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At least 289 records · Page 16

FY23 Status of Quality Assurance Plan for Out-of-Pile Test Data

The DOE Advanced Reactor Technology program has supported recovery and preservation of legacy metallic fuel data collected as part of the US fast reactor program, recognizing it as essential to development and licensing activities for advanced fast reactors. Databases were established as organized collections of experimental records and data generated from in-pile experiments at EBRII, FFTF, and TREAT as well as related out-of-pile examinations of irradiated fuels. The Out-of-Pile Transient Database (OPTD), includes records of over 150 out-of-pile furnace transient tests on metallic fuels conducted at Argonne’s Alpha-Gamma Hot Cell Facility to evaluate their transient performance and characterize fuel/cladding interaction. The database is accessible to registered users from US universities, laboratories, and nuclear industry. Because the data in OPTD has not been formally qualified, its applicability and ability to support licensing activities is limited. This report outlines progress and plans to quality assure data in OPTD, maximizing its impact for model validation and verification as well as qualification of fuels for safe and effective use in advanced reactor designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

FY22 Progress Report on BISON Metallic Fuel Model Development and V&V Using EBR-II Legacy Data

In this report, the activities and achievements made by Argonne National Laboratory for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) BISON code metallic fuel validation and verification project in FY2022 are summarized. The FIPD-BISON integration powered metallic fuel low-burnup swelling framework has been enhanced to cover both radial and axial swelling analyses. Axial-dependent as-fabricated fuel radius profiles have been used to improve the accuracy in evaluating radial swelling strain. The simulations of the IFR experiment X447, which was focused on FCCI/CCCI induced cladding degradation and failure, have been converted into a BISON metallic fuel assessment case with detailed documentation, which is also the first FIPD-BISON integration powered assessment case in BISON. A FIPD-BISON integration repository has been established to be an optional submodule of BISON to support FIPD-BISON integration powered assessment cases. Preliminary trials to simulate the in-cell out-of-pile transient experiment have been made after implementing some essential models such as cladding plasticity models and liquid phase penetration models. The initial results are promising and help identify gaps that need to be filled in FY2023 and beyond, to eventually achieve OPTD-BISON integration powered assessment cases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Cluster Dynamics Simulations of Intra-Granular Fission Gas Bubble Size and Pressure Evolution in UO 2

Fission gases such as xenon (Xe) play a critical role in determining the behavior and response of nuclear fuel. Given that Xe has little solubility in UO 2 , it accumulates and forms bubbles, which significantly impact fuel performance. Intra- and inter-granular bubble nucleation and growth can lead to fuel swelling, and once bubbles interconnect at grain boundaries, fission gas can be released into the plenum. At low temperatures, limited uranium vacancy mobility can restrict swelling, therefore causing the bubbles to become highly pressurized. Consequently, this can induce micro-cracking, promote fission gas release (increasing the likelihood of cladding failure), and even lead to fuel pulverization under accident conditions such as a loss of coolant accident. As bubble evolution is strongly influenced by local temperature and fission rate, markedly different behavior occurs across the radial profile of the fuel pellet. Capturing the mechanisms that underpin bubble evolution is therefore important to predict these behaviors in the fuel. Previous models describing important mechanisms informed by lower length scale simulations have been developed under the NEAMS program. These can describe the evolution of a single bubble type (i.e., single value for radius and pressure) at each position in the pellet, for instance using the Centipede cluster dynamic code. However, in reality, a full distribution in bubble sizes and pressures exists within the microstructure at a given position in the pellet. To address this the cluster dynamics code Xolotl, which can predict Xe and vacancy phase space (i.e., bubble distributions) for intra-granular bubbles, has been used before. Prior work benchmarked the Xolotl code against the Centipede cluster dynamics code to ensure compatibility and to verify that mobile defect properties are adequately transferred between the two codes, along with some physics improvements. In this work, we go further by introducing a physics-based set of improvements that will allow us to accurately predict bubble size distributions and internal bubble pressures under representative UO 2 irradiation conditions. The improvements include (i) coupling bubble-defect reaction energies to a virial equation of state (EOS), (ii) including a bubble surface tension contribution, (iii) incorporating radiation-induced re-solution of Xe and vacancies, (iv) enabling pressure-driven dislocation loop punching through an effective emission of interstitial clusters informed by interstitial loop energetics, (v) accounting for radiation induced athermal diffusion of Xe, and (vi) implementing a Booth-type grain boundary sink representation for all mobile defects and defect clusters. After these modifications, we observe good agreement of Xolotl fission gas bubble size and concentration predictions with legacy experimental measurements. Additionally, it allows the distribution of Xe bubble pressures and radius to also be predicted and compared to data produced through the Advanced Fuels Campaign (AFC) program. Here, we have done this by running simulations under conditions similar to the AFC post-irradiation examination (PIE) samples irradiated at North Anna 2 light water reactor (LWR). Our results shows excellent agreement with these experimental measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Space Weathering Investigations Enabled by NASA's Virtual Heliophysical Observatories

