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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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Manufacture, Characterization, and Fusion Welding of Thermoplastic Composites for Space Applications

The goals of NASA’s Thermoplastic Development for Exploration Applications (TDEA) Project include and assessment of thermoplastic composite joints for space structures by developing and maturing design capabilities, analysis tools and techniques, and manufacturing processes for thermoplastic composites. Through this effort the TDEA project will expand NASA’s in-house thermoplastic composite manufacturing capabilities, develop an understanding of advanced thermoplastic joining techniques relevant to space environments, evaluate the feasibility of reconfigurable composite structures, and advance structural analysis capabilities, including failure prediction of thermoplastic composites including joints. This presentation will provide an overview of the materials and manufacturing effort within the TDEA project. Three semi-crystalline and one amorphous thermoplastic composite system were selected for initial screening and provided an opportunity to gain manufacturing experience across a range of processing temperatures, rheological behavior, and composite properties. Equivalency to the National Center for Advanced Materials Performance (NCAMP) test report for Toray’s TC1225 material was established to provide confidence in in-house manufacturing and material properties were generated through baseline mechanical tests for each material. Material characterization data was generated for a subset of materials to (1) provide data for analysis and model development and (2) generate thermal, rheological and conductivity data across relevant temperatures. A key objective of the TDEA project is evaluation of thermoplastic composite joining processes suitable for in-space operations. This presentation will outline the project’s progress in fusion bonding candidate materials by common welding techniques to assess the quality, reproducibility and strength of the bond, as well as identify limitations to in-space manufacturing. The feasibility of joint disassembly and reassembly will be discussed.

thermoplastic composite↗
DOE OSTI2013

Performance characteristics of Jefferson Lab's new SRF infrastructure

In the past two years, Jefferson Lab has reconfigured and renovated its SRF support infrastructure as part of the Technology and Engineering Development Facility project, TEDF. The most significant changes are in the cleanroom and chemistry facilities. We report the initial characterization data on the new ultra-pure water systems, cleanroom facilities, describe the reconfiguration of existing facilities and also opportunities for flexible growth presented by the new arrangement.

Reece, Charles E.↗
DOE OSTI2015

Baseline Risk Assessment Supporting Closure at Waste Management Area C at the Hanford Site Washington

The Office of River Protection under the U.S. Department of Energy is pursuing closure of the Single-Shell Tank (SST) Waste Management Area (WMA) C under the requirements of the Hanford Federal Facility Agreement and Consent Order (HFFACO). A baseline risk assessment (BRA) of current conditions is based on available characterization data and information collected at WMA C. The baseline risk assessment is being developed as a part of a Resource Conservation and Recovery Act (RCRA) Facility Investigation (RFI)/Corrective Measures Study (CMS) at WMA C that is mandatory under Comprehensive Environmental Response, Compensation, and Liability Act and RCRA corrective action. The RFI/CMS is needed to identify and evaluate the hazardous chemical and radiological contamination in the vadose zone from past releases of waste from WMA C. WMA C will be under Federal ownership and control for the foreseeable future, and managed as an industrial area with restricted access and various institutional controls. The exposure scenarios evaluated under these conditions include Model Toxics Control Act (MTCA) Method C, industrial worker, maintenance and surveillance worker, construction worker, and trespasser scenarios. The BRA evaluates several unrestricted land use scenarios (residential all-pathway, MTCA Method B, and Tribal) to provide additional information for risk management. Analytical results from 13 shallow zone (0 to 15 ft. below ground surface) sampling locations were collected to evaluate human health impacts at WMA C. In addition, soil analytical data were screened against background concentrations and ecological soil screening levels to determine if soil concentrations have the potential to adversely affect ecological receptors. Analytical data from 12 groundwater monitoring wells were evaluated between 2004 and 2013. A screening of groundwater monitoring data against background concentrations and Federal maximum concentration levels was used to determine vadose zone contamination impacts on groundwater. Waste Management Area C is the first of the Hanford tank farms to begin the closure planning process. The current baseline risk assessment will provide valuable information for making corrective actions and closure decisions for WMA C, and will also support the planning for future tank farm soil investigation and baseline risk assessments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Consensus Standardization of Slurry Simulant Development Process to Reduce Design Risk Within the DOE Complex

