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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Material Plasma Exposure eXperiment (MPEX) High Heat Flux Bellows for Component Alignment in High Microwave Environment

The Material Plasma Exposure eXperiment (MPEX) device is a linear plasma device developed to perform plasma material interaction experiments under the conditions prototypic of a fusion reactor divertor. MPEX has multiple systems that must be precisely aligned to the plasma axis, including an electron cyclotron heating system that emits up to 400 kW of microwave power into the vacuum vessel. Five distinct systems require precise alignment on the MPEX device, thus requiring four bellows, all of which are adjacent to the plasma at a relatively high heat flux of approximately 47 kW/m2 and microwave power regions. The MPEX high heat flux bellows (HHFB) is designed to deliver 6 degrees of freedom positioning. The HHFB includes titanium–zirconium–molybdenum (TZM) inserts that are brazed into a Glidcop AL-15 body using a high-temperature braze alloy, thus blocking direct line of sight to an edge-welded bellows from the plasma and microwave screen to block microwaves from the bellows. Further, a custom ConFlat knife edge is machined into the Glidcop AL-15, so the vacuum flanges do not need a braze or weld joint on the vacuum interface. Fingerstock or copper mesh is used to restrict microwave power from entering the interstitial space between the water-cooled Glidcop AL-15 body and edge-welded bellows. Glidcop AL-15 was selected as the material of choice for the water-cooled body because it can maintain mechanical integrity at elevated temperatures, and it also allows for a high-temperature braze. TZM was selected for its machinability and compatibility with vacuum and plasma requirements. A test article is also being considered for a similarly shaped component, the MPEX limiter, that will demonstrate the integrity of the braze joint under high thermal load. Results from this testing will be extrapolated to deduce the lifetime and integrity of the HHFB design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Experimental and Computational Studies of Stress Corrosion Cracking of Alloys 308/309 and 82/182 Weldments in Corrosive and Radiation Environment

The goal of the project was to determine factors that influence SCC and IASCC in weldments found in LWR nuclear power plants. We focused on a SA508-304L SS weldment fabricated by EPRI using gas tungsten arc welding and used an aggressive BWR normal water chemistry (NWC) immersion environment. This weldment used 309L butter and 308L groove filler material. The microstructure of the 309L butter was non-uniform, exhibiting a 20–30μm thick martensitic layer closest to the SA508 interface, a 1–4 mm thick single γ austenite phase dilution zone, and a γ–δ duplex region extending to the 308L groove filler. The 308L groove filler had an entirely γ–δ duplex microstructure. Two approximately 1-inch thick 304L and SA508 plates approximately 12 by 6 square inches were joined using standard nuclear grade welding techniques. This included a post weld heat treatment of the 309L butter after application, a 0.32 cm fit-up root opening, and 57 bead lines of 308L groove filler applied in 18 layers. A 60 degree weld bevel angle was used and the 309L butter was 1.5 cm thick. Displacement cascade damage was induced using proton irradiation at the Michigan Ion Beam Laboratory. The incident proton energy was 2 MeV, the sample temperature was 360 ºC, and the calculated dpa value at 10 μm (60% of the Bragg peak depth of ~18 μm) was 5 dpa using the quick Kinchin-Pease model. Proton irradiation to this damage level required approximately 125 hours of beam time. Two types of samples were irradiated, tensile specimens and TEM bars. Samples were selected from all regions of the weldment, including the SA508-309L butter interface, the 309L-308L interface, and 308L-304L interface. The stainless steel alloys (304L, 308L, and 309L) within the heat affected zone are characterized by a duplex skeletal morphology of δ-ferrite and γ-austenite resulting from the recrystallization associated with weld fabrication. Approximately 7 to 8 mm of length along the specimen was irradiated. The gauge volume surfaces were mechanically polished and then electro-polished to remove mechanical damage from the mechanical polishing step prior to irradiation. Immersion tests were performed in a recirculating autoclave under BWR NWC conditions (2000 ppb wt. dissolved oxygen, neutral pH, 288 ºC, 10 MPa, and inlet water conductivity <100 nS/cm) to accelerate corrosion. Constant strain rate tests were performed either to failure or to approximately 5% strain. Strain rates of 10 -7 to 10 -6 mm/mm/s were used and typical immersion testing required four to six weeks to achieve failure or strains near 5%. Analysis primarily used advanced electron microscopy techniques of FIB lift out specimens.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Effect of thermal aging on microstructure and stress corrosion cracking behavior of Alloy 152 weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged to an equivalent of 60 years of service. Testing in a primary water environment of two heats of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. However, the SCC CGR response of the aged samples appears to show a deterioration in performance, confirming our previous observation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effect of Thermal Aging on Microstructure and Stress Corrosion Cracking Behavior of an Alloy 152 1st Layer Butter Weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40- 80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is critical for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for different durations up to 75,000h (equivalent to 60 years of reactor service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged at 370°C and 450°C to an equivalent of 60 years of service. However, the first weld butter layer has high levels of deformation and is highly susceptible to SCC even in its non-aged condition. Nanohardness testing confirmed the extreme hardening with aging (ΔHV ≅ 100) at two locations within this weldment. In absence of LRO, hardening is suspected to be due to thermally-induced Cr carbide precipitation and coarsening. Testing in a primary water environment of the 1st layer of Alloy 152 weld butter aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. Similarly, the SCC CGR response of the aged weld butter does not appear to show a deterioration in performance, however, the difficult-to-test weldment geometry may affect the test results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Reduction of Methane Leaks through Corrosion Mitigation Pre-treatments for Pipelines with Field Applied Coatings

