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At least 721 records · Page 40

Evaluation of the Impact of Additional Manganese from the Recycle Collection Tank (RCT) Glycolate Destruction Process on Glass Properties

The Defense Waste Processing Facility (DWPF) is planning to implement glycolic acid as a reductant within the waste processing flowsheet. An assessment of the glycolic acid flowsheet has revealed the potential for thermolytic production of hydrogen in the Concentration, Storage and Transfer Facilities (CSTF) from glycolate entrained in the DWPF recycle stream. To mitigate this potential scenario, a glycolate destruction process utilizing sodium permanganate (NaMnO 4 ) is being developed for use in the DWPF Recycle Collection Tank (RCT). The RCT is fed by the Slurry Mix Evaporator Condensate Tank (SMECT) and the Off-Gas Condensate Tank (OGCT). The SMECT could receive glycolate via a foamover from the Sludge Receipt and Adjustment Tank (SRAT) or the Slurry Mix Evaporator (SME), and the OGCT could receive glycolate via carryover of sludge particles in the purge from the melter during surge conditions. The use of NaMnO 4 additions in the RCT will result in additional manganese (Mn) in the waste stream and needs to be evaluated.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Foaming during Nuclear Waste Melter Feeds Conversion to Glass: Application of Evolved Gas Analysis

During the final stages of batch-to-glass conversion in a waste-glass melter, gases evolving in the cold cap produce primary foam, the formation and collapse of which control the glass production rate via its effect on heat transfer to the reacting batch. We performed quantitative evolved gas analysis (EGA) for several HLW melter feeds with temperatures ranging from 100 °C to 1150 °C, the whole temperature span in a cold cap. EGA results were supplemented with visual observation of batch-to-glass transition using the feed expansion tests. Upon heating, most of the gases—mainly H2O, CO2, NO, NO2, N2 and O2—evolve at temperatures below 700 °C and escape directly to the atmosphere through open porosity. However, as open porosity closes when enough glass-forming melt appears at 720 °C, the residual gas evolution leads to the formation of primary foam. We found that primary foaming is mostly caused by the decomposition of residual carbonates, though oxygen evolution from iron-redox reaction can also play a role. We also show that the gas evolution shifts to a higher temperature when the heating rate increases. The implications for the mathematical modeling of foam layer in the cold cap are presented.

Hujova, Miroslava↗

Crossflow membrane filtration system for operando fouling characterization using transmission x-ray scattering

Membrane-based separations are widely used for wastewater treatment due to their low cost and efficiency. However, membrane fouling, which is the unwanted deposition or attachment of contaminants on membrane surfaces and/or within membrane pores, remains a major challenge as it increases the mass transfer resistance and reduces membrane productivity. Membrane fouling is typically probed by macroscopic performance metrics, such as flux decline, and ex situ characterization. However, this does not capture the membrane and fouling layer evolution under operating conditions, potentially masking important mechanisms and nonequilibrium pathways that impact fouling. Here, we present a remotely controlled crossflow membrane system and a custom membrane cell for operando fouling characterization using transmission small/wide angle x-ray scattering (SAXS/WAXS). This approach allows direct observation of the nanoscale changes occurring at the membrane surface during pressurized water treatment processes, enabling a new way to understand the connections between dynamic fouling behaviors and membrane performance. Nanoparticle fouling of porous membranes during ultrafiltration was investigated using operando SAXS, and mineral scaling of reverse osmosis membranes was investigated using operando WAXS. Furthermore, this system allows for tracking membrane fouling in real time and under realistic conditions, providing fundamental physical insights into how water chemistry and operating conditions affect macroscopic membrane performance. Moreover, this system opens the door for future in situ and operando studies, and it serves as a testbed for evaluating novel materials/processes for membrane-based separations.