A critical responsibility of DOE’s Office of Environmental Management is the design, construction, and operation of equipment and facilities to process legacy radioactive waste slurries for safe, long-term disposal. The accomplishment of this mission requires the extensive use of physical and chemical simulants to represent the hazardous radioactive waste. Development and characterization of simulants can often be difficult to compare or utilize as questions exist as to the basis, techniques, or consistency associated with reported parameters. In addition, controversy has existed in the interpretation and application of the characterization data for the wastes that the simulants are intended to represent.

design↗
DOE OSTI2016

FIELD IMPLEMENTATION PLAN FOR A WILLISTON BASIN BRINE EXTRACTION AND STORAGE TEST

The Energy & Environmental Research Center (EERC) successfully completed all technical work of Phase I, including development of a field implementation plan (FIP) for a brine extraction and storage test (BEST) in the North Dakota portion of the Williston Basin. This implementation plan was commissioned by the U.S. Department of Energy (DOE) National Energy Technology Laboratory (NETL) as a proxy for managing formation pressure plumes and measuring/monitoring the movement of differential pressure and CO2 plumes in the subsurface for future saline CO2 storage projects. BEST comprises the demonstration and validation of active reservoir management (ARM) strategies and extracted brine treatment technologies. Two prospective commercial brine injection sites were evaluated for BEST to satisfy DOE’s goals. Ultimately, an active saltwater disposal (SWD) site, Johnsons Corner, was selected because it possesses an ideal combination of key factors making it uniquely suited to host BEST. This site is located in western North Dakota and operated by Nuverra Environmental Solutions (Nuverra), a national leader in brine handling, treatment, and injection. An integrated management approach was used to incorporate local and regional geologic characterization activities with geologic and simulation models, inform a monitoring, verification, and accounting (MVA) plan, and to conduct a risk assessment. This approach was used to design a FIP for an ARM schema and an extracted brine treatment technology test bed facility. The FIP leverages an existing pressure plume generated by two commercial SWD wells. These wells, in conjunction with a new brine extraction well, will be used to conduct the ARM schema. Results of these tests will be quantified based on their impact on the performance of the existing SWD wells and the surrounding reservoir system. Extracted brine will be injected into an underlying deep saline formation through a new injection well. The locations of proposed extraction and injection wells were selected during the Phase I efforts. These wells will be permitted as North Dakota Administrative Code Underground Injection Control Class II wells and will yield additional characterization data which will further refine the FIP in Phase II. An array of surface and downhole monitoring techniques will validate ARM performance against predictive simulation results. Infrastructure will be constructed to manage extracted fluids at the surface and provide brine to a treatment test bed facility. Treatment of extracted brine can provide a means of reducing extracted brine disposal volumes, an alternate source of water, and/or salable products for beneficial use. A test bed facility will be constructed to provide a means of demonstrating these technologies on a wide range of brine concentrations. Screening criteria based on a techno-economic and life cycle assessment were developed to select high-salinity brine treatment technologies for extended duration treatment (30–60 days) in Phase II. A detailed cost assessment determined total implementation costs for BEST of $19,901,065 million (DOE share $15,680,505). These costs are inclusive of all necessary equipment, infrastructure construction, operations and project closeout costs required to implement BEST. An ideal combination of key factors makes the Johnsons Corner site uniquely suited to be the BEST demonstration.