Structural and chemical impact of the heliospheric space environment on exposed planetary surfaces and interplanetary dust grains may be generally defined as space weathering . In the inner solar system, from the asteroid belt inwards towards the Sun, the surface regolith structures of airless bodies are primarily determined by cumulative meteoritic impacts over billions of years, but the molecular composition to meters in depth can be substantially modified by irradiation effects. Plasma ions at eV to keV energies may both erode uppermost surfaces by sputtering, and implant or locally produce exogenic material, e.g. He-3 and H2O, while more energetic ions drive molecular change through electronic ionization. Galactic cosmic ray ions and more energetic solar ions can impact chemistry to meters in depth. High energy cosmic ray interactions produce showers of secondary particles and energetic photons that present hazards for robotic and human exploration missions but also enable detection of potentially useable resources such as water ice, oxygen, and many other elements. Surface sputtering also makes ejected elemental and molecular species accessible for in-situ compositional analysis by spacecraft with ion and neutral mass spectrometers. Modeling of relative impacts for these various space weathering processes requires knowledge of the incident species-resolved ion flux spectra at plasma to cosmic ray energies and as integrated over varying time scales. Although the main drivers for investigations of these processes come from NASA's planetary science and human exploration programs, the NASA heliophysics program provides the requisite data measurement and modeling resources to enable specification of the field & plasma and energetic particle irradiation environments for application to space weather and surface weathering investigations. The Virtual Heliospheric Observatory (VHO), Virtual Energetic Particle Observatory (VEPO), Lunar Solar Origins Exploration (LunaSOX), and Space Physics Data Facility (SPDF) services now provide a wide range of inner heliospheric spacecraft data that can be applied to space weathering of potential exploration destinations including the Moon, asteroids, and the moons of Mars, as well to radiation hazard assessment for the spacecraft and human explorers. For example, the new VEPO service for time-averaging of multi-source ion flux spectra enables the specification of composite flux spectra from a variety of ongoing and legacy missions for applications to surface interaction modeling. Apollo to Artemis data resources of LunaSOX enable specific space weathering investigations for the Moon, while VHO more generally covers the space field and plasma environments of the inner and outer solar system from the sunward-most perihelia of the twin Helios spacecraft to the ongoing heliosheath passages of the twin Voyagers. Composite multi-source spectra from VEPO can also be applied to the continuing compilation of accumulated 1-AU fluence spectra, mostly contributed by solar wind plasma and energetic particle events, for determination of time-averaged particle compositional and kinetic energy output from the Sun and for modeling of long-term irradiation impacts on planetary surfaces.

Cooper, John F.↗

Benchmark for Fuel Shuffling and Depletion for Pebble-Bed Reactors

Pebble bed reactors have specific operational characteristics when their fuel-cycle and fueling operations are considered. They are specifically distinguished by other type of nuclear reactor designs by their online fuel recycling scheme, where the fuel elements that have not yet reached discharge burnup can be reloaded and recycled continuously during normal operation. The fuel in a pebble bed reactor is not stationary and stochastically moves through the core once or several times during its lifetime, which allows them to operate without requiring a large excess reactivity hold for the burnup. However, this characteristic of pebble bed reactors introduces challenges in simulation, as each pebble can take many different trajectories through the core, its composition depends on the details of the irradiation history that is unique to its aggregated path through the core. For predicting the safety performance characteristics, such as source term, maximum fuel temperatures and fuel failure rates, etc., it is important to accurately incorporate the movement of pebbles through the core during their lifetime in a multi-physics simulation together with other phenomena. The equilibrium core analysis for pebble bed reactors are performed with multi-physics tools including fuel depletion in a multi pass reload coupled to the fuel movement. Currently, there are only a few legacy multi-physics simulation tools that can implement the pebble flow characteristics and perform equilibrium core analysis for pebble bed reactors. However, there are development efforts on-going under Department of Energy's Nuclear Energy Advanced Modelling and Simulation program and also in private industry for including these capabilities into their modelling and simulation tools. Any new development in the modelling and simulation tools needs to be validated by using tools such as experiments, analytical solutions or code-to-code benchmarks. In this work, a code-to-code benchmark for the equilibrium core analysis capability of pebble bed reactors was developed. Multiple cases were identified to capture different fuel cycle strategies that can be used in PBRs. The results of each case are presented in terms of overall equilibrium core characteristics: the discharge burnup; spatial burnup distribution; spatial isotopic distributions; axial and radial neutron flux distributions and power history of fuel elements per pass through the core for both a prototypical pebble bed High Temperature Gas-cooled Reactor and a prototypical pebble bed Fluoride-salt cooled High temperature Reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Hanford Double Shell Waste Tank Corrosion Studies (Final Report FY2019)

