is essential. The Advanced Materials and Manufacturing Technologies program under the Department of Energy's Office of Nuclear Energy focuses on understanding the irradiation performance and damage evolution of laser powder bed fusion 316 stainless steel, with an emphasis on integrating ion and neutron irradiation data to accelerate the development and qualification of materials for advanced nuclear reactor applications. While ion irradiation is a cost- and time-effective method, modeling and simulation are required to interpret the data for the broader range of irradiation conditions encountered in advanced reactors. In fiscal year 2025, integrated multiscale modeling and simulations were conducted to assess irradiation damage in additively manufactured 316 stainless steel. Key outcomes include predictions of chromium enrichment at grain boundaries, nickel enrichment at dislocation cell walls and void surfaces, and heterogeneous void evolution under ion and neutron irradiation conditions. Cluster dynamics simulations revealed the coarsening of voids at high irradiation temperatures and the suppression of void growth by high network dislocation density, while also demonstrating significant growth and coarsening of voids and self-interstitial atom loops at low dose rates. Machine learning-accelerated atomistic simulations highlighted the impact of the local environment and chromium concentration on vacancy diffusivity, providing key insights on the influence of composition on void swelling and radiation-induced segregation. Additionally, molecular dynamics simulations demonstrated the presence of defect production bias and a significant effect of carbon content on defect cluster behavior. These combined efforts aim to predict the performance of additively manufactured materials under various reactor conditions, supporting their qualification for nuclear reactor applications by interpreting ion irradiation data. This report underscores the potential of integrated multiscale modeling to analyze ion irradiation data in the effort to accelerate the qualification of additively manufactured materials for nuclear reactor components.

316 stainless steel↗

Measurements of emanation of Ar-37 and Ar-39 from irradiated rocks and powders

The emanation fraction of radionuclides has been highlighted as a known source of uncertainty in the estimation of radionuclide source signatures from underground nuclear tests and other nuclear activities, particularly in the case of activation products. A system was developed at Pacific Northwest National Laboratory to quantify the emanation fraction of argon from samples ranging in particle size from powder to small rocks. Seven materials, two powders and five rock types, were irradiated with fission spectrum neutrons and the emanation fraction of 37 Ar was measured. Additional measurements were made of the 39 Ar emanation for four of these materials.

36 MATERIALS SCIENCE↗

An Atom Probe Tomography Study of the Through Wall Attenuation Effect on Cu-rich Precipitate Formation in a Reactor Pressure Vessel Steel

In this study, high-Cu weld material harvested from an ex service reactor pressure vessel (RPV) steel from Unit 1 of the decommissioned Zion Nuclear Generating Station has been characterized using atom probe tomography. Samples taken from 4 different positions through the thicknesses of the pressure vessel wall from the water-side to the air-side were characterized, along with an unirradiated baseline material. In the baseline material, no precipitates were found and the Cu was observed to be fully in solid solution; however, scanning transmission electron microscopy combined with energy dispersive spectroscopy (STEM-EDS) revealed the presence of ϵ -Cu that form during processing of the material and results in the concomitant decrease of matrix Cu. Following irradiation, a high number density of nano-scale Cu-rich precipitates (CRPs) were observed, uniformly distributed throughout the matrix. The Cu content within the CRPs was found to be ~30-35 at.% regardless of location in the wall. No statistically significant variation in the compositions, mean radius, number density, or volume fraction as a function of location within the wall was observed. The measured matrix Cu level excluding CRPs contribution was found to be ~90 appm higher than the solubility limit suggesting that further nucleation and growth of the CRPs under continued operations would have occurred. These results clearly demonstrate that the neutron energy attenuation has no significant effect on the precipitation kinetics of CRPs regardless of location in the wall in high-Cu RPV steels under irradiation.

36 MATERIALS SCIENCE↗

Pulsed-Laser Irradiation Space Weathering Of A Carbonaceous Chondrite

Grains on the surfaces of airless bodies experience irradiation from solar energetic particles and melting, vaporization and recondensation processes associated with micrometeorite impacts. Collectively, these processes are known as space weathering and they affect the spectral properties, composition, and microstructure of material on the surfaces of airless bodies, e.g. Recent efforts have focused on space weathering of carbonaceous materials which will be critical for interpreting results from the OSIRIS-REx and Hayabusa2 missions targeting primitive, organic-rich asteroids. In addition to returned sample analyses, space weathering processes are quantified through laboratory experiments. For example, the short-duration thermal pulse from hypervelocity micrometeorite impacts have been simulated using pulsed-laser irradiation of target material e.g. Recent work however, has shown that pulsed-laser irradiation has variable effects on the spectral properties and microstructure of carbonaceous chondrite samples. Here we investigate the spectral characteristics of pulsed-laser irradiated CM2 carbonaceous chondrite, Murchison, including the vaporized component. We also report the chemical and structural characteristics of specific mineral phases within the meteorite as a result of pulsed-laser irradiation.

Thompson, M. S.↗

Perspective on “code qualifying” new graphite grades for use in advanced nuclear reactors*

The American Society of Mechanical Engineers (ASME) publishes the Boiler and Pressure Vessel (BPV) Code, which include guidance for the safe development, construction, and operation of boilers and pressure vessels. ASME BPV Code Section III “Rules for Construction of Nuclear Facility Components” Division 5 focuses on “High Temperature Reactors”. Subsection HH, subpart A lists the different materials properties that are to be measured and how those properties change due to different environmental conditions (oxidation and irradiation damage) for a graphite to be accepted for use in a high temperature reactor core (i.e., “Code Qualified”). Currently there are no nuclear graphite grades that are “Code Qualified” (i.e., a reactor designer can select a graphite grade and build their reactor without any additional testing), which is due in part to development of new graphite grades in the last 20 years and the lack of comprehensive programs needed to produce the data for the code cases. This perspective is going to discuss the requirements, as called out in the ASME BPV Code, that are necessary to “code qualify” a nuclear graphite grade but will primarily focus on the practical and technical challenges associated with irradiation-induced property changes and how to address these to assist with getting graphite ready for use in advanced nuclear reactors. These same technical challenges can be expected to arise for other materials being developed for advanced reactor concepts.

advanced nuclear reactors↗

Arcjet Testing of Micro-Meteoroid Impacted Thermal Protection Materials

There are several harsh space environments that could affect thermal protection systems and in turn pose risks to the atmospheric entry vehicles. These environments include micrometeoroid impact, extreme cold temperatures, and ionizing radiation during deep space cruise, all followed by atmospheric entry heating. To mitigate these risks, different thermal protection material samples were subjected to multiple tests, including hyper velocity impact, cold soak, irradiation, and arcjet testing, at various NASA facilities that simulated these environments. The materials included a variety of honeycomb packed ablative materials as well as carbon-based non-ablative thermal protection systems. The present paper describes the results of the multiple test campaign with a focus on arcjet testing of thermal protection materials. The tests showed promising results for ablative materials. However, the carbon-based non-ablative system presented some concerns regarding the potential risks to an entry vehicle. This study provides valuable information regarding the capability of various thermal protection materials to withstand harsh space environments, which is critical to sample return and planetary entry missions.

Agrawal, Parul↗