10 SYNTHETIC FUELS↗
DOE OSTI2017

Improvements to Modeling Capabilities of ATF Concepts in the BISON Fuel Performance Code

Research into accident tolerant materials for the fuel and cladding in Light Water Reactors has been at the forefront of the nuclear fuel research community since the events that oc- curred at the Fukushima Daiichi nuclear power plant in March 2011. Accident tolerant materials are defined as those that provide significantly increased response time in the event of an accident while providing similar or improved performance as the conventional UO2/zirconium-based cladding fuel rods dur- ing normal operation [1]. In particular, qualitative attributes for materials with enhanced accident tolerance include im- proved reaction (e.g., oxidation) kinetics with steam, resulting in slower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical prop- erties, fuel-clad interactions, and fission-product behavior [1]. Through its Office of Nuclear Energy the United States De- partment of Energy (U.S. DOE) has accelerated research in this area through the Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC). The goal of the AFC Accident Tolerant Fuel (ATF) program is to guide se- lection of promising concepts for insertion into a commercial reactor as part of a lead test rod or assembly in 2022. Given the aggressive development schedule, it is impossi- ble to perform a comprehensive set of experiments to provide material characterization data. Therefore, the AFC is utilizing computational analysis tools in an effort to understand the proposed accident tolerant materials. The two materials of interest in this work are iron-chromium-aluminum (FeCrAl) cladding and U3Si2. From a fuel performance standpoint, the correlations available to describe the behavior of these two materials is limited. Therefore, as additional experimental data becomes available and correlations are updated, the fuel performance models must be updated to represent the current state of the art understanding of these materials. In this paper we demonstrate the importance of utilizing the latest state-of-the-art models for fuel performance simula- tions by comparing old and new thermal conductivity models for U3Si2 fuel and old and new oxidation models for FeCrAl cladding alloys using the BISON [2, 3] fuel performance code under normal operating conditions. The results indicate that the latest thermal conductivity model for U3Si2 results in fuel centerline temperatures that are on average approximately 3% higher than the previous model. For the oxidation behavior of FeCrAl alloys the new model predicts observable but small ox- ide thickness (~6 µm) compared to the previous model which predicted zero oxide scale formation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Emerging Issues in the Development of Geologic Models for Gas Hydrate Numerical Simulation

The numerical simulation of gas hydrate reservoirs has evolved greatly in recent years. Scientific and industrial drilling programs have acquired detailed characterization data from gas hydrate systems in Alaska, Japan, India, Korea, China, and the Gulf of Mexico. These data provide invaluable calibration points that serve to ground-truth hydrate reservoir modeling efforts.

Boswell, Ray↗
DOE OSTI2022

Impact of Tree Age and Anatomical Fraction on Forest Residue Pyrolysis and Hydrotreating

Utilization of cost-advantaged biomass and waste resources to produce clean, domestic biofuels will be a key factor in decarbonizing transportation in the United States. This study investigates how the distribution of anatomical fractions (stem wood, bark, needles, branches) from pine forest residues of different ages impacts pyrolysis and hydrotreating processes, including product composition, carbon efficiency, economics, and overall sustainability. Commercially sourced 13- and 23-year-old loblolly pine residues were chosen to represent available resources; whole tree thinnings, and tops and branches from merchantable timber, respectively. Detailed characterization of the residue samples, product distributions, bio-oil, char, and gas characterization data are presented for 13 experiments, including the individual fractions, blends, and air-classified samples to reduce ash content. Pyrolysis mass balances averaged 97 +/- 3%. As expected, bio-oil yields were lower for the bark, needles, and blends with high proportions of these (53-56% dry basis), and highest for the debarked stem wood (67-71% dry basis). Lower yields were attributed to the high ash content of the needles, but also to the high lignin content in the bark (>50%), which is hypothesized to have led to increased char formation during pyrolysis. Higher-oil yield was generally correlated with total volatiles, hydrogen, oxygen, glucan, and mannan; somewhat correlated with xylan and acetyl content; and negatively correlated with total carbon, fixed carbon, nitrogen, sulfur, and extractives. Air classification of the residues was effective in reducing ash content (1.5% to 0.64%), extractives (~7% to 2%), and sulfur (0.04% to 0.02%), resulting in a small increase in yield, a significant increase in the GC-detectable proportion of oil (27% to 34%) and a net economic benefit. These experimental results were used to conduct technoeconomic and life cycle analyses, which are also summarized here.