During fiscal year (FY) 2019, the Savannah River National Laboratory (SRNL) focused on four primary tasks for Hanford Double Shell Tanks (DSTs) corrosion studies. The first task, New Limits, was a continuation from the electrochemical work started in FY16, to expand electrochemical testing to elevated temperatures and elevated hydroxide concentrations and evaluate how these factors will influence pitting corrosion. Using the pitting factor (PF) equation, these new conditions were validated to provide a conservative estimate for the susceptibility of pitting corrosion on legacy carbon steel. The second task, Secondary Liner Corrosion, focused on Vapor Space Corrosion (VSC) and immersion testing studies using two commercially available vapor corrosion inhibitors (VCIs). For this FY, the VCI strategy was applied mid-experiment to determine performance on weathered coupons. The third task, Long-Term Open Circuit Potential (OCP) Drift, was performed using simplified simulated chemistries to evaluate change in pitting corrosion risk due to evolution of corrosion potential for mill-scale and freshly polished 600-grit surfaces. Finally, the fourth task, Microbiologically-Influenced Corrosion (MIC) studies, was focused on the study of leak detection pit (LDP) water sent from the Hanford tank farm for determination of bacteria that can be conducive to corrosion using commercially available kits: BART TM and MICkit® 5. In addition, this FY, a secondary task was added which was a continuation of the study of the Quartz-enhanced Photoacoustic Spectroscopy (QEPAS) system. Work was started but was ultimately not pursued. A summary of each task performed is presented below.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Early-stage uranium-hydrogen corrosion kinetics and mechanism

Research into the early-stage uranium-hydrogen degradation mechanism lacks unified interpretation and understanding due to inadequate temporal and spatial characterization resolution required to unify reaction observations. In this study, white-light interferometry was adapted to characterize the uranium-hydrogen reaction at a temperature of 50 °C and hydrogen pressure of 13.8 kPa. The uranium surface remained unchanged for 62 min during the hydride induction period before the first hydride blister appeared. This blister was monitored from induction through critical spallation and UH₃ powder release during rapid surface hydriding. The critical spallation dimensions measured for height, diameter, area, and perimeter of the first blister were 1.79 µm, 17.58 µm, 187.7 µm 2 and 59.79 µm. Hydriding growth kinetics were assessed by tracking the first hydride site over the reaction time. Analysis shows that the spall-front velocity of the hydride was changing at a rate of 0.91 µm/min which compares favorably with legacy Sievert’s experiments in literature. Total percentage of the surface area hydrided after 242 min was 42.7%, and post-characterization of the UH 3 powders shows both α-UH 3 and β-UH 3 . This work highlights the strength of characterizing the early-stage uranium-hydrogen reaction using white-light interferometry, but more importantly, unifying the understanding and mechanism of the uranium-hydrogen reaction kinetics.

Materials science↗

A Global End-Member Approach to Derive aCDOM(440) from Near-Surface Optical Measurements

