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The abstracts included in this document are: Responsiveness, Modernization, and Radiography Test at the Centrifuge Test Facility; LANL Hostile Blast Testing Capability Development; Digital Image Correlation Measurement of B61 Center Case Fragmentation Test; and Collaborative Analysis Testing Techniques (CATT) Phase 1.
The subject of this report is the 138-kilovolt (kV) power transmission line on the Nevada National Security Site (NNSS) (see Figure 1). The NNSS, formerly the Nevada Test Site (NTS), was the main continental nuclear test site for the United States between 1951 and 1992. Known as the battleground of the Cold War, the NNSS hosted a total of 928 nuclear tests during that period. It also served as the location of numerous scientific advancements in the fields of rocketry, big-hole drilling, nuclear waste management, and defense. From its construction in the 1960s through the present day, the 138 kV line has provided electrical power to the numerous buildings, facilities, experiments, and tests on the NNSS. The Department of Energy (DOE), in consultation with the Nevada State Historic Preservation Officer (SHPO), has determined that it may contribute to the character and significance of the Mercury Historic District (MHD) on the NNSS, as well as other historic districts that have not been recorded or evaluated yet. The purpose of this report is to fulfill two of the stipulations in the Memorandum of Agreement DEGM58-22NA25553 between the U.S. Department of Energy and the Nevada State Historic Preservation Officer Regarding Installation of a 138-kilovolt Transmission Line from the Mercury Switching Station to the U1a Facility and the Removal of the Historic 138-kilovolt Transmission Line from the Mercury Switching Station to the U1a Facility in Areas 1, 3, 5, 6, and 23 of the Nevada National Security Site (MOA). The MOA was executed as part of the DOE’s obligations under Section 106 of the National Historic Preservation Act (NHPA) to mitigate the adverse effects of two undertakings on historic properties. In particular, this report fulfills MOA Stipulations III.C.1-3, which state that the Department of Energy (DOE) will prepare the following: A historic context for the 138 kV transmission line and identification of the line’s characterdefining elements; High-quality digital images of the 26.45-mile segment of the 138 kV transmission line that is subject to the two undertakings for which the MOA was executed; and Photograph catalogs and map keys showing photograph viewpoints and directions.
The process to develop and qualify nuclear fuels for commercial nuclear application requires fundamental material development, characterization, and design; out-of-pile testing on unirradiated materials; integral fuel rod irradiations, testing, and post irradiation examinations (PIEs); and transient analyses. The historical approach depends on the generation of large empirical datasets and series of integral fuel rod irradiations, and this approach ultimately takes approximately 20 years—or sometimes longer—to acquire data through extensive sequential testing. Thus, the qualification and eventual deployment of new fuel systems constitute a long and drawn-out process. However, recent technological advancements have provided researchers the opportunity to perform out-of-cell, in-situ measurements to assess material performance for the duration of the experiment. One such example of this capability is the use of digital image coordination and thermal imaging to assess Zircaloy cladding performance under a simulated loss-of-coolant accident (LOCA) transient condition. In general, the in-situ measurements provide high-fidelity strain, strain rates, and temperature surface maps. This is critical for the United States nuclear industry, which is actively developing a technical basis to support extending the peak rod average burnup from 62 to ~75 GWd/tU and the deployment of accident-tolerant fuel. However, the Nuclear Regulatory Commission, through their research information letter, outlined a number of technical issues for the industry to address before extending burnup. One topic of specific interest is understanding the cladding balloon and rupture geometry during the LOCA heatup phase. Leveraging these advanced in-situ capabilities, this work intends to use insitu data generated from a simulated LOCA in the Severe Accident Test Station at Oak Ridge National Laboratory to better understand high-temperature creep and its impact on Zircaloy balloon and rupture performance. This work used the BISON fuel performance code to compare BISON’s high-temperature creep model predictions to in-situ data and identify limitations and gaps within the model. Additionally, BISON will subsequently be used to simulate relevant high-burnup LOCA tests. The results will be analyzed and compared to the available post-test data in a manner consistent with the approach outlined in the Nuclear Regulatory Commission research information letter.
