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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 829 records · Page 46
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Advanced Test Reactor National Scientific User Facility (ATR NSUF) (Monthly Report, December 2014)

PNNL has completed sectioning of the U.C. Berkeley hydride fuel rodlet 1 (highest burn-up) and is currently polishing samples in preparation for optical metallography. A disk was successfully sectioned from rodlet 1 at the location of the internal thermocouple tip as desired. The transition from annular pellet to solid pellet is verified by the eutectic-filled inner cavity located on the back face of this disk (top left) and the solid front face (bottom left). Preliminary low-resolution images indicate interesting sample characteristics in the eutectic surrounding the rodlet at the location of the outer thermocouple tip (right). This sample has been potted and is currently being polished for high-resolution optical microscopy and subsequent SEM analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Complete Initial Scoping Tests on the Incorporation of Novel Loaded Iodine Getters into GCM

This study encompasses initial scoping tests on the incorporation of a novel iodine loaded getter material into the Sandia developed low temperature sintering glass ceramic material (GCM) waste form. In particular, we studied the PNNL Ag-I-Aerogel. Optical microscopy indicates inhomogenous samples based on particle sizes and variations in color (AgI vs Ag/AgO on silica). TGA/MS data when heated in air indicates loss of iodine and organics (CO2) between 250-450°C a total of ~15wt% loss, with additional / small iodine loss when during 550°C hold for 1 hr. TGA/MS data when heated in N2 indicates less organic and slightly less iodine loss below 550°C, with no loss of iodine in 550°C 1 hour hold. Furthermore, a substantial mass loss of sulfur containing compounds is observed (m/e of 34 and 36) between 150 – 550°C in both air and N2 sintering atmospheres. In an effort to capture iodine lost to volatilization during heating (at temps below glass sintering temperature of 550°C), we added 5 wt% Ag flake to the AgIaerogel. Resulting data indicates the iodine is retained with the addition of the Ag flake, resulting in only a small iodine loss (< 1wt%) at ~350°C. No method of curtailing loss of sulfur containing compounds due to heating was successful in this scoping study.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Demonstration of thermal control, microstructure control, defect mitigation and process parameter database generation for Ti-6Al-4V Direct Digital Manufacturing - Understanding defect mitigation and process parameter database generation for direct digital manufacturing

Researchers from Manufacturing Demonstration Facility (MDF) at Oak Ridge National Laboratory (ORNL) worked with Applied Optimization (AO) to understand and evaluate the propensity for defect formation in builds manufactured using DM3D-POM laser direct metal deposition. The main aim of this collaboration was to understand the character of powder jet behavior as a function of the nozzle parameters such as cover gas, carrier gas, and shaping gas. In order to evaluate the sensitivities of the parameters used in model, various experiments were performed with in-situ monitoring of the powder stream characteristics using a high speed camera. A wide variety of conditions while keeping the hopper motor rpm constant, including laser power and travel speed were explored. The cross sections of the deposits were characterized using optical microscopy.

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

Microscopic Examination of the Corroded Hagan Container Used to Store Molten Salt Extraction Residue XBPS333

A microscopic examination was performed on a Hagan container after corrosion and a white powder were found on the outside surface of the lid and filter cover while in storage in the vault. The inside surfaces of the container were covered by a wipeable brown coating. The coating was removed and portions of the container were sectioned and imaged with optical microscopy. General corrosion was found throughout the inside of the container, and pitting corrosion was observed in several locations on the lid and sidewall. This report describes the type and extent of the corrosion observed and its impact on the storage lifetime of the container.

36 MATERIALS SCIENCE↗
DOE OSTI2018

U 3 Si 2 Fabrication and Testing for Implementation into the BISON Fuel Performance Code

A creep test stand was designed and constructed for compressive creep testing of U 3 Si 2 pellets. This is described in Chapter 3. Creep testing of U 3 Si 2 pellets was completed. In total, 13 compressive creep tests of U 3 Si 2 pellets was successfully completed. This is reported in Chapter 3. Secondary creep model of U 3 Si 2 was developed and implemented in BISON. This is described in Chapter 4. Properties of U 3 Si 2 were implemented in BISON. This is described in Chapter 4. A resonant frequency and damping analyzer (RFDA) using impulse excitation technique (IET) was setup, tested, and used to analyze U 3 Si 2 samples to measure Young’s and Shear Moduli which were then used to calculate the Poisson ratio for U 3 Si 2 . This is described in Chapter 5. Characterization of U 3 Si 2 samples was completed. Samples were prepared and analyzed by XRD, SEM, and optical microscopy. Grain size analysis was conducted on images. SEM with EDS was used to analyze second phase precipitates. Impulse excitation technique was used to determine the Young’s and Shear Moduli of a tile specimen which allowed for the determination of the Poisson ratio. Helium pycnometry and mercury intrusion porosimetry was performed and used with image analysis to determine porosity size distribution. Vickers microindentation characterization method was used to evaluate the mechanical properties of U 3 Si 2 including toughness, hardness, and Vickers hardness. Electrical resistivity measurement was done using the four-point probe method. This is reported in Chapter 5.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Understanding Part to Part Variability During Directed Energy Deposition Processes Using In Situ and Ex Situ Process Characterization

