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At least 811 records · Page 45

FY19 Annual Report Advanced Light-Duty SI Engine Fuels Research

This project furthers the science-base needed by industry stakeholder to co-evolve the next generations of highly efficient DISI engines and new gasoline-type fuels. Here, the research emphasis is on lean operation, which can provide high efficiency, using fuels that also support traditional non-dilute stoichiometric operation for peak load and power. Lean operation induces challenges with ignition stability, slow flame propagation and low combustion efficiency. Therefore, techniques that can overcome these challenges are studied. Specifically, fuel stratification is used to ensure ignition and completeness of combustion, but this technique has soot- and NO x - emissions challenges. For ultra-lean well-mixed operation, turbulent deflagration can be combined with controlled end-gas autoignition to render mixed-mode combustion for sufficiently fast heat release. However, such mixed-mode combustion requires appropriate autoignition reactivity, motivating fuel studies of autoignition under lean conditions.

02 PETROLEUM↗

Analysis of Tank 38H (HTF-38-18-78, -79) and Tank 43H (HTF-43-18-80, -81, -83) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. Compared with recent samples, the surface and sub-surface samples from Tank 38H show much lower concentrations of most species analyzed except for silicon. The Tank 38H samples show stratification in the tank with the concentration of species in the surface sample being much lower than in the sub-surface sample and both samples being much lower than recent results. The Tank 43H surface and two sub-surface samples generally showed consistent results across the three samples and compare well with recent analyses. The lowest sub-surface sample from Tank 43H contained slightly higher concentrations of uranium and plutonium likely due to the presence of a small amount of dark sludge solids in the sample.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Predictive LES Modeling and Validation of High-Pressure Turbulent Flames and Flashback in Hydrogen-enriched Gas Turbines

The overall objective of this project is to make advances in large-eddy simulation modeling of swirl-flame flashback at elevated pressure, where the reactants can be premixed or stratified, and composed of methane enriched with hydrogen. The modeling effort was conducted with a companion experimental validation effort that focused on high-speed measurements of flashback in a model swirl-flame combustor. The modeling effort focused on the development of manifold-based combustion model for flashback simulations. The main complexity in this configuration comes from the flames’ propagation through a stratified fuel-air mixture along with heat loss near the walls. For this purpose, a non-adiabatic flamelet model was developed and tested using canonical flow configurations. Further, extensive validation in premixed configurations is used to establish the model’s accuracy in capturing heat loss and flame propagation. Finally, the model was tested using the UT flame flashback experiment, and was shown to accurately capture the flame flashback process. The end products from this work are a) a full validated flameout-based approach for modeling flame flashback; b) a complete implementation of the model in the open source OpenFOAM framework, which has been made available to both industry and academia; c) an extensive suite of validation studies performed using the OpenFOAM solver. In addition, new techniques for flame modeling, subfilter modeling of scalars, and simulation of low Mach number flows were developed and tested. In the experimental validation effort, flashback was studied in a premixed and stratified swirl flame by using high-speed imaging diagnostics. Fuels tested included CH 4 enriched with H 2 , with H 2 enrichment up to 84% by volume. These flashback experiments are conducted at pressures ranging from 1 to 5 atm. A swirler-based fuel-injection system was used to create the fuel-air stratification in the radial direction. For the high-pressure measurements, an optically accessible elevated pressure chamber was used. The spatial distribution of the equivalence ratio under non-reacting conditions was characterized using planar laser-induced fluorescence with acetone as the fuel tracer. It was observed that fuel-air pockets were distributed across the mixing tube width, although in an average sense, the fuel-air mixture was radially stratified. The global behavior of upstream flame propagation was investigated for different levels of hydrogen-enrichment. For stratified hydrogen-rich flashback, the propagation path of the flame changed from the inner wall to outer wall, which was induced by the faster chemistry of stoichiometric mixtures that are frequently present near the outer wall. This behavior of hydrogen-rich flashback persists even at elevated pressures up to 5 atm, although the propagation of the flame occurs as a wide flame tongue as opposed to the acute-tipped flame structures present in the atmospheric cases. The end product of the experimental effort is greatly improved knowledge of flashback physics and a unique database that can be used for model validation.