36 MATERIALS SCIENCE↗

Hanford Site Composite Analysis Special Analysis: Inventory and Solid Waste Release Modeling for the LLBG Sensitivity Case - UCAQ-22-01 Inventory Discrepancies for 218-E-12B, 218-W-3A, and 218-W-3AE in the Hanford Site Composite Analysis

This environmental calculation file (ECF) documents the methodologies, assumptions, and results of four sensitivity analyses that reevaluate the representativeness of solid-waste radionuclide inventory and release rate from three solid waste sites included in the recently completed Hanford Site Composite Analysis (CA) (DOE-RL-2019-52, Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau (FY 2020), Rev. 1) (hereinafter called the CA Update). Specifically, this ECF reevaluates the representativeness of the base case inventory and radionuclide waste release rates from three solid waste sites (i.e., 218-E-12B, 218-W-3A, and 218-W-3AE) and two radionuclides (i.e., carbon-14 [C-14] and technetium-99 [Tc-99]). These three waste sites and two radionuclides were identified as being the most significant contributors to groundwater contamination and dose in the CA Update for the Inner Area boundary at times periods after the compliance period. This ECF first evaluates the representativeness of the C-14 and Tc-99 inventory and second, radionuclide release rates for the three waste sites1. If the inventory and release rates assumed in the CA Update are determined to be not representative, as hypothesized in the CA Update, then this ECF evaluates the impact of more representative inventories and release rates on the transfer of radionuclides to the vadose zone. The four analyses described in this ECF are as follows: 1) 218-E-12B C-14 Inventory and Release Rate Sensitivity Case – Determine the representativeness of the C-14 inventory and associated waste release rate from the 218-E-12B waste site assumed in the CA Update. If the inventory and associated waste release are more appropriately characterized as being different from the assumptions in the CA Update, then update the predicted C-14 release rate and compare the results to the results presented in the CA Update (DOE/RL-2019-52); 2) 218-W-3A C-14 Inventory and Release Rate Sensitivity Case – Determine the representativeness of the C-14 inventory and associated waste release rate from the 218-W-3A waste site assumed in the CA Update. If the inventory and associated waste release are more appropriately characterized as being different from the assumptions in the CA Update, then update the predicted C-14 release rate and compare the results to the results presented in the CA Update (DOE/RL-2019-52); 3) 218-W-3AE Tc-99 Release Rate Sensitivity Case – Determine the representativeness of the Tc-99 inventory and associated waste release rate from the 218-W-3AE waste site assumed in the CA Update. If the inventory and associated waste release are more appropriately characterized as being different from the assumptions in the CA Update, then update the predicted Tc-99 release rate and compare the results to the results presented in the CA Update (DOE/RL-2019-52); 4) 218-W-3AE Tc-99 Release Footprint Sensitivity Case – Determine the representativeness of the Tc-99 waste area footprint for the 218-W-3AE waste site assumed in the CA Update. If the footprint of the waste is more appropriately characterized as being different from the assumptions in the CA Update, then update the predicted Tc-99 release rate and compare the results to the result presented in the CA Update (DOE/RL-2019-52).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Advanced spacecraft fuel cell systems

The development and characteristics of advanced spacecraft fuel cell systems are discussed. The system is designed to operate on low pressure, propulsion grade hydrogen and oxygen. The specific goals are 10,000 hours of operation with refurbishment, 20 pounds per kilowatt at a sustained power of 7 KW, and 21 KW peaking capability for durations of two hours. The system rejects waste heat to the spacecraft cooling system at power levels up to 7 KW. At higher powers, the system automatically transfers to open cycle operation with overboard steam venting.

Thaller, L. H.↗

Development of a household waste treatment subsystem, volume 1

The domestic waste treatment subsystem was developed to process the daily liquid and non-metallic solid wastes provided by a family of four people. The subsystem was designed to be connected to the sewer line of a household which contained water conservation features. The system consisted of an evaporation technique to separate liquids from solids, an incineration technique for solids reduction, and a catalytic oxidizer for eliminating noxious gases from evaporation and incineration processes. All wastes were passed through a grinder which masticated the solids and deposited them in a settling tank. The liquids were transferred through a cleanable filter into a holding tank. From here the liquids were sprayed into an evaporator and a spray chamber where evaporation occurred. The resulting vapors were processed by catalytic oxidation. Water and latent energy were recovered in a combination evaporator/condenser heat exchanger. The solids were conveyed into an incinerator and reduced to ash while the incineration gases were passed through the catalytic oxidizer along with the processed water vapor.