The overall goal of this technical collaboration is the use of in situ and ex situ characterization techniques to understand part to part variability in Ti-6Al-4V parts fabricated using directed energy deposition (DED) process, and subsequently qualify parts fabricated using AM. The primary goal of the Phase-1 of this CRADA was to demonstrate the feasibility of capturing the above changes using a combination of high speed stereo imaging and in situ monitoring using two wavelength pyrometers. The in situ monitoring was then correlated using ex situ characterization performed using optical microscopy. The results from phase-1 clearly show that changes to the process and shielding gases have a significant impact on the geometric conformity of the build and the build layer height. The results from phase-1 show that in addition to using IR thermography as a tool to measure the temperature and gradients, we could use it to monitor the build topography and the layer height and use the information to develop closed loop control systems. The phase-1 results show that there is a narrow zone under which the process must operate in order to be “passively stable” or autocorrect for over building. In addition, the results demonstrate that when operated within this range the build autocorrects and there is a drastic reduction in the interlayer lack of fusion defect formation.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Developing Novel Multifunctional Materials for High-Efficiency Electrical Energy Storage

This is the final report for the program “Developing Novel Multifunctional Materials for High-efficiency Electrical Energy Storage”, funded by The U.S. Department of Energy’s (DOE) National Energy Technology Laboratory (NETL), and performed by University of Tennessee, from September 01, 2013 to August 31, 2018. Securing a sustainable energy economy by developing affordable and clean energy from renewable energy is a central element to the mission of DOE/NETL. To further this mission, NETL funds research and development of novel multifunctional materials that can be used for high-efficiency electrical energy storage. The main objective of this research program is to conduct fundamental and applied research that lead to successful development and demonstration of novel multifunctional materials, which can improve the efficiency and reduce the cost of hydrogen production with water electrolysis. The specific objectives of this research program include: 1) Design, fabrication and demonstration of thin/well-tunable liquid/gas diffusion layer (TT-LGDLs) with different pore morphologies, including pore sizes and pore porosities. 2) Development of comprehensive computational models of proton exchange membrane water electrolyzer cell (PEMECs) for the investigation of influence of different operating conditions and physical design parameters on its performance. 3) Surface modification of TT-LGDLs for better catalyst access and product/reactant removal, and smaller interfacial contact resistance (ICR). To achieve the above challenging goals, a multidisciplinary research team was formed to work on fundamental and applied research on the following three major technical areas: 1) Design, fabrication and demonstration of thin/well-tunable LGDLs (TT-LGDLs) by using photolithography and wet etching of thin titanium (Ti) foils. During the project period, the team has successfully fabricated TT-LGDLs with different pore morphologies, including pore sizes and pore porosities. The manufacturing steps mainly include cleaning thin Ti foil, supporting the foils, applying photoresist, patterning the photoresist with designed mask, exposing and developing, etching, removing photoresist, and lift-off and cleaning. The manufactured TT-LGDLs were characterized with optical microscopies and scanning electron microscopies. Their pore morphologies, including pore size and porosity, were derived based on advanced image process. The feasibility for scale-up manufacturing has been demonstrated. The TT-LGDLs have been successfully investigated for water splitting in PEMECs, which showing better performance than their counterparts fabricated with the traditional fiber-based materials. 2) Comprehensive computational modeling of a PEMEC. A comprehensive computational model has been successfully developed, and the influence of different operating conditions and physical design parameters on its performance has been investigated, including operating temperature, pressure, exchange current density, electrode thickness, membrane thickness, and interfacial resistance. Results obtained from the present model provide a comprehensive understanding of design parameter effects and consequently improve the design/performance in a PEMEC. 3) Surface modification of TT-LGDLs. The proposed surface modifications of TT-LGDLs by sputter coating and/ electroplating have been successfully demonstrated. It has been found that the performance of PEMECs can be improved by both methods, and the thickness of the electroplated gold thin layer has limited effects on the performance. With the help of the surface modifications, the PEMECs achieved the best ever performance compared with the public literatures. In addition to completion of the proposed method, microporous layers (MPLs) with both irregular micro and sphere nano titanium particles on novel TT-LGDLs was developed for the first time. The MPLs were fabricated by a low-temperature air spraying method, which provide a better catalyst access and product/reactant removal. In summary, under the close collaborations among the research team, all the tasks have been completed and the proposed research objectives have been fulfilled. All the technical milestones have been achieved and the project has been completed successfully. The technical breakthroughs obtained under this project provide novel LGDLs with well-tunable pore morphologies that can significantly improve the performance of water splitting and energy storage. This development can be directly used in existing and next generation energy storage and fuel systems.