biofuels↗
DOE OSTI2022

Experimental Steps toward a Density Law for Chlorine-Crediting Criticality Models of Aqueous Plutonium solutions

Criticality accident prevention is an essential safety consideration for all operations involving fissionable and fissile materials. Predictive criticality calculations using advanced neutron transport codes, such as Monte Carlo N-Particle Transport (MCNP), are invaluable tools for designing and implementing operational limits. While indispensable, these tools are limited by the quality and accuracy of the inputs that the user provides to define the modeled system. Parameters such as atomic composition, shape, and material density must be accurately defined to obtain a meaningful result. In the case of material property details for a plutonium model, accurate material characterization data is sometimes sparse. Thus limitations in calculation accuracy can be reduced by improving our knowledge of material properties in the targeted fissile systems. One of the biggest remaining challenges to accurately defining fissile systems is the description of aqueous fissile solutions. Even simple properties, such as density, are not well-known for fissile solutions relevant to nuclear energy and security. Described here are initial efforts undertaken to improve criticality calculation inputs for fissile plutonium chloride solutions in water. This effort is focused on experimentally determining accurate solution characteristics for ternary plutonium chloride/hydrochloric acid/water systems, by measurement of water activity and solution density. The effect of inputting experimental densities for these solutions into MCNP criticality calculations is compared to the traditional approach of modeling an idealized (and fictitious) plutonium-water mixture. Expansion of these efforts to a working density law for aqueous plutonium chloride solutions is also discussed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Targeted Health Assessment for Wastes Contained at the Niagara Falls Storage Site to Guide Planning for Remedial Action Alternatives - 13428

The U.S. Army Corps of Engineers (USACE) is evaluating potential remedial alternatives at the 191-acre Niagara Falls Storage Site (NFSS) in Lewiston, New York, under the Formerly Utilized Sites Remedial Action Program (FUSRAP). The Manhattan Engineer District (MED) and Atomic Energy Commission (AEC) brought radioactive wastes to the site during the 1940's and 1950's, and the U.S. Department of Energy (US DOE) consolidated these wastes into a 10-acre interim waste containment structure (IWCS) in the southwest portion of the site during the 1980's. The USACE is evaluating remedial alternatives for radioactive waste contained within the IWCS at the NFSS under the Feasibility Study phase of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process. A preliminary evaluation of the IWCS has been conducted to assess potential airborne releases associated with uncovered wastes, particularly during waste excavation, as well as direct exposures to uncovered wastes. Key technical issues for this assessment include: (1) limitations in waste characterization data; (2) representative receptors and exposure routes; (3) estimates of contaminant emissions at an early stage of the evaluation process; (4) consideration of candidate meteorological data and air dispersion modeling approaches; and (5) estimates of health effects from potential exposures to both radionuclides and chemicals that account for recent updates of exposure and toxicity factors. Results of this preliminary health risk assessment indicate if the wastes were uncovered and someone stayed at the IWCS for a number of days to weeks, substantial doses and serious health effects could be incurred. Current controls prevent such exposures, and the controls that would be applied to protect onsite workers during remedial action at the IWCS would also effectively protect the public nearby. This evaluation provides framing context for the upcoming development and detailed evaluation of remedial alternatives for the IWCS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

EPRI Report: Review of Geostatistical Approaches to Characterization of Subsurface Contamination - 17442

The key metric in the final release of a nuclear facility from radiological controls following plant decommissioning is the potential long-term radiation exposure to future users of the site. This is estimated from modeling of the exposure that could conceivably result from residual contamination remaining on the site in buildings, soil and groundwater after the completion of decontamination, dismantlement and remediation activities. A key input to such modelling is characterization data obtained from surveys and sampling. To date, the regulatory guidance for conducting the final status survey required to show compliance with regulator approved site release limits (i.e., radionuclide concentrations or contamination levels) has been limited to surface contamination. EPRI has completed a project that provides guidance for use of geo-statistics in the development of a site final status survey that will address the impact of dose to a future user of the site from subsurface contamination. (authors)