Gresko, T. M.↗

An Overview of In-Stu Treatability Studies at Marshall Space Flight Center, Huntsville, Alabama

Marshall Space Flight Center (MSFC) is located in Huntsville, Alabama (north-central Alabama), on approximately 1,840 acres near the center of the U.S. Army's Redstone Arsenal (RSA). MSFC is the National Aeronautics and Space Administration's (NASA's) principal propulsion development center. Its scientists, engineers, and support personnel play a major role in the National Space Transportation System by managing space shuttle mission activities, including the microgravity laboratory. In addition, MSFC will be a significant contributor to several of NASA's future programs, including the Reusable Launch Vehicle (X-33), International Space Station, and Advanced X-ray Astrophysics Facility, as well as research on a variety of space science applications. MSFC has been used to develop, test and manufacture space vehicles and components since 1960, when civilian rocketry and missile activities were transferred from RSA to MSFC. In 1994, MSFC was placed on the National Priority List for the management of hazardous waste sites, under the requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). One requirement of the CERCLA program is to evaluate the nature and extent of environmental contamination resulting from identified CERCLA sites, assess the public health and environmental risks associated with the identified contamination, and identify potential remedial actions. A CERCLA remedial investigation (RI) for the groundwater system has identified at least five major plumes of chlorinated volatile organic compounds (CVOCs) in the groundwater beneath the facility. These plumes are believed to be the result of former management practices at 14 main facility locations (termed "source areas") where CVOCs were released to the subsurface. Trichloroethene (TCE) is the predominant CVOC and is common to all the plumes. Perchloroethene (PCE) also exists in two of the plumes. In addition to TCE and PCE, carbon tetrachloride and 1,1,2,2-tetrachloroethane are contained in one of the plumes. The CVOCs are believed to exist as dense non-aqueous phase liquids (DNAPLs) beneath many of the source areas.

McElroy, Bill↗

NASA Utilization of Space Nuclear Systems for Robotic and Human Exploration Missions: Response to EO 13972: Promoting Small Nuclear Reactors for National Defense and Space Exploration

Space Nuclear Systems (SNS) technology development offers a wide range of capabilities to support NASA’s current and future missions. Executive Order (EO) 13972, “Promoting Small Modular Reactors for National Defense and Space Exploration” [1], issued 5 January 2021, directs NASA to define requirements for NASA utilization of nuclear energy systems for human and robotic exploration missions through 2040 and analyze the costs and benefits of such requirements.” Although it is premature to define requirements and cost for future exploration missions that have not yet been formulated, this report describes planned objectives and missions by 2040 that are enabled or enhanced by nuclear systems while taking into account a number of unique considerations for nuclear energy in the space environment. Nuclear energy systems are enabling for space missions and critical capabilities where conventional forms of energy production are impractical or impossible due to mass constraints, mission duration, or distance from the Sun. Space nuclear technologies available or in development for use by 2040 utilize radioisotope decay or nuclear fission and fall into three categories: heat, power, and propulsion. Current applications utilize radioisotope power systems that provide consistent and reliable performance in the sub-kilowatt power range. More advanced SNS can enable new mission objectives where high energy density solutions are critical, or where access to solar solutions is prohibitive. Higher power radioisotope and fission systems are under development within NASA for a wide variety of human exploration and science mission applications. Planned missions designed to use radioisotope systems include Dragonfly, a rotorcraft that will explore the surface of Titan, and Persephone, a mission concept for a Pluto orbiter. Nuclear fission systems have the key advantage of providing significantly higher power, lower mass solutions from tens to even thousands of kilowatts. Fission power is enabling to a sustained human presence on the Moon and developing a robust lunar economy. Fission propulsion is enabling for missions within and beyond cis-lunar space. This report examines NASA-envisioned mission applications and associated performance needs for SNS over the next twenty years leading to 2040 along with the unique technical considerations posed by space nuclear technology development. This includes engineering and operational logistics for ground handling, thermal management, survival of the space environment, operational safety, power requirements, and service longevity. Safety to the public, the NASA work force, and agency assets remains a top priority for NASA and particular attention is given to this aspect in the design, hardware assembly, ground operation, launch, and mission operation of an SNS. NASA relies on the Department of Energy as nuclear authority and its legacy of rigorous safety procedures as standards for ground development, test, transportation, and launch site operation. The principal concern is preventing unintended radiological release to the public or environment. Radioisotope system experience has established processes, including ground operation, transportation, and launch, that are considered directly applicable to emerging fission systems; however, fission systems have unique design needs that impact the safety and performance requirements. High efficiency power conversion from both fission and radioisotope systems requires high operating temperatures necessitating both passive and active thermal management to maintain safe and nominal operating conditions. Effective cooling and waste heat rejection have special considerations for space applications, whether in zero-g or reduced gravity. Fluid and heat transfer within the reactor system is not anticipated to be impacted by reduced or zero-g environments. Cryogenic working fluids and propellant supplies utilized in some space nuclear applications will need low mass, high capacity cryocoolers to meet the long-term storage and near zero-boiloff needs. Integrated, high power density SNS capable of being packaged in a single vehicle is a key consideration for NASA. Due to concerns for complexity and reliability, in space reactor assembly and reactor refueling are not current design considerations. Expanding into a new era for space exploration depends on mass-efficient, high-energy solutions to power deep-space vehicles, operate in harsh environments, and increase mission flexibility. NASA nuclear technology investments are targeting power for surface operations and propulsion for fast-transit, deep-space missions, all with the ability to reliably operate without the need for repair or refueling. NASA’s goals, enabled by nuclear technologies, provide for exciting advances in scientific objectives and human exploration, ushering in a new space age that enables a human presence on bodies beyond our Earth.