Corrosion of buried, coated steel pipelines transporting natural gas is a significant source of methane emissions, from pipeline venting required for maintenance and repairs and from pipeline leaks and incidents. Corrosion of steel under field applied coatings is an important safety concern for the pipeline industry. This project investigates the application of a field applied alloy over girth welds to mitigate external corrosion of buried coated steel pipelines. Various metallic coating options were considered, which were required to meet several criteria: (1) it must resist corrosion under open-circuit or mild cathodic protection conditions, (2) it must protect the substrate steel, and (3) it must not negatively affect the adhesion of the field coating. Finite element models and lab testing were performed of alloy coating compositions to identify promising alloy types underneath disbonded coatings. Polarization curves of coating alloys were generated to provide the boundary conditions for the COMSOL model to compute potential and current distributions around coated areas. Sacrificial and corrosion-resistant metal alloy coatings were evaluated and optimized using corrosion modeling and laboratory electrochemical testing, where aluminum alloy 5356 (5% Mg) and steel alloy B9 (9% Cr) were selected. Corrosion test coupons were designed and fabricated using thermal spray aluminum 5356 and welded B9 steel overlays on API 5L grade X42 line pipe steel. The corrosion test coupons, with simulated pipe coating damage, were tested in a laboratory soil box and a field pipeline site in Texas for 3-months. Corrosion test coupons were then tested for 6-months at field pipeline sites in Texas and Tennessee to quantify corrosion rates and performance of the aluminum and steel alloys under polyethylene tape and 2-part epoxy coatings, various coating holidays, and with and without cathodic protection.

03 NATURAL GAS↗
DOE OSTI2022

Dendrite-resolved, full-melt-pool phase-field simulations to reveal non-steady-state effects and to test an approximate model

In this report we study the epitaxial, columnar growth of (multiply oriented) dendrites/cells for a spot melt in a polycrystalline Al–Cu substrate using two-dimensional, phase-field, direct numerical simulations (DNS) at the full-melt-pool scale. Our main objective is to compare the expensive DNS model to a much cheaper but approximate “line” model in which a single-crystal phase-field simulation is confined to a narrow rectangular geometry. To perform this comparison, we develop algorithms that automatically extract quantities of interest (QoIs) from both DNS and line models. These QoIs allow us to quantitatively assess the assumptions in the line model and help us analyze its discrepancy with the DNS model. We consider four sets of heat source parameters, mimicking welding and additive manufacturing conditions, that create a combination shallow and deep melt pools. Our largest DNS simulation used 16K x 14K grid points in space. Our main findings can be summarized as follows. Under AM conditions, the QoIs of line models are in excellent agreement with the full DNS results for both shallow and deep melt pools. Under welding conditions, the primary spacing of the DNS model is smaller than the prediction of line model. We identify a geometric crowding effect that accounts for the discrepancies between the DNS and line models. We propose two potential mechanisms that determine the response of the microstructure to geometric crowding.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Microstructure and mechanical properties of ultrasonic spot welding of AA7075-T6 and A380 casting aluminum alloy