08 HYDROGEN↗
DOE OSTI2018

Postirradiation Examination of the ATF-1 Experiments - 2018 Status

The Advanced Fuels Campaign in collaboration with industry are working to develop enhanced accident tolerant fuels for light water reactors. The initial irradiations of these concepts were performed in the Idaho National Laboratory Advanced Test Reactor using simple drop-in style irradiations. These irradiations are collectively referred to as ATF 1. The postirradiation examination of ATF 1 has begun and some concepts have now completed initial examination. This report describes the results of examinations on ATF 1 irradiations designed to test the feasibility of enhancing UO 2 thermal conductivity with additives, ATF 1 irradiations of uranium silicide, and the current status of ATF 1 irradiations that are in process at the hot cell. Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO 2 SiC) and diamond (UO 2 diamond) have been irradiated to low burnup. The conditions of this irradiation test and subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction under representative light water reactor power conditions. Both non destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO 2 SiC pellets revealed enhanced cracking when compared to a UO 2 rodlet also irradiated under similar conditions. Additionally, instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300 °C. The microstructure of the diamond added UO 2 was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature. The postirradiation examination results cast doubt on the suitability of these additives to improve UO 2 fuel performance in a way that would lead to enhanced accident tolerance. Additionally, postirradiation examination data of U 3 Si 2 fuels at low burnup (i.e., < 20 GWd/tHM) for application in advanced Light Water Reactor (LWR) is presented. The U 3 Si 2 pellets show limited cracking in comparison to the expected behavior of UO 2 at same power level. In addition, gamma scanning data did not reveal migration of fission products. Minor homogeneous hardening along the pellet radius due to accumulation of fission products and radiation damage was measured by microindentation. The fission gas release and swelling remains very low. Formation of fission gas bubbles resolvable with optical microscopy occurs from the pellets center outward to approximately 60% of the fuel pellet radius. The overall data suggest a good performance of this accident tolerant fuel candidate at low burnup. Finally, the current status of postirradiation examination on other ATF 1 irradiations will be discussed. These irradiations include UO 2 clad in an FeCrAl alloy, an alternative uranium silicide compound U 3 Si 5 also clad in a FeCrAl alloy, and an interaction test between UO 2 and different FeCrAl alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Report for the Forensic Laboratory Analysis of Hose-in-Hose Transfer Line used in C-Farm Retrievals

Two 6-ft hose-in-hose transfer line sections used in the retrieval of C-Farm single-shell tanks were delivered to the 222-S Laboratory for forensic analysis to determine the amount of wear and degradation that had occurred during the lifetime of the hose in the tank farms on the Hanford Site. One hose section was used for supernatant liquid transfers, and the other section was used for slurry transfers. Cutting the hoses in a hot cell enabled visual and photographic examination of the inside surfaces of the hoses. The hoses were further investigated by optical microscopy and scanning electron microscopy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Processing and Characterization of U 3 Si 2 at the MiniFuel Scale

This document details efforts at Los Alamos National Laboratory (LANL) to process and characterize U 3 Si 2 at miniature fuel (MiniFuel) scale geometries. Eight MiniFuel pellets were fabricated using standard powder metallography techniques. Final specimens had diameters of ~2.8 mm and thickness ranging from ~244-290 µm. Densities of MiniFuel samples determined via geometric methods showed values ranging from 92.9-96.6% of the theoretical U 3 Si 2 density (12.2 g/cm 3 ). Microstructural characterization via scanning electron and optical microscopy show relatively dense microstructures with grain sizes ~52±1 µm. Powder processing parameters were shown to influence the overall quality of the pellet and is discussed in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Compatibility of polymers in super-critical carbon dioxide for power generation systems: High level findings for low temperatures and pressure conditions