08 HYDROGEN↗

Shock Compressed Forsterite Overturns Prior "Mixed Phase" Theories

To provide a better understanding of the interior structure and dynamics of the Earth’s mantle, it is imperative to know the equation of state and phase diagrams of MgO-MgSiO 3 thermodynamic systems. A more detailed understanding of partial melting and chemical stratification of the mantle is provided from extensive research focused on Mg 2 SiO 4 at high pressures over a wide range of thermodynamic conditions. However, there is no stable compound of Mg 2 SiO 4 in the Earth’s lower mantle. Thus, in order to get an equilibrium state of homogeneous forsterite, it must be decomposed into at least two compounds, periclase and bridgmanite. In-situ powder x-ray diffraction observations of forsterite crystal structure samples shocked to 44(3) and 73(5) GPa, done at the Dynamic Compression Sector (DCS) of the U.S. Department of Energy’s Advanced Photon Source (APS) show chemical segregation of forsterite does not occur on dynamic timescales, but the nonequilibrium states persist to 73(5) GPa, where the shock temperature is predicted to be ~2500 K.

58 GEOSCIENCES↗

Analysis of Tank 30H (HTF-30-19-33, -34), Tank 32H (HTF-32-19-35, -36), and Tank 37H (HTF-37-19-53, -54) Samples for Support of the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator

SRNL analyzed samples from Tank 30H, Tank 32H, and Tank 37H to support the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator system. The six samples from Tanks 30H, 32H, and 37H all contain highly concentrated salt solutions with sodium concentrations ranging from 13.8 M to 16.9 M. Hydroxide is the dominant anion in all of the samples followed by nitrate and nitrite. The preponderance of the analytical determinations (density, anions, cations) on all three tanks show slightly higher concentrations of solute in the sub-surface sample than in the surface sample, indicating some stratification within the tanks. With the exception of the Tank 30H VDS, all of the samples are somewhat more concentrated compared with the compositions of the previous samples. The sum of the major cations versus the sum of the major anions show a difference of <5% for all of the samples indicating good data quality for the non-radioactive analytes in the table. The silicon concentrations measured in the VDS and surface samples from Tank 30H, Tank 32H, and Tank 37H were all below detection.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A Deterministic and Probabilistic Framework for Forecasting of Time-Series Damage States and Associated End-of-lIfe of a Pressurizer Water Reactor Surge Line under Design-Basis and Grid-Load-Following Loading Conditions

Argonne National Laboratory in support of the Department of Energy’s Light Water Reactor Sustainability (LWRS) program. The present semi-annual report covers the period between May 2018 and September 2018. In this report, we present the work performed to further improve the capability for structural integrity prediction. Part of this work focused on validating a computational fluid dynamics model for modeling thermal stratification. We also present finite element (FE) model based thermal-mechanical stress analysis of a pressurized water reactor (PWR) surge line under grid-load-following loading condition. In addition, we present work related to end-of-life prediction of a PWR surge line based on ASME code and NUREG-6909 deterministic approaches under design-basis and grid-load-following conditions. Our research in this report period also majorly involved with the development of a strategy for probabilistic fatigue life estimation. First, we applied a Weibull probabilistic modeling approach based on end-of-life fatigue data (traditional S~N data under in-air and PWR water conditions) to estimate the probabilistic life for a given strain amplitude. Second, we developed a data analytics based Markov-Chain-Monte-Carlo (MCMC) model for probabilistic modeling of time-series fatigue damage states and the associated probabilistic end-of-life of a component. Unlike the Weibull type approach, the MCMC type may not require a large number of fatigue tests for estimating the end-of-life for given loading and environmental conditions. We demonstrated the use of this framework through symmetric fatigue tests (stress- or strain-ontrolled tests either under in-air or PWR-water conditions) obtained from our earlier experiments. The MCMC model can use base time-series data either obtained from a model or fatigue experiment. Since experimental data are typically more accurate than model-calculated data, we used experiment-based time-series data to demonstrate the MCMC framework. However, when performing a complex experiment is not possible (due to time and costs involved), a model-based approach can be used to generate the time-series data. Similar model-based data are presented in our earlier report (e.g. for cyclic hardening damage states of a PWR surge line under constant amplitude and design-basis loading [1]). However, for the MCMC work in this report, we used experiment-based time-series damage states obtained from fatigue experiments under design-basis and grid-load-following loading conditions. The experimental loading inputs were based on thermal-mechanical FE model results of a typical PWR surge line. Then, we used the experimental time-series damage state data for forecasting the probabilistic time-series damage states and end-of-life of the PWR surge line. Based on this MCMC modeling strategy, and with the assumed loading, environment and thermal-mechanical boundary conditions, we have estimated that a 316SS-pure base metal PWR surge line would survive at least 159 years with zero failure probability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Analysis of Tank 30H (HTF-30-19-33, -34), Tank 32H (HTF-32-19-35, -36), and Tank 37H (HTF-37- 19-53, -54) Samples for Support of the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator

SRNL analyzed samples from Tank 30H, Tank 32H, and Tank 37H to support the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator system. The six samples from Tanks 30H, 32H, and 37H all contain highly concentrated salt solutions with sodium concentrations ranging from 13.8 M to 16.9 M. Hydroxide is the dominant anion in all of the samples followed by nitrate and nitrite. The preponderance of the analytical determinations (density, anions, cations) on all three tanks show slightly higher concentrations of solute in the sub-surface sample than in the surface sample, indicating some stratification within the tanks. With the exception of the Tank 30H VDS, all of the samples are somewhat more concentrated compared with the compositions of the previous samples. The sum of the major cations versus the sum of the major anions show a difference of <5% for all of the samples indicating good data quality for the non-radioactive analytes in the table. The silicon concentrations measured in the VDS and surface samples from Tank 30H, Tank 32H, and Tank 37H were all below detection.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Tank 22 Settled Sludge Core Sample Characterization and Testing

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) Engineering (SRR-E) to provide sample characterization and analyses of Tank 22 settled sludge core sample and the supernatant liquids (supernate) in support of Tank 22 re-baseline data, inhalation dose potential/hydrogen generation and suspension/removal of settled sludge from the bottom of Tank 22. The approximately 9.5 inches of intact Tank 22 core sample from SRR-E, after extrusion in the SRNL Shielded Cells, showed three visually different stratification layers or regions of solids in the extruded core material. Part of the initial physical characterization of these three segments of the Tank 22 core sample included density, weight percent total solids, particle size distribution, X-ray diffractions and scanning electron microscope of each segment material. Given the dissimilarity of the three Tank 22 segment materials after the physical characterization results and the fact that there was no clear line of demarcation to indicate where each segment began and ended across the entire length of the core sample, SRNL in conjunction with the SRR-E customer decided to blend all three segments of the material to form a composite Tank 22 core sample for further analysis and characterization without taking into consideration the proportional weight or mass contribution of each segment. This compositing approach assumes that the errors introduced in any characterization of the Tank 22 core sample will be comparable to the errors when the three segments are separated and combined proportionally by weight to form the composite Tank 22 sample. This compositing approach maximized the quantity of composited Tank 22 material needed for radionuclide and other physical characterizations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Tank 38H (HTF-38-20-7, -8) and tank 43H (HTF-43-20-9, -10) samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The results indicate the concentrations of most species in the Tank 38H surface sample decreased significantly from the previous surface sample. The Tank 38H sub-surface sample shows only small changes in concentration for most major species in the solution from the previous sample. The large differences in the concentrations of major components between the Tank 38H surface and sub-surface samples indicate significant stratification of solution species within the tank. The Tank 38H sub-surface sample contained a small amount of sludge solids (<1 wt%).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Tank 38H (HTF-38-20-103, -104) and Tank 43H (HTF-43-20-105, -106) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The results indicate the concentrations of most species in the Tank 38H surface sample decreased from the previous surface sample. The Tank 38H sub-surface sample shows only minor changes in concentration for soluble species in the solution (e.g., Na, Al, Cs-137) but a large drop in concentrations for species typically associated with sludge solids (e.g., U, Pu, Fe, Si) likely because of a decrease in sludge solids from the previous sample. The large differences in the concentrations of major components between the Tank 38H surface and sub-surface samples indicate significant stratification of solution species within the tank.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Assessment of Model Hub Height Wind Speed Performance Using DOE Lidar Buoy Data