nuclear↗

Tank 40 final sludge batch 9 chemical and fissile radionuclide characterization results

A sample of Sludge Batch (SB) 9 was pulled from Tank 40 in order to obtain radionuclide inventory analyses necessary for compliance with the Waste Acceptance Product Specifications (WAPS)i. The SB9 WAPS sample was also analyzed for chemical composition, including noble metals, and fissile constituents, and these results are reported here. These analyses along with the WAPS radionuclide analyses will help define the composition of the sludge in Tank 40 that is fed to the Defense Waste Processing Facility (DWPF) as SB9. At the Savannah River National Laboratory (SRNL), the 3-L Tank 40 SB9 sample was transferred from the shipping container into a 4-L high density polyethylene bottle and solids were allowed to settle. Supernate was then siphoned off and circulated through the shipping container to complete the transfer of the sample. Following thorough mixing of the 3-L sample, a 547 g sub-sample was removed. This sub-sample was then utilized for all subsequent slurry sample preparations. Eight separate aliquots of the slurry were digested, four with HNO3/HCl (aqua regiaii) in sealed Teflon® vessels and four with NaOH/Na2O2 (alkali or peroxide fusioniii) using Zr crucibles. Three Analytical Reference Glass – 1iv (ARG-1) standards were digested along with a blank for each preparation. Each aqua regia digestion and blank was diluted to 1:100 with deionized water and submitted to Analytical Development (AD) for inductively coupled plasma – atomic emission spectroscopy (ICP-AES) analysis, inductively coupled plasma – mass spectrometry (ICP-MS) analysis, atomic absorption spectroscopy (AA) for As and Se, and cold vapor atomic absorption spectroscopy (CV-AA) for Hg. Equivalent dilutions of the alkali fusion digestions and blank were submitted to AD for ICP-AES analysis. Tank 40 SB9 supernate was collected from a mixed slurry sample in the SRNL Shielded Cells and submitted to AD for ICP-AES, ion chromatography (IC), total base/free OH-/other base, total inorganic carbon/total organic carbon (TIC/TOC) analyses. Weighted dilutions of slurry were submitted for IC, TIC/TOC, and total base/free OH-/other base analyses. Activities for U-233, U-235, and Pu-239 were determined from the ICP-MS data for the aqua regia digestions of the SB9 WAPS slurry using the specific activity of each isotope. The Pu-241 value was determined from a Pu-238/-241 method developed by SRNL AD and previously described.v