Ultrasonic spot welding (USW) was employed to join AA7075-T6 and casting Al alloy A380. Under a constant welding power, the influence of the welding time duration (from 1 s to 3 s) on peak temperature, microstructure, hardness, and mechanical strength of the joints were studied. It was found that the peak temperature increased with welding time. A sound solid-state metallurgical bonding was obtained at the joint interface. The cracks occurred at the edge of the joints were eventually eliminated as the welding time increased. A layer of refined grains formed in A380 adjacent to the bonding line. Further, hardness reduction occurred in AA7075 at the joint interface and the degree of hardness reduction and the volume increased with the peak temperature. By increasing welding time, lap shear tensile tests show that the peak load increased from 5.3 kN to 6.55 kN. The corresponding interfacial shear strength ranged from 96.29 MPa to 99.72 MPa, lower than the shear strength of both base materials (331 MPa for AA7075 and 185 MPa for A380). All joint samples exhibited interfacial fracture after lap shear tensile testing, with a thin layer of A380 remnant attached to AA7075 at the center of the fractured surface.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Subcritical Californium Source Drive Noise Analysis Measurements With Unreflected Uranium (93.15) Hydride

On July 6, 1989, subcritical californium source-driven noise analysis (CSDNA) measurements were performed with bare uranium hydride cylindrical assemblies at the Los Alamos National Laboratory Critical Experiments Facility in KIVA 1. Three configurations of ~7.5 cm radius-enriched (93.15 wt. % 235 U) uranium hydride with a density of ~10 g/cm 3 cylinders were assembled with uranium hydride heights of ~11, ~14, and ~16 cm. The uranium hydride (~2 and ~3 cm high) was in thin-welded stainless steel cans. The neutron multiplication factors obtained on-line from the measured ratios of spectral densities were 0.942 ± 0.002, 0.917 ± 0.003, and 0.867 ± 0.004 for uranium hydride heights of ~16, ~14, and ~11 cm., respectively. Neutron multiplication factors for these three configurations calculated by Los Alamos National Laboratory Monte Carlo methods were 0.952 ± 0.005, 0.922 ± 0.005, and 0.860 ± 0.005, all of which are in good agreement with the measurements. The break frequency noise analysis data could be fitted to obtain the prompt neutron decay constant at all subcritical states. Some data presented in this report are from notes that are not in the logbooks. A discussion of the two different point kinetics theories of the measurements that illustrated the deficiencies of the rigorous theory at low neutron multiplication factors (k eff < 0.80) is given in an Appendix. These measurements may be the only configuration of unreflected, highly enriched uranium hydride configuration assembled. The presence of hydrogen diluted the uranium density, decreasing the reactivity, but this is offset by slowing down the neutron energy to the energy ranges of higher-fission cross sections. Extrapolations to delayed critical indicate that adding one more 3 cm can and one more 2 cm can would result in the unreflected cylindrical system being slightly above critical with a critical mass of ~37 kg, which is considerably less than the critical mass of an unreflected and unmoderated highly enriched uranium metal sphere. An unreflected, highly enriched uranium hydride sphere would have a critical mass less than 37 kg.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Tuning of Multicell Superconducting Accelerating Cavities Using Pressurized Balloons

Plastic tuning of multicell superconducting accelerating cavities is crucial in the development cycle of cavities for particle accelerators. Cavities must meet stringent requirements regarding the operating mode frequency, field flatness, and eccentricity before lining them up in a cryomodule string. After dressing bare cavities with helium vessels, the welded vessel prevents access to individual cavity cells disallowing any further localized tuning. Currently, there is no straightforward way to tune dressed cavities other than cutting the vessel and then tuning the bare cavity and dressing it back, which would significantly impact cost and schedule. In this work, we present a novel tuning technique for already jacketed cavities that is non-invasive and cost-effective. The proposed scheme employs pressurized balloons to be temporarily deployed inside the cavity as a means to localize mechanical deformation in specific cells. The proposed tuning technique was successfully utilized to recover a 9-cell 1.3 GHz tesla-style cavity.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2022

ALDWP Operational Highlights 2021

The Weapons Pit production manufacturing team completed pit build 21-1 assembly operations on December 23, 2020. This build utilized all essential flow sheet steps and included the second -01 long tube cut off extension and welding activity. This is the best build produced in the MC4597-01 pit program to date. Final out-of-line operations, non-destructive evaluation, and radiography was completed during the week of January 4. Depending on the results from high energy radiography, this pit will be considered to be a candidate for CERT-02 and shipment to Lawrence Livermore National Laboratory in support of pit certification. It is notable that this build was completed while concurrently making Plutonium (Pu) metal, producing additional castings, machining Pu parts for Build 22, completing disassembly of two pits, supporting Nightshade B Pu target characterization, and precision machining of Jasper certification base plates.