Final report on evaluation of NanoSonic high pressure hydrogen hoses under CRADA 399. NanoSonic hose liners, fiber winding, and hose end crimp fittings were analyzed for material properties, design, and burst test performance. The hose burst pressures exceeded 100 MPa under -130C sub ambient conditions. Hose crimp fittings were found to have some defects from assembly that impacted burst pressures. The use of digital image correlation provided unique insight to the crimped fitting influence on the composite material and the winding pattern during burst testing.
Cracking of crystalline silicon photovoltaic cells remains a challenging topic in accurately assessing the long-term reliability and performance of affected modules. Cells can be damaged in every stage throughout the lifetime of a photovoltaic module, ranging from manufacturing, transportation, and installation to operation. Initially, the metallization can be able to bridge the gap of fractured cells and keep individual cell fragments electrically connected. However, photovoltaic modules and cells experience thermo-mechanical stresses during operation from temperature changes and pressure cycles of wind and snow loads. This causes the cell fragments to move, which, in turn, can lead to the wear out of the metallization and, consequently, to power loss or a safety hazard. The rate at which this degradation mechanism proceeds is currently unknown. Hence, in this work, we quantify the cell fragment movement of polycrystalline and monocrystalline mini-modules. By using digital image correlation, we were able to extract the normal crack opening and tangential sliding distances of adjacent cell fragments during heating of the mini-modules. Those distances are essential to develop wear-out models for the metallization and determine the rate of the degradation mechanism. We found that the interconnect technology has a significant impact on the direction and quantity of the cell fragment movements.
This facility operation report covers the refurbishment and initial testing of a vacuum creep frame used to conduct 1300°C tensile and creep tests on Mo- and Nb-based alloys. The report details the refurbishment of the frame and initial operation including 100 h creep experiments compared to similar tests in argon. Unlike the 25 mm dogbone specimens tested in argon, the vacuum load train requires 76 mm dog bone specimens. Testing in vacuum has eliminated concern of O embrittlement in argon and Nb alloy CB103 showed >120% tensile elongation at 1300°C and 1450°C. Also, a window in the vacuum chamber allows the use of digital image correlation (DIC) to measure strain rather than the load train displacement used in the argon frames. Comparison of the displacement and DIC results in the vacuum frame showed significant variation in primary creep deformation, likely due to the machine compliance, but similar secondary creep rates. Recently, ULTIMATE teams have submitted specimens for testing in vacuum. With the relatively low volume of creep testing, the facility performance metric of testing within 4 weeks of receiving specimens has been met.
Over the past 5 years, advances in digital imaging and image processing have enabled everyday computing hardware, such as that found on laptops and even phones, to easily create 3D models of scenes from a series of still images. This field is broadly known as photogrammetry. For safeguards applications, a safeguards inspector could evaluate a 3D model of a facility (such as an enrichment plant or the top of a reactor) constructed from multiple photos instead of reviewing each photo individually. This report compares the detailed capabilities of a particular commercial offering (Quidient) that Oak Ridge National Laboratory (ORNL) has investigated in depth with the general capabilities provided in the industry and recommends areas where the general technology could be pursued for potential safeguards applications.
Understanding stress distributions in solid materials is complicated by the fact that most materials are polycrystalline in nature, with each crystal having an elastically anisotropic reaction to force. This study is to gain a better understanding on how external forces placed on a polycrystal are related to internal reactions within and between the grains. The hypothesis is that the stress distribution in porous to fully dense materials are self-organized based on strong contacts between and within the individual grains created by force chains. Force chains, commonly known in loaded granular materials, and could be the phenomenon that connect micro to macro deformation. Scale bridging measurements conducted through Raman spectroscopy, Atomic Force Microscopy, and Digital Image Correlation will be used to create stress maps, modulus maps, and elastic strain maps across a variety of geological and pharmaceutical polycrystals. When possible, the resulting maps will be compared to current homogenization schemes and a full field models. Finite element modeling will be used to assess whether the patterning seen in the experimental map is a reasonable approximation based on the orientation data of the samples used. A minimum of three publications is projected to be accomplished focusing each on a different method to experimentally test and analyze stress distributions.