Polymers such as PTFE (polytetrafluorethylene or Teflon), PEEK (polyetheretherketone), EPDM (ethylene propylene cliene monomer) rubber, Viton, EPR (ethylene propylene rubber), Nylon, Nitrile rubber, and perfluoroelastomers are commonly employed in super critical CO 2 (sCO2) energy conversion systems. O-rings and gaskets made from these polymers face stringent performance conditions such as elevated temperatures, high pressures, pollutants and corrosive humid environments. Critical knowledge gaps about polymer degradation from sCO2 exposure need to be addressed. To understand these effects, we have studied nine commonly used polymers subjected to elevated temperatures under isobaric conditions of sCO2 pressure. The polymers (PEEK, Nylon, PTFE, EPDM, Nitrile rubber, EPR, Neoprene, perfluoroelastomer FF 202 and Viton) were exposed for 1000 hours at 100°C to 25 MPa sCO2 pressure in an autoclave. In a second study, elastomers perfluoroelastomer (FF202) and EPDM were exposed to 25 MPa sCO2 for 1000 hours at 150°C. Samples were extracted for ex-situ characterization at t = 200 hours and then at the completion of the test at t=1000 hours. The polymer samples were examined for physical and chemical changes by Dynamic Mechanical and Thermal Analysis (DMTA), Fourier Transform Infrared (FTIR) spectroscopy, and compression set. Density and mass changes immediately after removal from test and 48 hours later, and optical microscopy techniques were also used. Microcomputer tomography (micro CI) data was generated on select specimens. Super-critical CO 2 effects have been identified as either physical or chemical effects. For each polymer, the dominance of one type of effect over the other was evaluated. Attempts were also made to qualitatively link sCO2 effects such as lowering or increase in glass transition temperatures, storage modulus changes, mass and compression set changes, chemical changes seen in FTIR analyses and blister and void formation seen post-exposure to polymer microstructure-related mechanisms such as plasticization of the polymer matrix, escape of volatiles from the polymer during depressurization, and filler and plasticizer effects on microstructure with rapid depressurization rates.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Measurement of kernel swelling and buffer densification in irradiated UCO and UO 2 TRISO fuel particles from AGR-2

The second Advanced Gas Reactor irradiation (AGR-2) featured UCO tristructural isotropic (TRISO)-coated particle fuel and, for comparison purposes, UO2 TRISO fuel. Particles from three UCO fuel compacts and one UO2 fuel compact were chosen for analysis of kernel swelling and buffer shrinkage. Both kernel swelling and buffer shrinkage are common elements of TRISO fuel performance models. More recently, the AGR program has determined that buffer shrinkage may impact the integrity of the inner pyrolytic carbon (IPyC) layer, which may impact SiC layer integrity (particularly during high-temperature heating tests). Therefore, measurements of buffer irradiation-induced shrinkage may be useful in improving existing models and developing new models that capture buffer-IPyC-SiC-fission product interactions observed in recent AGR post-irradiation examination (PIE). TRISO particles were mounted in epoxy, ground, polished, and imaged via optical microscopy in four separate iterations. Each of the four iterations revealed progressively deeper cross sections within the particles. Images collected from each iteration were analyzed for the circumferences/radii of the kernel, buffer, IPyC, and SiC. Spheres were fit to each set of measurements to generate spherical radii representative of the components of the fuel particle. From these radii, volumes were computed and compared to the as-fabricated volumes. UCO kernel swelling was similar among the three AGR-2 compacts ranging from approximately 28 to 32%. This is only slightly higher swelling than was measured in an earlier study of an AGR-1 compact. However, the burnup at which this swelling occurred in AGR-2 was lower than in AGR-1, indicating a faster swelling rate for the AGR-2 kernels for the irradiation conditions covered by these compacts. The AGR-2 UO2 kernel swelling was less, at 9.9%, comparable to other studies of UO2 kernel swelling. While AGR-2 UCO and UO2 kernels demonstrated different rates of swelling, the extent of buffer shrinkage was similar among the two fuel types, ranging from 24-28%. For comparison, a prior study showed that AGR-1 Compact 1-3-1 experienced 39% buffer shrinkage. A commonly-observed post-irradiation particle morphology is where the buffer pulls away from the IPyC (at least in some regions of the particle). Besides the possibility of the buffer degrading the IPyC when it pulls away from the IPyC, this gap is significant because it may also reduce heat transfer within the particle. Average AGR-2 buffer-IPyC gaps were similar for the UCO and UO2 particles, ranging from 22.2 to 26.0 µm. This is similar to what was measured for AGR-1 Compact 1-3-1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Development of a Testing Protocol for C7 Carbon Fiber Core Conductors

