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Development and Demonstration of Material Properties Database and Software for the Simulation of Flow Properties in Cementitious Materials

This report describes work performed by the Savannah River National Laboratory (SRNL) in fiscal year 2014 to develop a new Cementitious Barriers Project (CBP) software module designated as FLOExcel. FLOExcel incorporates a uniform database to capture material characterization data and a GoldSim model to define flow properties for both intact and fractured cementitious materials and estimate Darcy velocity based on specified hydraulic head gradient and matric tension. The software module includes hydraulic parameters for intact cementitious and granular materials in the database and a standalone GoldSim framework to manipulate the data. The database will be updated with new data as it comes available. The software module will later be integrated into the next release of the CBP Toolbox, Version 3.0. This report documents the development efforts for this software module. The FY14 activities described in this report focused on the following two items that form the FLOExcel package; 1) Development of a uniform database to capture CBP data for cementitious materials. In particular, the inclusion and use of hydraulic properties of the materials are emphasized; and 2) Development of algorithms and a GoldSim User Interface to calculate hydraulic flow properties of degraded and fractured cementitious materials. Hydraulic properties are required in a simulation of flow through cementitious materials such as Saltstone, waste tank fill grout, and concrete barriers. At SRNL these simulations have been performed using the PORFLOW code as part of Performance Assessments for salt waste disposal and waste tank closure.

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Actinide Removal Process October 2014 Sample Analysis

Filtration within the Actinide Removal Process (ARP) currently limits the throughput in interim salt processing at the Savannah River Site. In this process, batches of salt solution with Monosodium Titanate (MST) sorbent are concentrated by crossflow filtration. The filtrate is subsequently processed to remove cesium in the Modular Caustic Side Solvent Extraction Unit (MCU) followed by disposal in saltstone grout. The concentrated MST slurry is washed and sent to the Defense Waste Processing Facility (DWPF) for vitrification. During past ARP processing, there has been a degradation of filter performance manifested as the inability to maintain high filtrate flux throughout a multi-batch cycles during Salt Batches 6, 5 and 4. The objectives of the current effort for Salt Batch 7 samples, which are similar to the most recent ARP sample analysis from 2013, were to characterize the feed streams, to determine if solids (in addition to MST) are precipitating and causing the degraded performance of the filters, and to assess the particle size and rheological data to address potential filtration impacts. The samples from the LWPT and LWHT were obtained from several stages of processing of Salt Batch 7B, Cycle 1, Batch 67.

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Technical Approach for Determining Key Parameters Needed for Modeling the Performance of Cast Stone for the Integrated Disposal Facility Performance Assessment

Washington River Protection Solutions (WRPS) and its contractors at Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) are conducting a development program to develop / refine the cementitious waste form for the wastes treated at the ETF and to provide the data needed to support the IDF PA. This technical approach document is intended to provide guidance to the cementitious waste form development program with respect to the waste form characterization and testing information needed to support the IDF PA. At the time of the preparation of this technical approach document, the IDF PA effort is just getting started and the approach to analyze the performance of the cementitious waste form has not been determined. Therefore, this document looks at a number of different approaches for evaluating the waste form performance and describes the testing needed to provide data for each approach. Though the approach addresses a cementitious secondary aqueous waste form, it is applicable to other waste forms such as Cast Stone for supplemental immobilization of Hanford LAW. The performance of Cast Stone as a physical and chemical barrier to the release of contaminants of concern (COCs) from solidification of Hanford liquid low activity waste (LAW) and secondary wastes processed through the Effluent Treatment Facility (ETF) is of critical importance to the Hanford Integrated Disposal Facility (IDF) total system performance assessment (TSPA). The effectiveness of cementitious waste forms as a barrier to COC release is expected to evolve with time. PA modeling must therefore anticipate and address processes, properties, and conditions that alter the physical and chemical controls on COC transport in the cementitious waste forms over time. Most organizations responsible for disposal facility operation and their regulators support an iterative hierarchical safety/performance assessment approach with a general philosophy that modeling provides the critical link between the short-term understanding from laboratory and field tests, and the prediction of repository performance over repository time frames and scales. One common recommendation is that experiments be designed to permit the appropriate scaling in the models. There is a large contrast in the physical and chemical properties between the Cast Stone waste package and the IDF backfill and surrounding sediments. Cast Stone exhibits low permeability, high tortuosity, low carbonate, high pH, and low Eh whereas the backfill and native sediments have high permeability, low tortuosity, high carbonate, circumneutral pH, and high Eh. These contrasts have important implications for flow, transport, and reactions across the Cast Stone – backfill interface. Over time with transport across the interface and subsequent reactions, the sharp geochemical contrast will blur and there will be a range of spatially-distributed conditions. In general, COC mobility and transport will be sensitive to these geochemical variations, which also include physical changes in porosity and permeability from mineral reactions. Therefore, PA modeling must address processes, properties, and conditions that alter the physical and chemical controls on COC transport in the cementitious waste forms over time. Section 2 of this document reviews past Hanford PAs and SRS Saltstone PAs, which to date have mostly relied on the lumped parameter COC release conceptual models for TSPA predictions, and provides some details on the chosen values for the lumped parameters. Section 3 provides more details on the hierarchical modeling strategy and processes and mechanisms that control COC release. Section 4 summarizes and lists the key parameters for which numerical values are needed to perform PAs. Section 5 provides brief summaries of the methods used to measure the needed parameters and references to get more details.