This report formally documents the completion of the Advanced Reactor Technologies (ART) Level 3 Milestone (M3TG-24OR0501054), “Continue activities related to Split Disc-DIC - complete analysis of effect of sample thickness on one fine grain graphite,” due May 31, 2024. Details within this report outline the status of activities aimed at elucidating the effect of sample thickness on the tensile strength measurement using the small-disc, split-disc testing standard in support of the US Department of Energy’s ART Graphite R&D Program. The ASTM D8289, Standard Test Method for Tensile Strength Estimate by Disc Compression of Manufactured Graphite, was developed to provide an alternative means for testing tensile strength on smaller specimens, which are compatible with available irradiation capsule volumes, and other compatibility measurements. ASTM D8289 specifies that acceptable specimen diameter can range from 6 to 12.7 mm and that the maximum allowed sample thickness should be half of the diameter. However, information is limited with respect to the effect of thickness on the measured splitting tensile strength. This report documents efforts to understand the effect of sample thickness on splitting tensile strength. The work involved testing Ø12.7 mm samples of fine-grain graphites 2114 and IG-110 of different thicknesses (6.35, 5, 4, and 3 mm). The digital image correlation (DIC) method was applied to the samples, along with the ASTM D8289 Standard, to help interpret the measured results.
The mechanical performance of accident-tolerant fuel (ATF) cladding candidates in light-water reactors (LWRs) must be similar to or better than that of current conventional nuclear fuel claddings to reduce dose to the public and ensure a that the core coolable geometry is maintained during a postulated reactivity-initiated accident (RIA) in light-water reactors (LWRs). During an RIA event, the rapid thermal expansion of nuclear fuel can deform the cladding once the fuel–cladding gap closes. In some cases, the pellet–cladding mechanical interaction (PCMI) can induce mechanical failure in ATF candidates. Thus, the mechanical response of ATF cladding must be investigated by mimicking the conditions of RIA and potentially performing Transient Reactor Test (TREAT) experiments to establish or verify the safety envelope. The work presented in this report investigated the failure behavior of as-received, hydrided, and chromium-coated (Cr-coated) Zircaloy-4 (Zry-4) cladding tube under strain-driven mechanical conditions, mimicking postulated RIA loading conditions. Mechanical testing was performed at 300°C via modified burst test (MBT) equipment with pulse width control previously developed under the Department of Energy’s (DOE’s) Advanced Fuel Campaign (AFC). The mechanical strains were determined using 2D digital image correlation (DIC) techniques. The base Zry-4 acquired by Cameco Inc. was in stress-relieved annealed (SRA) condition. Because of the observed large deformation of the cladding tubes, the failure strain definition was updated for the MBT, which can also be applied to other tube tests where significant bulging (out-of-plane deformation) is present. The failure strain was determined to be affected by the speed of the test or the RIA event. As the RIA-like event duration decreased from 75 to 15 ms, the failure strain decreased 5, 7 and 1% for as-received, hydrided, and Cr-coated specimens, respectively. Fractography on the Cr-coated specimens indicated the presence of two failure mechanisms: (i) crenulation at the outer surface of the coating due to the tensile tractional forces along with coatings grain microstructure and (ii) formation of critical defect at the coating/cladding interface that initiated coating rupture after severe plastic deformation of the Zry- 4 substrate. Based on the MBT results and fractography observations, performing mechanical property testing at high strain rates—in particular on Cr-coated tubes—and semi-integral TREAT experiments are required future efforts to ensure ATF cladding performance during transients. The testing recommended would also inform the development of long-term generalized cladding technologies.