This study establishes an optical inversion scheme for deriving the absorption coefficient of colored (or chromophoric, depending on the literature) dissolved organic material (CDOM) at the 440 nm wavelength, which can be applied to global water masses with near-equal efficacy. The approach uses a ratio of diffuse attenuation coefficient spectral end members, i.e., a short and long wavelength pair. The global perspective is established by sampling "extremely" clear water plus a generalized extent in turbidity and optical properties that each span three decades of dynamic range. A unique data set was collected in oceanic, coastal, and inland waters (as shallow as 0.6 m) from the North Pacific Ocean, the Arctic Ocean, Hawaii, Japan, Puerto Rico, and the east and west coasts of the United States. The data were partitioned using subjective categorizations to define a validation quality subset of conservative water masses, i.e., the inflow and outflow of properties constrain the range in the gradient of a constituent, plus 15 subcategories of water masses that were not evolving conservatively. The dependence on subcategories was confirmed with an objective methodology based on cluster analysis techniques. The latter defined five distinct classes with validation quality data present in all classes, but which also decreased in percent composition as a function of increasing class number and optical complexity. Four different algorithms based on different validation quality end members were validated with accuracies of 1.–6.2 %, wherein pairs with the largest spectral span were most accurate. Although algorithm accuracy decreased with the inclusion of more subcategories containing non-conservative water masses, changes to the algorithm fit were small when a preponderance of subcategories were included. The high accuracy for all end-member algorithms was the result of data acquisition and data processing improvements, e.g., increased vertical sampling resolution to less than 1mm and a boundary constraint to mitigate wave focusing effects, respectively. An independent evaluation with a historical database confirmed the consistency of the algorithmic approach and its application to quality assurance, e.g., to flag data outside expected ranges, identify suspect spectra, and objectively determine the in-water extrapolation interval by converging agreement for all applicable end-member algorithms. The legacy data exhibit degraded performance (as 44 % uncertainty) due to a lack of high-quality near-surface observations, especially for clear waters wherein wave-focusing effects are problematic. The novel optical approach allows the in situ estimation of an in-water constituent in keeping with the accuracy obtained in the laboratory.

Stanford B Hooker↗

Legacy Metallic Fuel U(Pu)Zr Data Qualification

The Metallic Fuels Irradiation and Physics Database (FIPD) is an organized collection of legacy metallic fuel U(Pu)Zr data, measurements, and reactor conditions from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The database provides a wealth of information and is easily accessed and utilized by the U.S. nuclear industry and the Nuclear Regulatory Commission (NRC). The database contains three categories of data. The first category covers fuel pin fabrication specifications, including fuel slug diameter and length, cladding diameter, smeared density, etc. The second consists of operational conditions, which include axial distributions of power, temperature, fluence, burnup, isotopic density, etc. The third category contains post-irradiation examination (PIE) results, largely collected at facilities now part of the Idaho National Laboratory (INL), formerly part of Argonne National Laboratory -West (ANL-W) and at the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory, comprising fission gas release and chemistry analysis, profilometry measurements, neutron radiography data, etc. All these data have been organized and preserved in FIPD.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A real-time search for Type Ia Supernovae with late-time interactions with circumstellar material in ZTF data

While it is generally accepted that Type Ia supernovae (SNe Ia) are the terminal explosions of white dwarfs (WDs), the nature of their progenitor systems and the mechanisms that lead up to these explosions remain widely debated. In rare cases, the SN ejecta interact with circumstellar material (CSM) that had previously been ejected from the progenitor system. The longer the delay between the creation of the CSM and the SN explosion, the greater the distance between the SN explosion site and the CSM and the later the onset of the interaction. The unknown distance between the CSM and SN explosion site makes it impossible to predict when the interaction will start. If the time between the SN explosion and the onset of the CSM interaction is of the order of several months to years, the SN has generally faded and it is no longer actively followed up on. This makes it even more difficult to detect the interaction while it is happening. In this work, we report on a real-time monitoring programme running between 13 November 2023 and 9 July 2024. It monitored 6914 SNe Ia for signs of late-time rebrightening using the Zwicky Transient Facility (ZTF). Flagged candidates were rapidly followed up on with photometry and spectroscopy to confirm the late-time excess and its position. We report the discovery of a ∼50 day rebrightening event in SN 2020qxz around 1200 rest-frame days after the peak of its light curve. SN 2020qxz exhibited signs of an early CSM interaction, but had faded from view over two years before its reappearance. Initial follow-up spectroscopy revealed the presence of four emission lines, while later follow-up spectroscopy showed that these had faded shortly after the end of the ZTF-detected rebrightening event. Our best match for these emission lines are H β (blueshifted by ∼5900 km s −1 ) and Ca II λ8542 , N I λ8567 , and K I λλ8763, 8767 (all blueshifted by 5100 km s −1 ; although we note that the line identifications are uncertain). This shows that catching and following up on late-time interactions as they occur can offer new clues on the nature of the progenitor systems that produce these SNe by putting constraints on the possible type of donor star. The only way to do this systematically is to use large sky surveys such as ZTF and the upcoming Vera C. Rubin Observatory’s Legacy Survey of Space and Time (LSST) to monitor a large sample of objects for the rare events that reappear long after the object has faded from view.