To avoid the high costs and environmental and regulatory challenges of building new lines to meet the power demand, utilities are searching for ways to increase power transmission capacity on existing lines and thereby. This could involve upgrading the network lines with conductors capable of transmitting greater capacity. In recent years, conductor manufacturers have begun producing new conductors that can operate at higher temperatures without violating present electrical clearances to ground and other objects. One new conductor technology is a carbon composite core with PPS polymer matrix and PEEK capping layer composite conductor, manufactured by Southwire Inc. The conductor uses new materials, which have not previously been used for overhead lines. The project made an initial characterization of the composite core to define the material used in the manufacture of the composite. Several analytical procedures were then conducted to evaluate the physical-chemical impacts on the composite as a result of thermal exposure. These procedures included thermogravimetry, dynamic mechanical analysis, thermal mechanical analyses, and optical microscopy. Mechanical three-point flexural tests, tensile tests, creep tests, and dynamic vibration tests were then performed to estimate the potential degradation after thermal aging. The report provides the basis to develop a qualification testing protocol for the carbon composite core at targeted service environments. The work covers the design of the qualification testing protocol and the follow-on validation testing to support the proposed methods. The research includes the following approaches: initial characterization, performance of thermal aging, characterization of the aged material, comparison of aged properties or characteristics to the original, and concluding by the findings through the developed qualification procedures.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2020

Intermediate-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at The LWR Relevant Temperature

Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) on the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Based on a comprehensive microstructure and property screening performed in Phase-1 of the ARRM program, a total of five alloys, together with 316L and X-750 as references, were down-selected for Phase-2 neutron irradiation studies. Because thermal aging could exert a synergistic effect on neutron irradiation due to the low neutron damage rate on the order of 10 –7 displacements per atom per second (dpa/s), Grade 92 (one of the five down-selected alloys) and two heats of 316L were selected in this task to study the effect of aging at 350°C for 12.5–12.7 kh on microstructure and mechanical properties. The aging time is approximately corresponding to 5 dpa neutron irradiation, which can be served as a reference for the 5-dpa-irradiated samples of the alloys to help understand the independent neutron damage influence on microstructure and mechanical properties. Optical microscopy and scanning electron microscopy were used for microstructural characterization. Hardness, tensile, Charpy impact toughness, and fracture toughness in the ductile regime of the aged samples were examined. The aging of Grade 92 led to the formation of many Laves phase in sizes of ~100–200 nm and resulted in some reduction in hardness and yield/ultimate tensile strength with some increases in uniform and total plastic elongations, which may have helped the enhancement of Charpy impact toughness, e.g., ~4 J increase in upper-shelf energy and 20.5°C reduction in ductile-brittle transition temperature compared with the unaged condition. The fracture toughness of the aged Grade 92 showed decent toughness of ~303 MPa√m (K Jq ) with ~87 tearing modulus at 22°C, which reduced to ~241 MPa√m (~20% reduction) with ~75 tearing modulus at 300°C. The two heats of 316L are differentiated by their amounts of δ-ferrite, i.e., ~1 vol% in heat T1103 and ~4 vol% in heat N5B8. The two heats of 316L were subjected to 15% cold work (CW) as this condition is often used in nuclear reactors to help trapping radiation-induced defects and thus postpone the steady state swelling stage. The aging did not result in any noticeable microstructure changes, except for possible segregations at grain boundaries that need to be further investigated. The aging resulted in slight reduction in hardness of 316L-T1103 and slight increase in hardness of 316L-N5B8, which is consistent with their tensile testing results. The 15%CW might have introduced some inhomogeneity, resulting in large standard deviations in hardness, which were reduced after the aging. The aged 316L followed the same trend as the unaged condition for the strength and elongation results from the tensile testing. Minimum elongations appeared within ~300–600°C, above which the elongations seem to increase. Unlike the negligible or minor changes in microstructure, hardness and tensile results of the two heats of 316L, their fracture toughness showed noticeable difference. The aged 316L-T1103 showed good toughness of ~350 MPa√m with ~80 tearing modulus at 22°C, which reduced to ~271 MPa√m (~23% reduction) with ~83 tearing modulus at 300°C. In contrast, the aged 316L-N5B8 showed lower toughness of ~244 MPa√m with ~53 tearing modulus at 22°C, which reduced to ~180 MPa√m (~26% reduction) with ~31 tearing modulus at 300°C. The preliminary results indicate that the presence of high volume of δ-ferrite would noticeable impair fracture toughness although it may not influence hardness and tensile properties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Material Property Determinations of P-Tunnel Core in Support of UNESE