42 ENGINEERING↗
DOE OSTI2024

Machine Learning for Joint Quality Control

The use of lightweight material combinations has been highly demanded in manufacturing automotive structures. However, making robust dissimilar material joints of such lightweight materials is still challenging. A significant barrier to achieving high-quality and repeatable joint performance is a deficient understanding of the relationship between the welding process, joint attributes, and joint performance. In this context, welding factors refer to material, equipment, environment, and process parameters, while joint features comprise specific microstructural attributes of the weld such as nugget size, heat affected zone (HAZ) topology, intermetallic layer thickness, and sheet thickness reduction. Joint performance is quantified in terms of strength (e.g., tensile shear, coach peel, cross-tension), weld size, and hardness, among other factors. While there have been many attempts to establish this process-structure-property relationship by developing a model derived from the associated physics and first principles, the complexity of the joining processes compounded by the complex interactions with different materials in an automotive assembly line environment, has hindered the usefulness of such attempts. The complexity is further exacerbated using different stacking materials, especially comprising dissimilar material combinations. In practice, the common approach has been the laborious process of creating welds, characterizing them, and then physically testing them through experimentation. With the emergence of artificial intelligence (AI) methods, an alternative pathway to eliciting the desired process-structure-property relationship at an accelerated pace is to use a data-driven approach by employing machine-learning (ML) techniques. This approach is benefitted by the availability of large streams of data, generated through years of research and testing by original equipment manufacturers, in the form of material, process, environmental, equipment, microstructural, and bulk-scale performance information from multimodal, multiscale sensors making measurements from laboratory-scale to production-scale processes. During Phase I efforts, which ended in fiscal year (FY) 2021, the Oak Ridge National Laboratory and Pacific Northwest National Laboratory (ORNL/PNNL) team demonstrated the effectiveness of different ML/AI frameworks in modeling complex relationships between resistance spot welding (RSW) process parameters, weld attributes, and joint properties using a subset of data from General Motors (GM). In FY 2022, the project team further refined and expanded their respective ML models to analyze additional welds with new weld stack-ups and materials to enhance the ML model predictive capability. ORNL extended its unified deep neural networks (DNN) ML training and prediction framework with new data streams of process parameters, and PNNL extended its model describing RSW process parameters’ associations with weld attributes. In FY 2023, the project team completed the development of the AI/ML architecture for analyzing aluminum/steel joints manufactured by GM via RSW and transitioned into the inline welding quality monitoring task for steel/steel RSW joints provided by GM.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Inspection Tools for Hanford Tanks and Waste Transport Systems - 20444

Recent integrity issues in the double-shell tanks (DST) at Hanford have motivated the need for developing innovative tools that can provide information regarding the health of the tanks. These issues include the primary linear failure of AY-102 and recent concerns of thinning in the DST secondary liners. Other concerns include erosion or corrosion on transfer lines and processing pipes. In recent years, Florida International University (FIU) has supported DOE-EM by developing tools that can assist in understanding the health of tanks and waste transport system at Hanford. More specifically, FIU has developed a miniature rover that can be deployed through small risers and gain access to refractory slots in the Hanford DSTs. It traverses through the slots on the primary liner upside down via magnets to avoid debris in the slots. Recent modifications include a spring loaded magnetic arm that allows the unit to traverse over large weld seams but does not reduce the overall functionality of the system. The system has also incorporated a control capsule that will sit within the annulus of the tank and improves on issues related to signal and voltage loss over long tether lines. Lastly, the unit is also being augmented with a second module that will include a means to prep the tank surface and obtain a thickness measurement using a dual element ultrasonic sensor. FIU has also developed a 6-inch peristaltic pipe crawler that operates similar to the previous pipe crawlers developed at FIU. This crawler is a marsupial type crawler that will navigate through the 6-inch drain lines, and deploy a small rover at the entrance of the drain slots. This rover will also traverse upside down on the secondary liner and provide information regarding the health of the liner. An initial prototype for the crawler unit has been developed, assembled and tested. The system is pneumatic and consists of three modules - two grippers and one extender. The validation tests are conducted in a mockup of the drain lines and demonstrate the system's ability to navigate through long pipe lengths and elbows. In this paper, a general overview of each of the aforementioned systems is presented along with results from testing and demonstrations that shows the applicability of each system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