circumstellar matter↗

Multimodal, microspectroscopic speciation of legacy phosphorus in two US mid-Atlantic agricultural soils

To understand phosphorus (P) mobility in agricultural soils and its potential environmental risk, it is essential to directly measure solid phase P speciation. Often, bulk P K-edge X-ray absorption near edge structure (XANES) spectroscopy followed by linear combination fitting (LCF) is utilized to determine the solid P phases in soil. However, this method may limit results to only a few major phases. Additionally, XANES spectra for different P species may have very similar features, leading to an over- or underestimate of their contribution to LCF. Here, an improved P speciation by pairing multimodal microbeam-X-ray fluorescence (µ-XRF) mapping coupled with µ-XANES (microbeam-X-ray absorption near edge structure) analysis to directly speciate major and minor P phases on the micron scale is provided. We combined maps of both tender (P, sulfur, aluminum, and silicon) and hard energy (calcium, iron [Fe], and manganese) elements to evaluate the elemental co-locations with P. To better account for uncertainty assigning XANES peaks to individual compounds, a more quantitative fingerprinting by “spectral feature analysis” was completed. With this analysis, an R-factor is reported for the fit. These results were compared to traditional LCF. Pre-edge fitting results revealed the presence of a two-component pre-edge feature for phosphate adsorbed to ferrihydrite. Additionally, phytate co-precipitated with ferrihydrite (Phytate-Fe-Cop) had a pre-edge feature, indicating direct association with Fe. Lastly, a unique P species associated with manganese oxide was identified in the soil via multimodal mapping and µ-XANES. These results allow for better prediction of P dissolution and mobility.

36 MATERIALS SCIENCE↗

Improved Benchmarking of Cohesive Elements in Abaqus Standard for Predicting Disbond and Delamination in Composite Structures

Traditional approaches for aircraft certification require the assumption of an initial flaw condition, either represented as barely visible impact damage (BVID) or through inclusion of a Teflon insert to serve as surrogate damage. Based on the initial composite damage state, the structure must be shown to demonstrate structural durability and damage tolerance (DaDT) according to the following criteria: a. Damage displays no detrimental growth under cyclic loading b. The structure is able to sustain design limit load (DLL) Currently, the only available manner for validating structural performance is through test. Since damage can occur over a wide variety of areas within a structure, this approach has proven to be increasingly expensive and time consuming for composite airframes and acreage structure within the design-test-certification building block. A further complicating factor is the requirement to accurately capture the most critical damage morphologies as a starting condition. To understand the severity of the damage, it is either required to experimentally determine the most critical areas at tremendous expense or rely on legacy data of similar structural testing, which limits design space expansion. A preferred solution is to use advanced analysis to provide improved understanding of load margins for critical locations based on a wide variety of potential starting damage conditions. The standard industry approach for DaDT certification adheres to the use of the traditional virtual crack closure technique (VCCT) method. VCCT is generally a preferred method because it conforms to the current certification principles of damage from a known flaw, and when used correctly, can be effective at predicting delamination propagation under static and cyclic loading. The VCCT method requires the inclusion of an initial flaw in the finite element (FE) model requiring a-priori knowledge of the flaw location. This in turn requires a plethora of analysis cases to be examined to cover a reasonable span of potential damage states. Additionally, the VCCT approach requires node-to-node connectivity rendering it incompatible with the best practices and approaches for using continuum damage mechanics (CDM) based progressive damage and failure analysis (PDFA) tools within a typical FE solver. Alternatives to VCCT have emerged in the form of cohesive elements which utilize the cohesive zone model (CZM). Unlike VCCT which models linear elastic fracture mechanics, cohesive elements couples continuum and fracture based responses through the use of bilinear traction separation laws. These laws are defined based on a penalty stiffness, a cohesive strength, and a strain energy release rate. The approach can be mesh regularized with native cohesive elements within many FE solvers such as Abaqus and LS-DYNA. In Phase I of the NASA Advanced Composites Consortium (ACC) post-buckled stiffened panel with BVID, Strength and Life [1], the performance of cohesive elements were benchmarked in comparison to VCCT and LEFM solutions and showed good agreement using Abaqus explicit [2]. To realize savings on current and future programs, it is still necessary to close technical gaps related to the use of cohesive elements with Abaqus Standard. Within a program environment, standard finite element analysis is the preferred analytical capability for quasi-static loading as it eliminates uncertainty due to oscillatory behavior commonly seen with explicit analysis. This oscillatory behavior creates difficulties in writing margins of safety based on the analysis. The use of negative tangent stiffness material models complicates convergence which typically requires the use of numerical controls such as viscous damping to overcome. To date, there has not been a comprehensive study on how to establish best practices for cohesive element convergence for predictive capability within the Abaqus implicit solver. In pursuit of these goals, under the NASA ACC program, several numerical benchmark problems were proposed including pure mode I (double cantilevered beam – DCB), pure mode II (end notch flexure – ENF), and symmetric/unsymmetric evolving mixed mode (single leg bend – SLB). This paper focuses on the use of cohesive elements to model the delamination through the use of CZM. Specifically, finite element models for the DCB, ENF, symmetric SLB, and unsymmetric SLB, are developed and various solution controls for convergence are studied to develop a best practice. Once the best practice has been developed, the predictive capability of the objective CZM model is used to analyze the hat pull-off strength of a standard hat stiffened configuration under various loading conditions.