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Vitrification Testing of HLW with High Phosphate

Projections of the number of high level waste (HLW) canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) (e.g., [1]) are based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier Department of Energy (DOE) melter designs, incorporates a glass bubbler system. The bubblers create active glass pool mixing and thereby improve heat and mass transfer and glass melting rate. The WTP HLW melters each have a glass surface area of 3.75 m 2 and depth of ~1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by optimization of the feed and glass formulations, increasing the melter operating temperature above 1150⁰C, and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Numerical Modelling of Radionuclide Migration in the Context of a Near-Surface LILW Disposal Facility - 18618

The reference solution for the management of low and intermediate level short-lived waste (LILW) in France consists in their emplacement in concrete vaults built at the near-surface. The corresponding disposal facility under operation by the French waste management organization (Andra), is located in North-eastern France and planned to be closed in 2060. The safety case developed by Andra has to show the evidence that, amongst other aspects, the radionuclide transfer through the confinement barriers will occur in such a way that the human and environmental exposure will be as low as reasonably achievable (ALARA) and in accordance with the dose constraints. For this purpose, modelling is used to integrate uncertainties about the characteristics of concrete components, waste and water table and their evolution in time and space. At last, modelling is finally used to investigate the radiological impact of specific scenarios in which the failure of safety barriers occurs. In the frame of its technical review of the safety assessment, the French Institute for Radiological Protection and Nuclear Safety (IRSN) also performs such kind of numerical modelling of radionuclide fate from the waste to the hydrogeological outlets, in order to check whether the assessment performed by Andra takes into account sufficiently conservative assumptions. The present work aims to evaluate the role of components of the vaults on radionuclide transport during the post-closure period over a few thousand years. For this purpose, IRSN carries out numerical calculations with MELODIE software, which simulates water flow and transport of species, in order to simulate radionuclide migration from waste packages emplaced in a concrete vault to the surrounding groundwater. The developed transport model focuses on 3-H radionuclide and integrates Andra's main assumptions (release model from the waste packages, chemical and hydraulic properties of the vault components, shrinkage and cracks in concrete components). An adaptive mesh method, based on an a posteriori error estimator, was used to model the cracks and the consideration of a strong heterogeneity in the hydraulic and transport parameters values in the different components of the model. This method shows an improvement of results of about 40 times, whereas the mesh size was increased by only 2%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Nitrate-To-Ammonia Electroconversion at Neutral pH on Polycrystalline Vanadium Sulfide Derived from Vanadium Disulfide

The electrochemical nitrate reduction reaction (NO3RR) offers a pathway to produce NH3 for fuel and fertilizer from waste NO3-. In this work, a polycrystalline vanadium sulfide (VSx), which is derived from solvothermally grown and annealed VS2, is shown to exhibit excellent NO3RR activity (2.3 +- 0.6 mg.cm-2 geo..h-1 @ -0.92 VRHE) and Faradaic efficiency to NH4+ (69 +- 6% at -0.69 VRHE) in buffered neutral pH electrolyte containing 0.1 M NO3-. A variety of characterization techniques are leveraged to support the VSx assignment, including X-ray photoelectron spectroscopy, near-edge X-ray absorption fine structure spectroscopy, selected area electron diffraction, and X-ray diffraction measurements. The VS2 annealing step reduces the oxide character and generates VSx, which, based on the improved NO3RR activity, results in the creation of active sites for NO3- binding. To help shed light on NO3RR on VSx, VS2 is used as a model system, and a grand-canonical density functional theory (GC-DFT) investigation of VS2 shows strong evidence that S vacancies are active sites for NO3RR, where NO3- outcompetes H+ for adsorption at the S-vacancy sites. Moreover, GC-DFT results highlight a thermodynamically favorable reaction to generate NH4+ in an aqueous electrolyte at relevant cathodic potentials. As an annealed material, VSx may contain undersaturated V sites, which show an electronic structure similar to the theoretically calculated S-vacancy site of VS2, and these sites may contribute to the observed increase in NO3RR activity and selectivity for NH4+ on VSx versus unannealed VS2. Finally, kinetic isotope effect measurements suggest that the kinetic rate-limiting step of the NO3RR on VSx is not proton-coupled, indicating it may be the first electron transfer to adsorbed NO3*.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Impact of Crystalline Silicotitanate Particle Size on Cesium Removal Efficiency