Full-field, multi-measurand diagnostics provide rich validation data necessary to improve the product life cycle time of nuclear safety components. Thermophosphor digital image correlation (TP+DIC) is a method of simultaneously measuring strain and temperature fields using patterned phosphor coatings deposited with aerosol deposition (AD). While TP+DIC produces a functional diagnostic, the coating’s reproducibility and the effect of the patterned features on the inferred temperature remains uncharacterized. This NSR&D project provided the opportunity to study two areas: 1) the tunability and repeatability of aerosol deposition and 2) the robustness of aerosol deposition phosphor on deforming substrates. The first area explores the process-property relationship of parameters elucidating the significance of each on the coating. The second area explores the relationship between the features’ characteristics (namely thickness) and the phosphor emission and inferred temperature. Together, the results will lead to the improved accuracy and functionality of TP+DIC for qualification testing of nuclear safety components.
Boiling water reactors can experience anticipated operational occurrences, where an expected deviation from normal operating conditions is witnessed for short time scales. One example of such transients arises from cladding dryout, where stable boiling becomes unstable, resulting in cladding heat up before re-wetting occurs. To understand the impact of such transients on fuel cladding lifetime and performance, cyclic dryout conditions were imposed upon fresh fuel cladding in the absence of irradiation. Pressurized cladding tube segments were exposed to rapid temperature increases to intermediate temperatures below a cladding rupture threshold and deformation was monitored in situ via 3-D digital image correlation. The impact of cycle frequency on cladding deformation was assessed. A clear understanding of elastic and plastic strain responses were developed for Zr alloy deformation under temperature transients, and these results can be used to support material model development and validate fuel performance codes.
Microstructure drives component behavior. Contemporary crystal plasticity studies compare strain measurements of polycrystal specimens to models. Because each specimen is unique, it is impossible to know which differences are significant. In this project, we invented microstructure clones and explored their use in understanding crystal plasticity. Microstructure clones are specimens with nearly identical microstructures, which allows for multiple destructive tests of a microstructure, insight into how a specimen will deform, variability quantification, and the ability to measure the effects of microstructural changes. Several sets of microstructure clones, pure nickel tensile bars, were tested. The techniques of digital image correlation, crystal plasticity finite element analysis, high resolution electron backscatter diffraction, transmission electron microscopy, and dislocation dynamics were used to understand the structural behavior of these microstructures. This work reshapes the fields of crystal plasticity and structure-property relationships by providing a technique to control for specific variables, quantify microstructural stochasticity, and replicate experiments.
The DOE Nuclear Energy Advanced Reactor Technologies (ART) Program has supported the creation of several databases with information describing the safety performance of fast reactors, components, and fuels. This growing collection of legacy experimental data, operating data, and analysis is available online to registered users. Metallography data is one of the most important types of PIE data being collected, organized and stored in several ART Fast Reactor Databases (https://frdb.ne.anl.gov), including the Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT (the Transient Reactor Test Facility) Experimental Relational Database (TREXR). These databases contain three main sets of metallography data. The first set is the metallography data from Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) irradiated fuel pins measured in the Hot Fuel Examination Facility (HFEF); the second set is the metallography data from EBR-II irradiated fuel pins measured in the Alpha-Gamma Hot Cell Facility (AGHCF); the third set is the metallography data from transient tested fuel pins (the transients tests include the out-of-pile tests and TREAT tests) measured in AGHCF. Since both the second and third sets of data were measured in AGHCF, they are governed by the same specification. The metallography data in the databases are digital images scanned from either positive or negative photos. The quality of the images, including their resolution and contrast, relies on the preserved quality of the pictures and the scanning conditions. To analyze the microstructure of a fuel pin, a series of preparatory steps must be undertaken. These include sectioning, epoxy mounting, mechanical grinding and polishing, and often etching. The resulting samples were then transferred to a secondary hot cell, or glovebox (depending the strength of radiation field) for microscopy examination. The specifications provided herein focus on the metallography examinations. The sample grinding, polishing and etching processes are also discussed. The procedures involving sectioning and epoxy mounting are out of the scope of the current specification; details can be found in the corresponding operation manuals. If more data are collected and added to the ART Fast Reactor Databases, this specification will be updated to accommodate the additional data.