A critical component of the Underground Nuclear Explosion Signatures Experiment (UNESE) program is a realistic understanding of the post-detonation processes and changes in the environment that produce observable physical and radio-chemical signatures. Rock and fracture properties are essential parameters for modeling underground nuclear explosions. In response to the need for accurate simulations of physical and radio-chemical signatures, an experimental program to determine porosity, hydrostatic and triaxial compression, and Brazilian disc tension properties of P-Tunnel core was developed and executed. This report presents the results from the experimental program. Dry porosity for P-Tunnel core ranged from 8.7%-55%. Based on hydrostatic testing, bulk modulus was shown to increase with increasing confining pressure and ranged from 1.3GPa-42.3GPa. Compressional failure envelopes, derived from wet samples, are presented for P-Tunnel lithologies. Brazilian disc tension tests were conducted on wet samples and, along with triaxial tests, are compared with dry tests from the first UNESE test bed, Barnwell. P-Tunnel core disc tension test strength varied nearly two orders of magnitude between lithologies (0.03MPa-2.77MPa). Material tested in both tension and compression is weaker wet than dry with the exception of Strongly Welded Tuff in compression which is nearly identical in compressive strength for confining pressures of OMPa and 1 OOMPa. In addition to the inherent material properties of the rocks, fractures within the samples were quantified and characterized, in order to identify differences that might be caused by the explosion-induced damage. Finally, material property determinations are linked to optical microscopy observations. The work presented here is part of a broader material characterization effort; reports are referenced within.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Sister Rod Destructive Examinations (FY20) Appendix B: Segmentation, Defueling, Metallograhic Data and Total Cladding Hydrogen

As a part of the US DOE-NE High Burnup Spent Fuel Data Project, ORNL is performing destructive examinations (DEs) of high burnup (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the “sister rods,” are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO®, and M5®. The DEs are being conducted to obtain a baseline of the HBU rod’s condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding are being tested. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities.This report documents the status of the ORNL Phase 1 DE activities related to: rough segmentation, defueling, optical microscopy, and cladding total hydrogen measurements for 7 Phase 1 sister rods and outlines the DE tasks performed and the data collected to date. The results of these detailed examinations and others are summarized in the Sister Rod DE Status Report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Connecting Dynamic Strain Aging to Deformation Processing in Magnesium-Calcium-Based Alloys

Fuel consumption can be improved by reducing car weight with substitution of aluminum and steel parts with magnesium (Mg). Magnesium formability at room temperature is limited due to strong basal texture generated prior processing. Shear band nucleation, observed under conditions in which dynamic strain aging (DSA), have been attributed to be responsible for texture weakening in dilute Mg – Rare Earths (RE) alloys. A low–cost substitution of RE is calcium (Ca) with similar atomic size. It is hypothesized that deformation of Mg – Ca alloys under DSA conditions improves Mg formability. In this study, Mg–Ca and Mg–Ca–Zn alloys were characterized using optical microscopy, XRD, and SEM/EDS. Compression testing was performed between 200 and 300C at different strain rates to observe DSA behavior. This work was performed under DOE Award Number DE-EE0007756.

Manuel, Michele↗
DOE OSTI2021

High Temperature Steam Oxidation of Irradiated FeCrAl in the Severe Accident Test Station

FeCrAl-UO2 test capsules were fabricated at Oak Ridge National Laboratory (ORNL) and irradiated at the Advanced Test Reactor. Following irradiation, samples were sectioned from the irradiated rod and oxidation kinetics were evaluated to access the candidate cladding high-temperature oxidation performance following irradiation. The irradiation was performed at approximately 400°C to a burnup of 10 GWd/MT. The high-temperature oxidation tests were conducted in the ORNL Severe Accident Test Station at 1200 and 1300°C. Weight measurements were taken before and after oxidation testing. Cross-sections of the cladding were metallographically mounted and optical microscopy was performed. Measurements of the oxidation layer before and after high-temperature testing were collected. The results indicate the irradiated FeCrAl C35M alloy provided good thermal stability up to 1200°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