Washington River Protection Solutions is working on a Tank-Side Cesium Removal (TSCR) system to remove Cs from Hanford tank waste supernate. Removal of 137 Cs from tank waste supernate by TSCR results in a low-activity waste (LAW) form that can be fed directly to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW vitrification facility, where the supernate can be vitrified into a glass form. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIV TM R9140-B), was selected for use as the ion exchange media to remove Cs at TSCR. The particle size used in testing is expected to impact the kinetic performance of the ion exchange media. Therefore, two single-column systems were tested with CST Lot 2002009604. One column contained <25 mesh sieve cut fraction and the other column contained <35 mesh sieve cut fraction. Flowrate through both test systems was adjusted to match the CST contact time expected for the full-scale operation, i.e., matched bed volumes per hour (BV/h) at 1.83 BV/h. The feed was processed downflow through the columns until ~1000 BVs of feed were processed. The waste acceptance criteria (WAC) limit for the WTP LAW vitrification facility is <3.18E-5 Ci 137 Cs per mole Na. For a tank waste containing 156 µCi/mL 137 Cs and 5.6 M Na, up to 0.114% of the influent 137 Cs concentration may be delivered to the WTP (based on 137 Cs content in AP-107 tank waste); this required a Cs decontamination factor of 876. The WAC limit was used as one of the relevant process parameters, along with the 50% breakthrough point, Cs loading per gram of CST, and the mass transfer zone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Tank 40 Final SB7b Chemical Characterization Results

A sample of Sludge Batch 7b (SB7b) was taken from Tank 40 in order to obtain radionuclide inventory analyses necessary for compliance with the Waste Acceptance Product Specifications (WAPS). The SB7b WAPS sample was also analyzed for chemical composition including noble metals and fissile constituents, and these results are reported here. These analyses along with the WAPS radionuclide analyses will help define the composition of the sludge in Tank 40 that is currently being fed to the Defense Waste Processing Facility (DWPF) as SB7b. At the Savannah River National Laboratory (SRNL) the 3-L Tank 40 SB7b sample was transferred from the shipping container into a 4-L high density polyethylene bottle and solids were allowed to settle over the weekend. Supernate was then siphoned off and circulated through the shipping container to complete the transfer of the sample. Following thorough mixing of the 3-L sample, a 558 g sub-sample was removed. This sub-sample was then utilized for all subsequent analytical samples. Eight separate aliquots of the slurry were digested, four with HNO 3 /HCl (aqua regia) in sealed Teflon® vessels and four with NaOH/Na 2 O 2 (alkali or peroxide fusion) using Zr crucibles. Two Analytical Reference Glass - 1 (ARG-1) standards were digested along with a blank for each preparation. Each aqua regia digestion and blank was diluted to 1:100 mL with deionized water and submitted to Analytical Development (AD) for inductively coupled plasma - atomic emission spectroscopy (ICP-AES) analysis, inductively coupled plasma - mass spectrometry (ICP-MS) analysis, atomic absorption spectroscopy (AA) for As and Se, and cold vapor atomic absorption spectroscopy (CV-AA) for Hg. Equivalent dilutions of the alkali fusion digestions and blank were submitted to AD for ICP-AES analysis. Tank 40 SB7b supernate was collected from a mixed slurry sample in the SRNL Shielded Cells and submitted to AD for ICP-AES, ion chromatography (IC), total base/free OH - /other base, total inorganic carbon/total organic carbon (TIC/TOC) analyses, and Cs-137 gamma scan. Weighted dilutions of slurry were submitted for IC, TIC/TOC, and total base/free OH - /other base analyses. Activities for U-233, U-235, and Pu-239 were determined from the ICP-MS data for the aqua regia digestions of the Tank 40 WAPS slurry using the specific activity of each isotope. The Pu-241 value was determined from a Pu-238/-241 method developed by SRNL AD and previously described. The following conclusions were drawn from the analytical results reported here: (1) The ratios of the major elements for the SB7b WAPS sample are different from those measured for the SB7a WAPS sample. There is less Al and Mn relative to Fe than the previous sludge batch. (2) The elemental composition of this sample and the analyses conducted here are reasonable and consistent with DWPF batch data measurements in light of DWPF pre-sample concentration and SRAT product heel contributions to the DWPF SRAT receipt analyses. The element ratios for Al/Fe, Ca/Fe, Mn/Fe, and U/Fe agree within 10% between this work and the DWPF Sludge Receipt and Adjustment Tank (SRAT) receipt analyses. (3) Sulfur in the SB7b WAPS sample is 82% soluble, slightly less than results reported for SB3, SB4, and SB6 samples but unlike the 50% insoluble sulfur observed in the SB5 WAPS sample. In addition, 23% of the soluble sulfur is not present as sulfate in SB7b. (4) The average activities of the fissile isotopes of interest in the SB7b WAPS sample are (in μCi/g of total dried solids): 4.22E-02 U-233, 6.12E-04 U-235, 1.08E+01 Pu-239, and 5.09E+01 Pu-241. The full radionuclide composition will be reported in a future document. (5) The fission product noble metal and Ag concentrations appear to have largely peaked in previous DWPF sludge batches, with the exception of Ru, which still shows a slight increase in SB7b.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Effect of Feeding Rate on the Cold Cap Configuration in a Laboratory-Scale Melter - 13362