Fresh Zircaloy cladding segments were charged with hydrogen and subjected to simulated Loss of Coolant Accident (LOCA) conditions to assess the impact of hydrogen on the cladding mechanical performance and rupture resistance. The cladding rupture temperature decreased by 10 °C for every 100 wppm H added into the system for two different internal pressurization cases. Post-test analysis showed a loss of ductility with increasing hydrogen content, evident through decreased balloon sizes, smaller rupture openings, and sharper rupture interfaces. Digital image correlation was used to map the cladding strain in situ for two cladding segments. An elevated intermediate temperature uniform strain was visible for the hydride cladding, occurring prior to the decreased ductility during balloon and rupture determined from posttest fractography and microscopy. The thermal stability of Zr alloys as a function of hydrogen concentration and the thermal stability of Zr hydride as a function of temperature are likely responsible for the uniform deformation behavior at intermediate temperatures. The present framework for rapid assessment of hydrogen effects on cladding rupture can be utilized as an initial comparison to high burnup cladding thermomechanical response to a LOCA, where local regions may contain high concentrations of precipitated Zr hydride phase.
This report details an analysis package, Ring Pull Strain Analysis (RPSA), that can be used to present and quantify digital image correlation (DIC) data as it relates to a gaugeless ring pull test. Gaugeless ring pull is a testing technique for mechanical testing of small annular samples, usually cut from a thin-walled tube. DIC data is often necessary for this kind of test because bending moments present on the ring cause a non-uniform strain distribution and localized measurements are necessary. In addition, the annular geometry of a ring lends itself to a polar representation, which is not present with typical DIC analysis methods. RPSA was made to calculate and plot the polar representation of strain from standard pre-processed DIC data of a gaugeless ring pull test. Further analysis can be done on ring pull including a quasi-uniaxial tensile analysis and coating analysis, which are also performed by RPSA. In addition, due to the universality of DIC plotting and ring pull test analysis, RPSA can accommodate a wide variety of tests, though it is tailored for ring pull testing. This report details how RPSA works, including the theory, assumptions, and logic behind the calculations and the structure of the program.
Differently processed Cr coatings on Zircaloy cladding were assessed for coating strain to failure. This report includes i) quantification of Cr coating strain to failure for several coating variants, ii) incorporation of high throughput analysis developed utilizing artificial intelligence, and iii) develop a hoop direction coating strain to failure testing approach for high temperatures that doesn’t require digital image correlation strain measurements. Room temperature strain to failure measurements indicated that heat treatment had an impact on strain to failure of the coating, with increasing heat treatment temperature increasing the average strain to failure of the Cr coating.
This report presents the results of advanced mechanical testing conducted on miniature tensile specimens excised from an irradiated baffle-former bolt, a commercial pressurized water reactor component. The specimens were extracted from the midsection of the bolt shank, where the estimated damage dose reached 23 displacements per atom. The mechanical testing included the following tests: (1) conventional tensile testing at room temperature, augmented by digital image correlation (DIC) to enable noncontact strain measurements, and (2) in situ tensile testing within a scanning electron microscopy (SEM) instrument equipped with energy-dispersive x-ray spectroscopy (EDS) and electron backscatter diffraction (EBSD) detectors to evaluate active deformation and fracture mechanisms. Additionally, SEM fractographic analysis revealed alterations in the fracture mechanism from a predominantly ductile fracture in nonirradiated steel to a mixed fracture mode (still dominating ductile and minor cleavage spots) in irradiated specimens derived from the baffle bolts. The findings indicate a complex strain localization behavior for in-service irradiated steel specimens. Beyond conventional necking at the macro scale and defect-free channel formation at the micro scale, DIC results identified the presence of deformation bands approximately 100 μm in width. These bands consist of chains of grains exhibiting elevated local strain levels and may be considered an intermediate or mesoscale level of strain localization. These bands become discernible near the yield stress as localized hot spots, areas of elevated strain, persisting throughout most of the experiment. The formation of mesoscale deformation bands as a strain localization mechanism may exacerbate the defect-free channel formation in irradiated materials, further promoting irradiation-assisted stress corrosion crack initiation.