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

Concluding a steam injection remediation project at a dense non-aqueous phase liquid source zone at the Savannah River site - 15303

Multiple subsurface dense non-aqueous phase liquid (DNAPL) source zones served as the sources for a groundwater plume within the A/M Area of the Savannah River Site. Between 1952 and 1985 over 1,600 metric tons of trichloroethylene, tetrachloroethylene, and 1,1,1-trichloroethane were released to the environment resulting in a 648 hectare plume with the M-Area Settling Basin receiving over half of the solvent discharges. In 1991, DNAPL was found in a water table well and steam injection technology was selected to remediate the source zone. Based on characterization data, the treatment area was 1.21 hectare with depths between 13.7 to 50.3 m. The remediation system included 96 vertical, horizontal, and angled steam injection and soil vapor extraction wells and a thermal monitoring system. Construction of the system began in 2004 and operation began in 2005. Steam was applied to the deep vadose zone first, then progressed to the aquifer, followed by the mid-vadose zone. Multiple steam strategies were utilized to enhance mass removal and were concluded in 2009. Soil vapor extraction continues, with residual temperatures in the deep permeability zones still exceeding 65 deg.C. To date over 204 metric tons of volatile organic compounds (VOCs) have been removed. Post-remediation soil, vapor, temperature, and groundwater data were used to evaluate steam injection effectiveness and demonstrated significant reduction in soil concentration, including removal of DNAPL that was evident in several soil cores prior to treatment. Heating the subsurface increased VOC solubility in groundwater and increased mass removal rates outside the target zone. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Addressing the Legacy Waste Challenge - IAEA Activities in Support of Member States - 16304

The International Atomic Energy Agency (IAEA) promotes the safe and peaceful uses of nuclear energy including the safe management of radioactive waste. Although safe and effective radioactive waste processing technologies and practices are well established and widespread, the issue of legacy waste remains a challenge for many IAEA member states. The primary source of the world-wide legacy waste inventory has its origins in past nuclear weapons programs and associated research and development activities. Other sources include historical waste in old storage and disposal facilities that were either never intended to be a permanent end-state or are no longer considered adequate from a safety perspective. Additionally, many IAEA member states with less advanced nuclear energy programs, often have small inventories of legacy radioactive waste arising from medical, research and industrial activities but lack the processing capabilities to put these wastes into a safe form for storage and disposal. These wastes have the potential to create a safety and environmental hazard and pose a radioactive waste management challenge for many reasons including lack of characterization data and process knowledge, waste package degradation, or the presence of hazardous or mixed waste components. The Waste Technology Section (WTS) within the Department of Nuclear Energy at the IAEA has an extensive program of activities to support the safe retrieval, processing, storage and disposal of radioactive waste including legacy waste. These include publications, co-ordinated research projects, technical co-operation activities, and information exchange forums including workshops, technical meetings and virtual networks. This paper will present an overview of recent IAEA activities and accomplishments aimed as assisting member states with addressing their legacy waste inventories. (authors)