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Coupling of Nuclear Waste Form Corrosion and Radionuclide Transports in Presence of Relevant Repository Sediments

Assessments of waste form and disposal options start with the degradation of the waste forms and consequent mobilization of radionuclides. Long-term static tests, single-pass flow-through tests, and the pressurized unsaturated flow test are often employed to study the durability of potential waste forms and to help create models that predict their durability throughout the lifespan of the disposal site. These tests involve the corrosion of the material in the presence of various leachants, with different experimental designs yielding desired information about the behavior of the material. Though these tests have proved instrumental in elucidating various mechanisms responsible for material corrosion, the chemical environment to which the material is subject is often not representative of a potential radioactive waste repository where factors such as pH and leachant composition will be controlled by the near-field environment. Near-field materials include, but are not limited to, the original engineered barriers, their resulting corrosion products, backfill materials, and the natural host rock. For an accurate performance assessment of a nuclear waste repository, realistic waste corrosion experimental data ought to be modeled to allow for a better understanding of waste form corrosion mechanisms and the effect of immediate geochemical environment on these mechanisms. Additionally, the migration of radionuclides in the resulting chemical environment during and after waste form corrosion must be quantified and mechanisms responsible for migrations understood. The goal of this research was to understand the mechanisms responsible for waste form corrosion in the presence of relevant repository sediments to allow for accurate radionuclide migration quantifications. The rationale for this work is that a better understanding of waste form corrosion in relevant systems will enable increased reliance on waste form performance in repository environments and potentially decrease the need for expensive engineered barriers.Our current work aims are 1) quantifying and understanding the processes associated with glass alteration in contact with Fe-bearing materials; 2) quantifying and understanding the processes associated with glass alteration in presence of MgO (example of engineered barrier used in WIPP); 3) identifying glass alteration suppressants and the processes involved to reach glass alteration suppression; 4) quantifying and understanding the processes associated with Saltstone and Cast Stone (SRS and Hanford cementitious waste forms) in various representative groundwaters; 5) investigating positron annihilation as a new tool for the study of glass alteration; and 6) quantifying and understanding the processes associated with glass alteration under gamma irradiation.

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Geochemical Data Package for Performance Assessment Calculations Related to the Savannah River Site

The Savannah River Site (SRS) disposes of low-level radioactive waste (LLW) and stabilizes high-level radioactive waste (HLW) tanks in the subsurface environment. Calculations used to establish the radiological limits of these facilities are referred to as Performance Assessments (PA), Special Analyses (SA), and Composite Analyses (CA). The objective of this document is to revise existing geochemical input values used for these calculations. This work builds on earlier compilations of geochemical data (2007, 2010), referred to a geochemical data packages. This work is being conducted as part of the on-going maintenance program of the SRS PA programs that periodically updates calculations and data packages when new information becomes available. Because application of values without full understanding of their original purpose may lead to misuse, this document also provides the geochemical conceptual model, the approach used for selecting the values, the justification for selecting data, and the assumptions made to assure that the conceptual and numerical geochemical models are reasonably conservative (i.e., bias the recommended input values to reflect conditions that will tend to predict the maximum risk to the hypothetical recipient). This document provides 1088 input parameters for geochemical parameters describing transport processes for 64 elements (>740 radioisotopes) potentially occurring within eight subsurface disposal or tank closure areas: Slit Trenches (ST), Engineered Trenches (ET), Low Activity Waste Vault (LAWV), Intermediate Level (ILV) Vaults, Naval Reactor Component Disposal Areas (NRCDA), Components-in-Grout (CIG) Trenches, Saltstone Facility, and Closed Liquid Waste Tanks. The geochemical parameters described here are the distribution coefficient, Kd value, apparent solubility concentration, k s value, and the cementitious leachate impact factor.