THE ROLE OF FRACTURE PROPERTIES ON LAP JOINT STRENGTH OF FRICTION STIR WELDED AA7055-T6 SHEETS

Friction stir lap welded (FSLW) joints have weight-saving potential in aluminum-intensive automotive assembly. However, FSLW also modifies the material microstructure close to the joint. Optimizing the FSLW process requires understanding the relationship between the strength and the joint's microstructure. In previous studies, efforts have been dedicated to determining the effects of local softening, the shape of the oxide line, and porosity. However, the impact of changes to the fracture properties on the joint's strength has not been studied. In this work, strength testing, and simulations, aided by material characterization, were used to determine the role of fracture properties on the shear strength of a 3-sheet (aluminum alloys 7055-7055-6022) lap joint. Characterization involved testing for fracture properties in the weld region. This data was then implemented into finite element simulations. As a result, the joint strength was predicted with a deviation of less than 6% from the experimental value. Comparison with strength prediction using only the base metal properties indicates that fracture property in the nugget region determines the strength of AA7055 FSLW.

Finite element analysis, Friction stir welding, ma↗
DOE OSTI2024

Neutron Diffraction Residual Stress Study of a AA2219-T87 Self-Reacting Friction Stir Weld

2219-T87 is a precipitation hardenable aluminum-copper alloy which sees wide use in structural aerospace components. Thick panels of this alloy are joined via self-reacting friction welding (SRFSW); however, this thermomechanical process can result in significant loss of the alloy’s strengthening precipitates and large residual stresses which are detrimental to mechanical behavior and performance. High-resolution maps of the residual strain and stress states along the normal, transverse, and longitudinal directions of 2219-T87 SRFSW were obtained using neutron diffraction measurements. Residual stress had the highest tensile value in the heat-affected zone (HAZ) and the largest compressive stress in the base metal region of the normal and transverse directions. Line profile residual stress distributions displayed an “M”-shaped distribution in each direction with the longitudinal direction being the most pronounced while 2D residual stresses displayed an hourglass-shaped pattern. Further, average maximal longitudinal residual stress values ranged from 60.6 to 85.6% of yield. Microhardness testing across the transverse weld section produced a “V”-shaped curve with slight hardness recovery in the stir zone. Maximal and minimal microhardness values were observed in the base metal (154 HV) and thermomechanically affected zone (82 HV), respectively. Microstructural evolution was recorded using optical microscopy and showed decreasing grain size from the HAZ to the stir zone.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A Preliminary Radiological Risk Assessment Model for Disposition of Remote-Handled Transuranic Wastes at Los Alamos National Laboratory Area G - 20116

The U.S. Department of Energy (DOE) operates a low-level radioactive waste (LLW) disposal site at Material Disposal Area G, in Los Alamos, New Mexico, USA. Area G has been the primary LLW disposal site for Los Alamos National Laboratory (LANL) since the 1960's. In addition to LLW, Area G is host to a variety of other wastes, the disposition of which must be determined before closure of the site. A probabilistic Radiological Risk Assessment (RRA) for Area G is used in order to support decision making regarding some wastes that are not addressed in the extant Area G Performance Assessment (PA) and Composite Analysis (CA). Between 1979 and 1987, 33 special shafts were augered into the Bandelier Tuff at Area G. This volcanic tuff is present across Pajarito Plateau on the eastern slopes of the Jemez Mountains, and varies widely in its consistency, from weakly indurated non-welded layers to welded layers that uphold the mesa cliffs of the Plateau. These mesas are home to LANL, Area G, and the townsites of Los Alamos and White Rock, with residences about 1400 m from Area G. The 33 Shafts were lined with steel casing, and contain remote-handled (RH) transuranic wastes (TRU) resulting from experiments and analysis performed in special glove boxes at the Chemistry and Metallurgy Research (CMR) facility at LANL. Some of these wastes originated as used nuclear fuel. The purpose of the Area G RRA is to evaluate the potential future risk to humans and the environment from the RH TRU in the 33 Shafts in the context of the risk associated with the surrounding wastes at Area G. The analysis is responsive to expectations outlined in DOE Order 458.1, Radiation Protection of the Public and the Environment, and is informed by the Manual and Guidance accompanying DOE O 435.1, Radioactive Waste Management. Because the waste meets the definition of TRU, the regulatory context necessarily takes into consideration the regulation governing the disposal of TRU from the U.S. Environmental Protection Agency (EPA): 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. Given the broader regulatory context for the RRA, the analysis is subject to different assumptions from those made in the existing DOE O 435.1 PA and CA, such as allowing for future occupation of the site. The analysis begins with a comprehensive evaluation of features, events, processes, and exposure scenarios (FEPS) for Area G and the wastes it contains. These FEPSs are screened to eliminate from further consideration those of extremely low probability and/or consequence, and a conceptual site model (CSM) is subsequently developed. The scope and structure of the Area G RRA Model is informed by this CSM, and the Area G RRA Model is developed using the GoldSim systems analysis modeling platform. This paper presents the initial version of a defensible, transparent, and reasonably realistic model, which is based on the state of knowledge of the wastes, the site, and the FEPSs that govern contaminant transport from wastes into the environment and subsequent exposures to humans and other biota. Probabilistic model input distributions represent uncertainties inherent in the real and modeled systems. The results of the Area G RRA Model inform decisions regarding the disposition of the RH TRU in the 33 Shafts. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Extending Ultrasonic Welding Techniques to New Material Pairs (FY 2023 Annual Progress Report)