High-level-waste melter feed is converted into glass in a joule-heated melter, where it forms a floating layer of reacting feed, called the cold cap. After the glass-forming phase becomes connected, evolving gases produce bubbles that form a foam layer under the feed. The bubbles coalesce into cavities, from which most of the gases are released around the edges of the cold cap while gases also escape through small shafts in the reacting feed. The foam layer insulates the cold cap from the heat transferred from the molten glass below. The cold cap behavior was investigated in a laboratory-scale assembly with a fused silica crucible. A high-alumina waste simulant was fed into the crucible and the feed charging rate was varied from 3 to 7 mL min{sup -1}. After a fixed amount of time (35 min), feed charging was stopped and the crucible was removed from the furnace and quenched on a copper block to preserve the structure of the cold cap during cooling. During the rapid quenching, thermal cracking of the glass and cold cap allowed it to be broken up into sections for analysis. The effect of the charging rate on the height, area and volume of the cold cap was determined. The size of the bubbles collected in the foam layer under the feed increased as the cold cap expanded and the relationship between these bubbles and temperature will be determined for input into a mathematical model. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NUHOMS{sup R} EOS (Extended Optimized Storage): Robust, Advanced, and Dual Purpose Dry Storage System - 16472

AREVA TN has completed the design, analysis, and US NRC license submittal for the NUHOMS{sup R} Extended Optimized Storage (EOS) system. NUHOMS{sup R} EOS offers a large capacity and high heat load dry storage and transportation solution for UNF. EOS is not a new system, but an evolution of the already-proven safe and low-risk horizontal NUHOMS{sup R} storage system, which is one of the most widely used UNF dry storage systems in the United States. The EOS system features a 37PTH dry shielded canister (DSC) for 37 PWR UNF assemblies and an 89BTH DSC for the 89 BWR UNF assemblies. The EOS DSCs will be stored inside a horizontal storage module (HSM) that is a robust, fortress-like concrete storage overpack. Designed with the lessons learned from the Fukushima event in Japan in mind, EOS has robust margins for beyond design basis (BDB) events and offers the lowest risk and highest safety margin of any dry storage system design for recovery from BDB events. EOS also offers innovative aging management features with an aging management program (AMP) to assure safe and low dose long-term storage at the utility sites. EOS incorporates ports for AMP inspections and offers the option of using duplex stainless steel (DSS) for the DSC shell in order to negate corrosion issues. EOS is stored horizontally and above ground, making it easy and quick to access for visual inspection, monitoring, and mitigating actions during AMP. NUHOMS{sup R} EOS is optimized for high capacity, high burnup, high enrichments, and very low cooling times for UNF assemblies. It enables the highest heat load per canister (50 kW) and low dose rates. Even with high heat loads, EOS has significant margins to design basis limits. As the future of used fuel disposition will require the need to move fuel off-site to interim or consolidated storage, EOS is the ideal solution as it enables on-site loading and unloading operations that mitigate risk when transferring fuel on-site or transporting fuel off-site. The EOS design can be safely located in a traditional outdoor setting with fortress-like concrete reinforced modules that offer superior resistance to attack or can be safely operated inside of a building enclosure for those utilities desiring a more discreet siting for its used fuel storage systems. EOS has an innovative basket design that is made of high strength low alloy (HSLA) material and is easy to fabricate due to its non-welded design. Its innovative inlet and outlet air vent design, which incorporates dose reduction hardware, allows for optimal ambient airflow for highly efficient heat transfer with minimal radiation dose. The NUHOMS{sup R} EOS storage modules are interlocked in a side-by-side configuration offering superior and robust seismic performance. The license application for NUHOMS{sup R} EOS is currently under US NRC review and this system will be available for deployment in 2017. (authors)