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OPC Paste Samples Exposed To Aggressive Solutions. Cementitious Barriers Partnership

The study presented in this report focused on a low-activity wasteform containing a high-pH pore solution with a significant level of sulfate. The purpose of the study was to improve understanding of the complex concrete/wasteform reactive transport problem, in particular, the role of pH in sulfate attack. Paste samples prepared at three different water-to-cement ratios were tested. The mixtures were prepared with ASTM Type I cement, without additional admixtures. The samples were exposed to two different sodium sulfate contact solutions. The first solution was prepared at 0.15M Na 2 SO 4 . The second solution also incorporated 0.5M NaOH, to mimic the high pH conditions found in Saltstone. The data collected indicated that, in Na 2 SO 4 solution, damage occurs to the pastes. In the case of the high-pH sulfate solution (Na 2 SO 4 + NaOH), no signs of damage were observed on any of the paste mixtures. These results indicate that the high sulfate content found in the wasteform pore solution will not necessarily lead to severe damage to concrete. Good-quality mixtures could thus prove durable over the long term, and act as an effective barrier to prevent radionuclides from reaching the environment.

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SCM Paste Samples Exposed To Aggressive Solutions. Cementitious Barriers Partnership

This report summarizes experimental work performed by SIMCO Technologies Inc. (SIMCO) as part of the Cementitious Barriers Partnership (CBP) project. The test series followed an experimental program dedicated to the study of ordinary Portland cement (OPC) hydrated cement pastes exposed to aggressive solutions. In the present study, the scope is extended to hydrated cement pastes incorporating supplementary cementitious materials (SCM) such as fly ash and ground granulated blast furnace slag (GGBFS). Also, the range of aggressive contact solutions was expanded. The experimental program aimed at testing aggressive contact solutions that more closely mimic the chemical composition of saltstone pore solution. Five different solutions, some of which incorporated high levels of carbonate and nitrate, were placed in contact with four different hydrated cement paste mixes. In all solutions, 150 mmol/L of SO 4 2– (14 400 ppm) were present. The solutions included different pH conditions and different sodium content. Two paste mixes were equivalent to Vault 1/4 and Vault 2 concrete mixes used at SRS in storage structures. Two additional paste mixes, cast at the same water-to-cement ratio and using the same cements but without SCMs, were also tested. The damage evolution in samples was monitored using ultrasonic pulse velocity (UPV) and mass measurements. After three and twelve months of exposure conditions, samples were taken out of solution containers and analyzed to perform migration tests and porosity measurements. Globally, results were in line with the previous study and confirmed that high pH may limit the formation of some deleterious phases like gypsum. In this case, ettringite may form but is not necessarily associated with damage. However, the high concentration of sodium may be associated with the formation of an AFm-like mineral called U-phase. The most significant evidences of damage were all associated with the Vault 2 paste analog. This material proved very sensitive to high pH. All measurement techniques used to monitor and evaluate damage to samples indicated significant alterations to this mix when immersed in contact solutions containing sodium hydroxide. It was hypothesized that the low cement content, combined with high silica content coming from silica fume, fly ash and GGBFS led to the presence unreacted silica. It is possible that the pozzolanic reaction of these SCMs could not be activated due to the low alkali content, a direct consequence of low cement content. In this scenario, the material end up having a lot of silica available to react upon contact with sodium hydroxide, possibly forming a gel that may be similar to the gel formed in alkali-silica reactions. This scenario needs further experimental confirmation, but it may well explain the poor behavior of mix PV2 in presence of NaOH.