Abaqus↗

Using Robotic Manipulators for Radioactive Waste Inspection - 20090

The global nuclear industry has a growing volume of nuclear waste which needs to be scanned, sorted according to its activity and material type, then processed into the correct waste packages for long term storage and disposal. It is vital that there is a detailed understanding of the waste inventory stored in long term waste containers, as knowledge of their contents could predict or prevent any adverse effects in storage. The numerous 'scan and sort' tables which are currently used at many different facilities around the world to sort waste into their correct containers are human operated and require very slow gamma scanning procedures combined with educated guesswork to manually sort the waste. This often leads to excessive conservatisms, with placement of lower activity wastes in higher activity containers, which in turn costs significantly more to store. In the United Kingdom it costs UK Pounds 46 k per cubic meter to store intermediate level waste compared to just UK Pound 2.9 k per cubic meter to store low level waste according to a 2008 Department of Energy and Climate Change report in the UK. A proposed solution to this problem, is the use of a robotic manipulator to automatically inspect the 'scan and sort table' in order to produce an accurate 3D model of the table's waste contents and attach an overlaid radiation map. The radiation map contains spectrometry data and can in consequence be used to distinguish and locate specific radioisotopes. The 3D model should be as accurate as possible in order to allow for a second robot arm with an attached gripper to grasp the objects and place them into their designated long-term storage container. Various scanning procedures are explored in this study including basic raster scanning, adaptive raster scanning and point sampling. The optimal solution will in practice be defined by the required application and activity level of the wastes being inspected. The results presented in this study indicate that it is possible to produce a centimeter accurate 3D model of a mixed assortment of components on a nuclear waste 'scan and sort' table. In addition, it was shown that the waste objects emitting radiation could be accurately identified and located, with an overlaid radiation map. This study is applicable across the nuclear waste management sector. Many of the ideas and concepts developed in this study are applicable in other decommissioning settings for example, dismantling of legacy gloveboxes or routine inspection of nuclear waste packages in storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Impact of Ethylenediaminetetraacetic Acid (EDTA) on the Sorption of Nd(III), Th{sup (IV)}, and U(VI) onto Dolomite in WIPP-Relevant Brines, GWB and ERDA-6 - 20356