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Closure Report for Corrective Action Unit 104: Area 7 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada

This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 104, Area 7 Yucca Flat Atmospheric Test Sites, and provides documentation supporting the completed corrective actions and confirmation that closure objectives for CAU 104 were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management. CAU 104 consists of the following 15 Corrective Action Sites (CASs), located in Area 7 of the Nevada National Security Site: · CAS 07-23-03, Atmospheric Test Site T-7C · CAS 07-23-04, Atmospheric Test Site T7-1 · CAS 07-23-05, Atmospheric Test Site · CAS 07-23-06, Atmospheric Test Site T7-5a · CAS 07-23-07, Atmospheric Test Site - Dog (T-S) · CAS 07-23-08, Atmospheric Test Site - Baker (T-S) · CAS 07-23-09, Atmospheric Test Site - Charlie (T-S) · CAS 07-23-10, Atmospheric Test Site - Dixie · CAS 07-23-11, Atmospheric Test Site - Dixie · CAS 07-23-12, Atmospheric Test Site - Charlie (Bus) · CAS 07-23-13, Atmospheric Test Site - Baker (Buster) · CAS 07-23-14, Atmospheric Test Site - Ruth · CAS 07-23-15, Atmospheric Test Site T7-4 · CAS 07-23-16, Atmospheric Test Site B7-b · CAS 07-23-17, Atmospheric Test Site - Climax Closure activities began in October 2012 and were completed in April 2013. Activities were conducted according to the Corrective Action Decision Document/Corrective Action Plan for CAU 104. The corrective actions included No Further Action and Clean Closure. Closure activities generated sanitary waste, mixed waste, and recyclable material. Some wastes exceeded land disposal limits and required treatment prior to disposal. Other wastes met land disposal restrictions and were disposed in appropriate onsite landfills. The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) requests the following: · A Notice of Completion from the Nevada Division of Environmental Protection to NNSA/NFO for closure of CAU 104 · The transfer of CAU 104 from Appendix III to Appendix IV, Closed Corrective Action Units, of the FFACO

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Composite Analysis Vadose Zone Transport of Selected Hanford Site Composite Analysis: Vadose Zone Transport of Selected Radionuclides Released from DOE O 435.1 Waste Sites

The main objectives of the composite analysis (CA) compliance case vadose zone (VZ) facet are to simulate transport of radionuclide releases from the surface to the water table, and to provide radionuclide transfer rates to groundwater for inclusion into the Plateau-to-River (P2R) saturated zone (SZ) model (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). The parallel exascale version of the Subsurface Transport Over Multiple Phases (eSTOMP) simulator is used to simulate flow and transport for the CA VZ models. The CA compliance case approach to address the scale and distribution of contaminant sources in the VZ facet is to subdivide the Hanford Site Central Plateau into 26 models that contain contaminant sources and liquid discharges likely to commingle during migration through the VZ to the water table.

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