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

Advanced Retrieval Demonstrations of the Hanford Waste End Effector - 19383

This paper describes advanced retrieval demonstrations of the Hanford Waste End Effector (HWEE) at the Washington River Protection Solutions (WRPS) Cold Test Facility conducted by Pacific Northwest National Laboratory (PNNL) in co-operation with the U.S. Department of Energy - Office of River Protection and WRPS. The HWEE is a water-based scarifier originally developed by PNNL for the U.S. Department of Energy tank retrieval systems in the 1990's as the confined sluicing end effector and later used with additional development in the waste retrieval of tanks at Oak Ridge National Laboratory. The HWEE is being assessed for potential use in Hanford tank retrievals because it is expected to use less water during waste retrieval than other water-based methods (e.g., sluicing) previously used in Hanford tanks. This is primarily because the HWEE uses three rotating high-pressure jets for dislodging the waste and a high-velocity conveyance system to retrieve the water and the waste locally at the point of waste dislodging. The retrieval demonstrations build on previous development testing (Phase I) completed in 2017 by addressing key tank waste challenges, including retrieval of a mixed layered simulant and retrieval of uneven simulant profiles like that found near tank walls. The tests involved retrieval of water and kaolin (sludge) simulant to determine pumping rate, dilution factors, and a demonstration of vertical pumping capability. Upgrades to the HWEE for Phase II testing included a more robust inlet screen, internal flush jet modification for lower water usage, and removal of the external flush mole that was not used during Phase I retrieval testing. In addition, a rotational positioning capability was added to the three-dimensional gantry to gain additional movement capability and flexibility needed to implement more complex retrieval mining techniques. Two simulant configuration options using existing simulant materials were defined to represent key Hanford tank waste properties: 1) layered waste (e.g., strong crust over weaker sediment) and 2) topography that is representative of tank waste (uneven surface). Historical information from Hanford waste, including tank history, sampling reports, and images, was investigated to define the simulant characteristics and physical configuration. Using the layered simulant configuration, retrieval tests focused on demonstrating the same operational parameters irrespective of simulant strength. The focus of the topographical retrieval test was to demonstrate the ability to clean next to the tank wall using a tilted HWEE and to demonstrate water confinement using a specific mining strategy. This paper will describe the Phase II test equipment, instrumentation, simulant preparation procedures, and testing methodology. Testing results that will be presented include HWEE water usage versus operating pressure, simulant characterization data, dilution ratios, and retrieval rates. Reaction forces and other factors that would affect the remote operation of the HWEE will also be presented. (authors)

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

A Python Package for Fuel Cycle and Waste Management Analysis - 19545

The future of nuclear power utilization in the United States may involve a transition from the once-through light water reactor fuel cycle to a so-called advanced fuel cycle. An advanced fuel cycle may offer improvements in natural resource utilization, reactor safety, and environmental impact. In addition, adoption of an advanced fuel cycle will result in changes in the amount, composition, and disposal of the nuclear wastes they produce. Characteristics intrinsic to the wastes themselves, such as mass and radioactivity at specific times, were used as waste management metrics in the Department of Energy 2014 Fuel Cycle Evaluation and Screening (FCES) study. Binning of these metrics informed fuel cycle according to relative potential benefit. However, waste properties alone cannot qualify management strategy. Therefore, comparing fuel cycles from the perspective of waste management can provide additional insights. To that end, an object-oriented Python package has been developed to perform fuel cycle waste management calculations. The package, named nwpy (short for Nuclear Waste analysis in Python), mimics the mass flow operations reported in the FCES study and interfaces with the ORIGEN-S code to assess metric data as reported in the FCES. In addition, this Python package determines the loading of waste streams into forms suitable for final disposal in a geological repository and uses such data for repository performance assessment. This work describes the Python package methodology, assumptions, and capabilities. As input data, the package requires information about the fuel cycle such as the mass of fuel discharged, the cooling time between reactor discharge and subsequent operations, and the separations process efficiencies and waste forms it produces. In addition, the isotopic composition of the reactor discharge stream, such as that available from the FCES database, is required. With that, it can perform prescribed mass flow and decay operations throughout the back end of the fuel cycle, the latter carried out by the ORIGEN-S code distributed with the SCALE6.2 software suite. The result is time-dependent data characterizing the waste on a per-package basis. Package loading for waste streams depends on the stream form (i.e. UNF vs. HLW, or glass HLW from aqueous reprocessing vs. ceramic HLW from electrochemical processing) and is based on literature review and engineering assumption. As an example of the application of the code, calculations for the areal footprint of direct-contact repositories for different fuel cycles is shown. Development is ongoing, and the package is expected to grow incorporating capabilities for waste management analyses such as waste transportation and radionuclide transport through the biosphere. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Zero-Valent Iron Permeable Reactive Barrier to Remediate Volatile Organic Compounds in Groundwater - 20209