Storage of legacy, transuranic waste at the Waste Isolation Pilot Plant (WIPP) requires accurate risk models for performance assessments. These models must predict the long-term fate and transport of contaminants in the WIPP and surrounding environment. Laboratory support is necessary for the development of an accurate model through better understanding of contaminant behavior in the WIPP. The WIPP is a deep geologic repository that is characterized by high ionic strength due to evaporite deposits in the Salado formation (porewaters up to 7.4 M). High concentrations of sodium chloride (NaCl), calcium chloride (CaCl{sub 2}), and magnesium chloride (MgCl{sub 2}) are particularly of concern due to their potential impact on solubility and ultimately transport of contaminants of concern. Additionally, ligands capable of forming strong complexes with metals are found in significant concentrations in the WIPP environment due to their disposal from waste processing and formation during degradation of repository components (e.g. cellulose degradation by calcium hydroxide in cement). Consequently, there is a need to investigate the fate and transport of contaminants in the WIPP environment with a focus on the effects of high ionic strength and strongly coordinating ligands. Contaminants of high concern include the actinides (U(VI), Pu(IV), and Am(III)). Laboratory support of the WIPP modeling efforts was provided by this research. A series of batch experiments were utilized to study the impact of ionic strength, contaminant concentration, and the presence of ethylenediaminetetraacetic acid (EDTA) on the sorption of Nd(III), Th(IV), and U(VI) onto dolomite, a carbonate mineral found within the Culebra formation. The Culebra formation is the most transmissive layer above the WIPP, making transport of contaminants the most likely route in previous risk assessments. Two brines specifically designed to imitate conditions within the WIPP environment were used in these experiments. The U.S. Energy Research and Development Administration Well 6 (ERDA-6, 95%) brine simulates a low Mg environment, while the generic weep brine (GWB, 95%) simulates the WIPP environment with a high Mg concentration. Mg is added to the repository in the form of MgO as an engineered barrier and is naturally present in some formations. It is expected to scavenge carbonate from solutions decreasing formation of the highly soluble and competitive uranium-carbonate complexes. EDTA is used during manipulations of nuclear materials due to its ability to form strong, stable complexes with the actinides and may be found in high concentrations (up to 0.08 mM) within the WIPP as waste canisters degrade. Further, it may provide additional avenues for unwanted transport of contaminants. Nd(III) and Th(IV) were used as stable analogues of Am(III) and Pu(IV), respectively, to represent the most common oxidation states of actinides in the WIPP environment in addition to U(VI). The potential for colloid formation was also investigated with low and high concentrations of contaminants (10 and 1000 μg/L) using three size-dependent particle separation steps. The results of this research provide a better understanding of potential behavior of actinides in the presence of EDTA and within WIPP-relevant brines. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

MASCOT: an ESO–ARO legacy survey of molecular gas in nearby SDSS-MaNGA galaxies - I. First data release, and global and resolved relations between H 2 and stellar content

We present the first data release of the Mapping Nearby Galaxies at Apache Point Observatory (MaNGA)–Arizona Radio Observatory (ARO) Survey of CO Targets (MASCOT), a European Southern Observatory public spectroscopic survey conducted at the ARO. We measure the CO(1–0) line emission in a sample of 187 nearby galaxies selected from the MaNGA survey that has obtained integral field unit (IFU) spectroscopy for a sample of ~10 000 galaxies at low redshift. The main goal of MASCOT is to probe the molecular gas content of star-forming galaxies with stellar masses >10 9.5 M ⊙ and with associated MaNGA IFU observations and well-constrained quantities like stellar masses, star formation rates, and metallicities. In this paper, we present the first results of the MASCOT survey, providing integrated CO(1–0) measurements that cover several effective radii of the galaxy and present CO luminosities, CO kinematics, and estimated H 2 gas masses. We observe that the decline of the galaxy star formation rate with respect to the star formation main sequence increases with the decrease of molecular gas and with a reduced star formation efficiency, in agreement with results of other integrated studies. Relating the molecular gas mass fractions with the slope of the stellar age gradients inferred from the MaNGA observations, we find that galaxies with lower molecular gas mass fractions tend to show older stellar populations close to the galactic centre, while the opposite is true for galaxies with higher molecular gas mass fractions, providing tentative evidence for inside-out quenching.

79 ASTRONOMY AND ASTROPHYSICS↗

Closure Report for the Project Shoal Area: Subsurface Corrective Action Unit 447 Shoal,Nevada, Site

The U.S. Department of Energy (DOE) Office of Legacy Management (LM) has prepared this Closure Report for subsurface Corrective Action Unit (CAU) 447 at the Shoal, Nevada, Site, which was formerly called the Project Shoal Area. The Shoal site was the location of an underground nuclear test in 1963 that resulted in residual radionuclide contamination at the detonation depth of 1211 feet (ft) (DOE 2015a). The environmental restoration process and Corrective Action strategy are conducted in accordance with the Federal Facility Agreement and Consent Order (FFACO) (FFACO 1996, as amended) and all applicable Nevada Division of Environmental Protection (NDEP) statues and regulations. Responsibility for the environmental site restoration was transferred from the DOE, National Nuclear Security Administration, Nevada Field Office, to LM on October 1, 2006. The Shoal site is south of U.S. Highway 50, approximately 30 miles southeast of Fallon, in Churchill County, Nevada (Figure 1). The site was selected for underground nuclear testing in 1961, as part of the Vela Uniform Program. Vela was a research and development program directed toward locating, detecting, and identifying underground and high-altitude nuclear detonations. The Vela Uniform program was specific to underground detonations (AEC 1964). The objective of Project Shoal was to detonate a nuclear device underground in an active seismic area to improve the United States’ ability to detect, identify, and locate underground nuclear detonations. The underground nuclear test at Shoal was a joint effort between the U.S. Department of Defense (DOD) and the U.S. Atomic Energy Commission (AEC), which was the predecessor agency to DOE. Figure 1 is a map showing the site location.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A Spectacular Viscoelastic Model Calibration for 828/D230/Alox Generated from Legacy Sandia Data