Modern multimaterial vehicles require joining of various lightweight materials, such as aluminum (Al) and magnesium (Mg) alloys and carbon fiber reinforced polymers (CFRP), with advanced high-strength steels together to form a high-performance and lightweight body structure. A variety of joining methodologies (e.g., resistance spot welding, adhesive bonding, linear fusion welding, hemming, clinching, bolting, riveting) have been attempted by the automotive industry to join different materials. Often, these joining techniques are limited to only certain material combinations. For capital and operational cost, automobile original equipment manufacturers need to limit the number of joining technologies implemented on an assembly line.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Getting things done onsite: Heat Source production never ramps down

The COVID-19 pandemic has changed many norms in the world and at LANL. During the rapid transition of sending many to work from home in March, some programs were deemed mission critical, with the expectation to continue driving their missions forward. One of those essential programs is Heat Source Technologies, a group within the Actinide Material Power & Processing (AMPP) Division at TA-55. Within the Associate Laboratory Directorate of Weapons Production (ALDWP), the group responsible for the Pu-238 heat sources continued their work nearly uninterrupted as many others made the mandatory transition to work remotely. This group, known as AMPP-1, Heat Source Technologies, works on the production line producing heat sources from Plutonium (Pu)-238, which is an essential mission for National Security. This group is comprised of nearly 90 employees and operates out of the TA-55 Plutonium Facility (PF-4 and PF-5) producing heat sources. Heat source production includes the operations that purify, process, and weld Pu-238 oxide into encapsulated heat sources that are essential for space and defense missions.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2024

Transmission electron microscopy of ion irradiated ODS MA956 samples

Oxide dispersion strengthened (ODS) alloys are promising candidate materials for the next generation advanced nuclear reactors due to their superior irradiation resistance and mechanical properties. To better understand the effect of irradiation on MA956, it is essential to study higher dose (50-100 dpa) samples, so that the general trend of microstructural evolution and the resulting radiation-hardening can be deduced. Currently, ion irradiations are considered the only way to achieve doses beyond ~50 dpa in a practical time frame relevant to alloy and welding development programs. This dataset contains TEM characterization results of ODS MA956 samples that were ion irradiated under different conditions: Sample #5 (1.25 dpa at 190℃); Sample #9 (50 dpa at 190℃); Sample #15 (1.25dpa under 320℃); and Sample 17 (25 dpa at 320℃). TEM characterization focused on the irradiation induced defects (dislocation lines and loops) using the on-zone axis bright field STEM technique. These data were collected using a FEI Tecnai G2 F30 S/TEM at Microscopy and Characterization Suite (MaCS) at Center for Advanced Energy Studies (CAES), Idaho Falls, ID. This project (ion irradiation and TEM studies) was supported by the U.S. Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-05ID14517 as part of Nuclear Science User Facilities award #18-14784 (PI: Ramprashad Prabhakaran, PNNL).

Prabhakaran, Ramprashad↗