The US DOE-SRS, the US EPA, and the South Carolina Department of Health and Environmental Control determined it was appropriate to perform a non-time critical removal action at the P-Area Groundwater Operable Unit at the SRS to reduce the mass and downgradient transport of trichloroethylene in the P-Area groundwater plume. Contaminated groundwater discharges to a nearby stream, Steel Creek, within the SRS boundaries, resulting in trichloroethylene concentrations above the maximum contaminant level. Impact to surface water is limited in areal extent and supported by recently collected characterization data. The P-Area Groundwater Operable Unit encompasses the groundwater beneath an industrial area within SRS, P Area, where the P-Reactor once operated. The boundaries of the P-Area Groundwater Operable Unit extend northwest to Steel Creek, northeast toward PAR Pond, and southeast to Meyers Branch. Groundwater in the Upper Three Runs Aquifer of the P-Area Groundwater Operable Unit has been impacted by reactor and facility operations between 1954 and 1991, including tritium and volatile organic compounds. The P-Area surface units contributing to groundwater contamination were remediated as part of the P-Area Operable Unit in 2011. The P-Reactor closure is one of the first of its kind in the DOE Complex and is one of only a few full-sized production reactors in the US to undergo completion of final closure activities. The nature and extent of groundwater contamination was determined using a variety of investigative approaches such as groundwater monitoring wells, direct-push technology, and surface water samples. Groundwater contamination associated with trichloroethylene is primarily exhibited in a narrow plume that extends from the source area at P-Reactor and west to Steel Creek. Maximum contaminant level exceedances in groundwater occur over an area of ∼6.9 hectares for trichloroethylene with concentrations as high as 7.7 milligrams per liter. To the west of the P-Area facility area, the trichloroethylene groundwater plume is controlled by a buried geologic feature, assumed to be an old stream bed, that further narrows the groundwater plume in what has been designated as the 'neck area.' This narrowing of the groundwater plume provides an ideal location for a treatment barrier. The non-time critical removal action alternative chosen is to install a zero-valent iron permeable reactive barrier within the neck area of the trichloroethylene groundwater plume, perpendicular to groundwater flow direction. This technology will provide a treatment barrier that will reduce trichloroethylene groundwater concentrations by 90% and has an anticipated useful life of at least 25 years. A pre-design investigation was performed in the neck area to confirm site lithology, hydrogeology, geochemistry, and extent of trichloroethylene contamination prior to a final design. A treatability study, conducted as part of the pre-design investigation, indicated that the subsurface and groundwater in the PArea Groundwater Operable Unit is compatible with the zero-valent iron and will not lead to excessive buildup from mineralization/precipitation or biofouling. Probabilistic modeling was conducted using field and laboratory data to determine the expected performance of the zero-valent iron permeable reactive barrier. The model simulations indicated that a 3.81-centimeter thick barrier would provide greater than 90% reduction of trichloroethylene groundwater concentration. The final design of the zero-valent iron permeable reactive barrier is a barrier that will extend 80.5 linear meters in a 'zigzag' orientation to best transect the trichloroethylene plume and account for varying groundwater flow. The barrier will be installed from 13.7 meters below ground surface to 41.1 meters below ground surface for 65.8 linear meters and from 13.7 meters below ground surface to 36.6 meters below ground surface for 14.6 linear meters, the base of which is 'keyed' into a low permeability zone. The barrier is designed to a thickness of 10.2 centimeters, which was determined to reduce the trichloroethylene groundwater concentrations by greater than 90% with a safety factor of 2.67. A total of approximately 689 metric tons of zero-valent iron will be injected through 22 injection wells spaced 3.66 meters apart, using guar to suspend the zero-valent iron. Zero-valent iron permeable reactive barrier construction will be monitored through 23 installed resistivity receivers offset 7.32 meters from the zerovalent iron permeable reactive barrier. The zero-valent iron will be energized with a low-voltage 100 Hertz signal during injection and will be monitored using the resistivity receivers to ensure complete coalescence of the zero-valent iron permeable reactive barrier. The zero-valent iron was sized to have a hydraulic conductivity greater than the natural subsurface, thus promoting groundwater flow through the barrier. As contaminated groundwater contacts the zero-valent iron, volatile organic compounds, including trichloroethylene, are immediately degraded to harmless compounds such as ethylene. The performance of the zero-valent iron permeable reactive barrier will be monitored using three upgradient monitoring well clusters, six downgradient monitoring well clusters, and four in-wall monitoring wells. The in-wall monitoring wells will indicate immediate reduction of trichloroethylene mass in the groundwater and allow analyses of the zero-valent iron permeable reactive barrier health. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Improvements to Modeling Capabilities of ATF Concepts in the BISON Fuel Performance Code