The nonlinear viscoelastic Spectacular model is calibrated to the thermo-mechanical behavior of 828/D230/Alox with an alox volume fraction of 20 %. Legacy experimental data from Sandia’s polymer properties database (PPD) is used to calibrate the model. Based on known densities of the epoxy 828/D230 and the alox filler, the alox volume fractions listed on the PPD were likely reported incorrectly. The alox volume fractions are recalculated here. Using the recalculated alox volume fractions, the PPD contains experimental data for 828/D230/Alox with alox volume fractions of 16 %, 24 %, and 33 %, so the thermo-mechanical behavior at 20 % alox volume fraction is estimated by interpolating between the bounding cases of of 16 % and 24 %. Because the Spectacularmodel can be fairly challenging to calibrate, the calibration procedure is described in detail. Several of the calibration steps involve inverse parameter identification, where an experiment is simulated and parameters are iteratively updated until the model response matches the experimental data. As the PPD does not fully describe all experimental procedures, the experimental simulations use assumed thermal and mechanical loading rates that are typical for the viscoelastic characterization of epoxies. Spectacular uses four independent relaxation functions related to volumetric (ƒ 1 ), shear (ƒ 2 ), thermal strain (ƒ 3 ), and thermal relaxations (ƒ 4 ). The previous SPEC model form, also known as the universal_polymer model, uses two independent relaxation functions related to volumetric and thermal relaxation (ƒ ν = ƒ 1 = ƒ 3 = ƒ 4 ) and shear relaxation (ƒ s = ƒ 2 ). The two constitutive choices are briefly evaluated here, where it is found that the four relaxation function approach of Spectacular was better suited for fitting the coefficient of thermal expansion during both heating and cooling.

36 MATERIALS SCIENCE↗

Radiation Hardened Foam Cold Test Plan – Phase-I: Foam Adhesion, Contamination Fixation, Moisture Stresses, Pipe Cutting, and Thermal Profile Testing

This document outlines the Phase-I test objectives and implementation plan for a down-selective foam fixative technology intended to facilitate activities in support of the Savannah River Site (SRS) F/H labs deactivation and decommissioning (D&D) efforts. It is a collaborative effort between Savannah River National Laboratory (SRNL), Florida International University (FIU), and the SRS F/H labs team intended to test and evaluate the potential of an intumescent, fire-retardant foam in mitigating the release of contamination during dismantling operations on radioactively contaminated piping in legacy facilities. The cold test plan addresses specific requirements highlighted by site and safety personnel and will be executed in FIU’s Outdoor Test and Evaluation Facility using a mock-up that replicates the operational conditions at the proposed hot test location at F/H labs. Results from the cold test plan will inform the hot test at F/H labs, which will use the foam fixative to confine and/or isolate residual contamination within a 3-dimensional void space of Hastelloy C-22 piping designated for removal from the site and transported to a proper disposal facility. Phase-I testing will address eight test objectives: (1) evaluation of the adhesion and bonding properties of foam fixative in piping, (2) evaluation of the adhesion of the foam fixative in piping under varying moisture conditions, (3) determination of the heat profile of the foam fixative during curing, (4) determination of the relationship between pipe diameter and foam fixative quantity, (5) determination of the internal pipe pressure after foam deployment and curing, (6) development of a leak test standard operating procedure to test for the effectiveness of the foam plug, (7) initiation of a literature review to determine if using a hot tap is a viable method to deliver foam into piping, and (8) initiation of a 10-foot mock up test that will be used in Phase-II cold testing. The cold test will be conducted at FIU and all testing activities will comply with SRNL Conduct of R&D Protocols (SRNL-IM-2020-00019).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