As the offshore wind industry continues to gain momentum both in the U.S. and globally, reanalysis products remain essential tools for the lifecycle of a wind farm. From project siting and production estimates to determining construction and maintenance windows, reanalysis products play an important role. These regional and global long-term models also serve as inputs to higher resolution models. With such wide-ranging impacts on the offshore wind industry, the need for validation of reanalysis products is strong, particularly at typical turbine hub heights. Pacific Northwest National Laboratory (PNNL) operates two AXYS WindSentinelTM buoys for the U.S. Department of Energy (DOE) in order to collect meteorological and oceanographic data, including hub height wind speed, in areas of interest for offshore wind development. This work uses DOE buoy observations off the coasts of New Jersey and Virginia to evaluate the performance of Modern Era Retrospective Analysis for Research and Applications-2 (MERRA-2), the North American Regional Reanalysis (NARR), the European Center for Medium-Range Weather Forecasts (ECMWF) Reanalysis 5th generation (ERA5), and the National Oceanic and Atmospheric Administration (NOAA) assimilation system Rapid Refresh (RAP). Biases and degrees of correspondence are determined for each reanalysis product in order to provide insights on the performance and uncertainty for long-term wind resource characterization. With the baseline metrics determined, this work proceeds with an investigation into the sources of large deviations between the modeled and observed hub height wind speeds during the East Coast lidar buoy deployments. Consistent sources of reanalysis model error include stably stratified flow conditions, high wind shear, flow parallel to the coastline, and tropical and winter storms. For the near shore New Jersey location, offshore winds (i.e. blowing from land to water) tend to be associated with model overestimation of observed hub height wind speed, while winds parallel to the coastline are correlated with model underestimation of wind speed. Similarly, near the Virginia buoy, during southerly winds (parallel to the coastline) models underestimate the observed wind speed due to coastal upwelling. Reanalysis model wind speeds underestimate the hub height winds with increasing stable atmospheric stratification. Reanalysis model biases are seen when the observed wind speeds approach or exceed typical turbine cut-out speeds (around 20 m s-1). Large model errors during winter and tropical storm events, high wind speed events, and ramp events are often the result of wind speed timing and magnitude offsets. Overall, ERA5 provides the most successful representation of observed offshore hub height wind speeds at the U.S. East Coast buoy locations and is therefore best suited as an input boundary condition for model case studies. ERA5 performs well for a variety of atmospheric phenomena, including storms and sea breezes, however RAP, the next most successful reanalysis model, performs best in capturing the frequency of ramp events.

17 WIND ENERGY↗

Characterization of Tank 9H Dissolution Batches in Support of Tank Closure Cesium Removal (TCCR) 1A Batch 1 Preparations