Research into accident tolerant materials for the fuel and cladding in Light Water Reactors has been at the forefront of the nuclear fuel research community since the events that occurred at the Fukushima Daiichi nuclear power plant in March 2011. Accident tolerant materials are defined as those that provide significantly increased response time in the event of an accident while providing similar or improved performance as the conventional UO{sub 2}/zirconium-based cladding fuel rods during normal operation. In particular, qualitative attributes for materials with enhanced accident tolerance include improved reaction (e.g., oxidation) kinetics with steam, resulting in lower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical properties, fuel-clad interactions, and fission-product behavior. Through it's Office of Nuclear Energy, the United States Department of Energy (U.S. DOE) has accelerated research in this area through the Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC). The goal of the AFC Accident Tolerant Fuel (ATF) program is to guide selection of promising concepts for insertion into a commercial reactor as part of a lead test rod or assembly in 2022. Given the aggressive development schedule, it is impossible to perform a comprehensive set of experiments to provide material characterization data. Therefore, the AFC is utilizing computational analysis tools in an effort to understand the proposed accident tolerant materials. The two materials of interest in this work are iron-chromium-aluminum (FeCrAl) cladding and U{sub 3}Si{sub 2}. From a fuel performance standpoint, the correlations available to describe the behavior of these two materials is limited. Therefore, as additional experimental data becomes available and correlations are updated, the fuel performance models must be updated as well to assure that the material models reflect the latest experimental findings. In this paper we demonstrate the importance of utilizing the latest state-of-the-art models for fuel performance simulations by comparing the change in the updated thermal conductivity model for U{sub 3}Si{sub 2} fuel and oxidation model for FeCrAl cladding alloys using the BISON fuel performance code under normal operating conditions. The results indicate that the latest thermal conductivity model for U{sub 3}Si{sub 2} results in fuel centerline temperatures that are on average approximately 3% higher than the previous model. For the oxidation behavior of FeCrAl alloys the new model predicts observable but small oxide thickness (6 μm) compared to the previous model which predicted essentially zero oxide scale formation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Hydraulic fracture characterization by integrating multidisciplinary data from the Hydraulic Fracturing Test Site 2 (HFTS-2)

Various technologies have traditionally been used to monitor and describe hydraulic fractures from different perspectives. This work demonstrates the value of data integration for hydraulic fracture characterization when multiple data resources are available. The Hydraulic Fracturing Test Site 2 (HFTS 2) is a hydraulic fracturing research project in the Delaware Basin with multiple surveillance techniques including fiber optics sensing, microseismic, pressure/temperature gauges, etc. We integrated the multidisciplinary data from the HFTS-2 to characterize hydraulic fractures. The integrated data revealed interesting fracture propagation features including layering, vertical propagation affected by pore pressure gradient, and different microseismic activities due to difference in-situ conditions. Furthermore, these findings can be insightful for understanding hydraulic fracture propagation. The comparison among multiple surveillance data also helps us to evaluate the roles of various surveillance technologies and provides us experience to make informative decisions depending on different monitoring objectives.

58 GEOSCIENCES↗