Savannah River Remediation (SRR) is currently preparing the first batch of material to be processed through the Tank Closure Cesium Removal (TCCR) 1A system. The feed for TCCR 1A will consist of dissolved saltcake from Tank 9H. Two batches of salt (Batch 1A and Batch 1B) have been dissolved in Tank 9H and subsequently transferred to Tank 10H to prepare Batch 1 for TCCR 1A. Savannah River National Laboratory (SRNL) received samples from each batch of dissolved salt prior to transfer for characterization. SRNL received both a surface and a variable depth sample from Batches 1A and 1B. In both cases no solids were observed in the surface sample, but were observed in the depth sample. For Batch 1A the variable depth sample was only slightly cloudy, while for Batch 1B the variable depth sample contained a significant amount (10.14 wt%) of solids. The solids were determined to be primarily aluminum containing phases, with only a small fraction (0.22 wt%) being sludge solids. In general, the samples from Batch 1A were more concentrated salt solutions than Batch 1B, with sodium concentrations of 8.53 and 8.57 M for the surface and filtered depth samples in Batch 1A, respectively. The sodium concentrations in Batch 1B samples ranged from 4.27 M for the surface sample to 7.57 M for the depth sample filtrate, indicating some stratification within the tank. The 137 Cs activity as well as the total Cs concentration in the filtered Batch 1A depth sample were approximately double the activity and concentration measured in the filtrate from the Batch 1B depth sample. The total Cs concentration in the Batch 1A depth sample filtrate was 22.4 mg/L, while for the Batch 1B depth sample filtrate the total Cs concentration was calculated to be 12.0 mg/L. These Cs concentrations are significantly higher than was measured in Batches 1-3 from Tank 10H dissolved saltcake which was previously processed through the original TCCR unit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Tank 38H (HTF-38-21-95, -96) and Tank 43H (HTF-43-21-97, -98) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The results indicate the concentrations of most species in the Tank 38H surface sample increased from the previous surface sample. The Tank 38H sub-surface sample shows only minor changes in concentration for soluble species in the solution (e.g., Na, Al, Cs-137) relative to the previous sample, but a small decrease in concentrations for species typically associated with sludge solids (e.g., U, Pu, Fe, Si) likely because of a decrease in sludge solids from the previous sample. The large differences in the concentrations of major components between the Tank 38H surface and sub-surface samples indicate significant stratification of solution species within the tank.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Tank 38H (HTF-38-22-31, -32) and Tank 43H (HTF-43-22-33, -34) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The results indicate the concentrations of most soluble species in the Tank 38H surface sample increased from the previous surface sample. The Tank 38H sub-surface sample shows changes in concentration for soluble species in the solution with some increasing and some decreasing. The current Tank 38H sub-surface sample contains less sludge solids than the previous sample based on visual appearance. The small differences in the concentrations of major components between the Tank 38H surface and sub-surface samples indicate only minimal stratification of solution species within the tank.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

SAS4A/SASSYS-1 Commercial Grade Dedication Example Report for a Generic Sodium Pool-Type Fast Reactor Application

In the U.S., a key component of the commercialization of advanced reactors is completion of a license application, which must ultimately be approved by the Nuclear Regulatory Commission (NRC). The approval of the license application by the NRC is contingent on satisfactory demonstration of the design basis and the response of the advanced reactor design to transient and accident scenarios using accepted codes and methods. This report describes the qualification and dedication requirements that the advanced reactor safety analysis system software SAS4A/SASSYS-1 are expected to need to fulfill to be used for sodium-cooled pool-type fast reactor licensing. The qualification and dedication requirements are identified through performance critical characteristics and evaluation model acceptance criteria representative of the advanced reactor design considered for licensing. This document captures, additionally, the verification process developed to demonstrate that the software fulfills the qualification and dedication requirements for a generic sodium-cooled pool-type fast reactor as part of the commercial grade dedication process. Like most software that has primarily existed in the research and development space, the most significant challenge facing SAS4A/SASSYS-1 for use in a licensing framework is the availability of a documentation basis describing the code pedigree. SAS4A/SASSYS-1 has been used for licensing of the fast flux test facility (FFTF) and the JOYO sodium-cooled fast reactor in Japan, as well as the design of the CRBR Plant. However, the historical verification and validation (V&V) activities supporting SAS4A/SASSYS-1 development do not align with modern software quality assurance (SQA) and V&V requirements. Two approaches to use of SAS4A/SASSYS-1 in a commercial licensing framework have been identified: commercial-grade dedication (CGD) and software qualification. The methods and requirements prescribed in the ASME NQA-1-2008/2009 Standard and Regulatory Guide 1.203 on the evaluation model development and assessment process (EMDAP) have been used as guidance to define the CGD and qualification processes, respectively. A qualification and dedication requirements matrix has been developed which utilizes fundamental software verification. In this process, software verification is defined as a software quality process aimed at defining software requirement specifications, developing software design documentation, and performing and documenting acceptance testing of the code against requirements. A key element of software qualification and dedication includes determination of software acceptance with respect to critical characteristics relevant to the functional requirements of the software. To assist with identification of cross-cutting transient phenomena and functional requirements, domestic SFR vendor designs have been reviewed to identify a reference SFR design. For this report, the reference design is defined as a pool-type reactor with metal alloy fuel, a liquid-metal intermediate heat transport system, and passive decay heat rejection systems. Given this reference, a series of high-level cross-cutting phenomena was identified for a general class of single-fault undercooling or reactivity insertion transients that scopes the design basis space, with the goal of assisting with prioritization of documentation development efforts for key transient models in SAS: 1) Reactivity feedback response prior to scram; 2) System-wide thermal inertia; 3) Transition in natural circulation flow regime in heat removal systems; 4) Decay heat generation; 5) Steady-state fuel characterization; 5) Clad/fuel behavior at elevated temperatures; 6) Point kinetics and decay heat; 7) Pump coastdown behavior; 8) Core flow redistribution in loss of forced convection; 9) Pool stratification. As a demonstration of CGD of SAS4A/SASSYS-1 for a sodium pool reactor, a software qualification and dedication gap analysis as it relates to code documentation has been performed. This effort leverages the framework established as part of the SAS4A/SASSYS-1 SQA Program. This CGD demonstration provides a framework that vendors can build upon to demonstrate the applicability of the SAS4A/SASSYS-1 software for licensing a sodium-cooled pool-type fast reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Refinement of Pitting Factor Basis to Support the Corrosion Control Program (Interim Report)

At Savannah River Site (SRS), High-Level Waste is stored in below-grade tanks constructed of carbon steel. This waste is composed of sludge, salt cake, and/or supernate. In part, preparation of this waste for future processing involves dissolution of the salt cake layer. The salt dissolution process can create conditions that leave the carbon steel tanks susceptible to a number of corrosion processes. The salt to be dissolved contains high concentrations of nitrate, that once released, create an environment that may be conducive to pitting corrosion or stress corrosion cracking (SCC) of carbon steel. Additionally, during the salt dissolution process, in the absence of mixing, stratification of the supernatant liquid may occur. This can result in less dense, more dilute waste layers occurring higher in the tank. In these more dilute waste layers, the susceptibility to localized corrosion could potentially differ from that of the more concentrated salt solutions evaluated in previous testing, as the amount of inhibiting and aggressive species, not just the ratios, can affect susceptibility. Evaluation of the susceptibility to localized corrosion in these more dilute waste chemistries could provide insight into the amount of inhibitors required to effectively mitigate pitting corrosion in stagnant salt dissolution environments, as well as other tank farm operations involving dilute waste streams.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Tank 38H (HTF-38-23-95, -96) and Tank 43H (HTF-43-23-93, -94) Samples for Support of the Enrichment Control and Corrosion Control Programs

Savannah River National Laboratory analyzed samples from Tank 38H and Tank 43H to support Enrichment Control Program (ECP) and Corrosion Control Program (CCP). The results indicate the concentrations of most soluble species in the Tank 38H surface sample increased significantly from the previous Tank 38H surface sample. The current Tank 38H subsurface sample shows similar Na, free hydroxide, and anions versus the previous subsurface sample. However, the 38H subsurface sample shows higher concentrations of Al, Ca, Fe, Mn, and Si vs. the previous Tank 38H subsurface sample. The current Tank 38H subsurface sample contained visible sludge solids in excess of the previous sample based on visual appearance. Weight percent solids measurements indicate presence of 3.0 ± 0.1 wt.% insoluble solids in the Tank 38H subsurface sample. The significant differences in the concentrations of major components between the Tank 38H surface and subsurface samples indicate significant stratification of solution species between these two locations within the Tank 38H. Savannah River Mission Completion (SRMC) personnel indicated that there were no tank-to-tank transfers into Tank 38H since early January 2023 and the 2H (16H) Evaporator was shut down on 3/26/2023 and has not operated since that time. There have been many pumped non-waste transfers of water from the H-Area diversion box 7 (HDB-7) sump into Tank 38 since the 3/26/2023 date.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