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CBP Toolbox Version 3.0 “Beta Testing” Performance Evaluation

One function of the Cementitious Barriers Partnership (CBP) is to assess available models of cement degradation and to assemble suitable models into a “Toolbox” that would be made available to members of the partnership, as well as the DOE Complex. To this end, SRNL and Vanderbilt University collaborated to develop an interface using the GoldSim software to the STADIUM @ code developed by SIMCO Technologies, Inc. and LeachXS/ORCHESTRA developed by Energy research Centre of the Netherlands (ECN). Release of Version 3.0 of the CBP Toolbox is planned in the near future. As a part of this release, an increased level of quality assurance for the partner codes and the GoldSim interface has been developed. This report documents results from evaluation testing of the ability of CBP Toolbox 3.0 to perform simulations of concrete degradation applicable to performance assessment of waste disposal facilities. Simulations of the behavior of Savannah River Saltstone Vault 2 and Vault 1/4 concrete subject to sulfate attack and carbonation over a 500- to 1000-year time period were run using a new and upgraded version of the STADIUM @ code and the version of LeachXS/ORCHESTRA released in Version 2.0 of the CBP Toolbox. Running both codes allowed comparison of results from two models which take very different approaches to simulating cement degradation. In addition, simulations of chloride attack on the two concretes were made using the STADIUM @ code. The evaluation sought to demonstrate that: 1) the codes are capable of running extended realistic simulations in a reasonable amount of time; 2) the codes produce “reasonable” results; the code developers have provided validation test results as part of their code QA documentation; and 3) the two codes produce results that are consistent with one another. Results of the evaluation testing showed that the three criteria listed above were met by the CBP partner codes. Therefore, it is concluded that the codes can be used to support performance assessment. This conclusion takes into account the QA documentation produced for the partner codes and for the CBP Toolbox.

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Defense Waste Processing Facility Simulant Chemical Processing Cell Studies for Sludge Batch 9

The Savannah River National Laboratory (SRNL) received a technical task request from Defense Waste Processing Facility (DWPF) and Saltstone Engineering to perform simulant tests to support the qualification of Sludge Batch 9 (SB9) and to develop the flowsheet for SB9 in the DWPF. These efforts pertained to the DWPF Chemical Process Cell (CPC). CPC experiments were performed using SB9 simulant (SB9A) to qualify SB9 for sludge-only and coupled processing using the nitric-formic flowsheet in the DWPF. Two simulant batches were prepared, one representing SB8 Tank 40H and another representing SB9 Tank 51H. The simulant used for SB9 qualification testing was prepared by blending the SB8 Tank 40H and SB9 Tank 51H simulants. The blended simulant is referred to as SB9A. Eleven CPC experiments were run with an acid stoichiometry ranging between 105% and 145% of the Koopman minimum acid equation (KMA), which is equivalent to 109.7% and 151.5% of the Hsu minimum acid factor. Three runs were performed in the 1L laboratory scale setup, whereas the remainder were in the 4L laboratory scale setup. Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on nine of the eleven. The other two were SRAT cycles only. One coupled flowsheet and one extended run were performed for SRAT and SME processing. Samples of the condensate, sludge, and off-gas were taken to monitor the chemistry of the CPC experiments.

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Critical Radionuclide and Pathway Analysis for the Savannah River Site, 2016 Update

During the operational history of Savannah River Site, many different radionuclides have been released from site facilities. However, as shown in this analysis, only a relatively small number of the released radionuclides have been significant contributors to doses to the offsite public. This report is an update to the 2011 analysis, Critical Radionuclide and Pathway Analysis for the Savannah River Site. SRS-based Performance Assessments for E-Area, Saltstone, F-Tank Farm, H-Tank Farm, and a Comprehensive SRS Composite Analysis have been completed. The critical radionuclides and pathways identified in those extensive reports are also detailed and included in this analysis.

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Nitric-glycolic flowsheet evaluation with the slurry-fed melt rate furnace

The Savannah River National Laboratory (SRNL) was tasked to support validation of the Defense Waste Processing Facility (DWPF) melter offgas flammability model for the nitric-glycolic (NG) flowsheet. The work supports Deliverable 4 of the DWPF & Saltstone Facility Engineering Technical Task Request (TTR)1 and is supplemental to the Cold Cap Evaluation Furnace (CEF) testing conducted in 2014.2 The Slurry-fed Melt Rate Furnace (SMRF) was selected for the supplemental testing as it requires significantly less resources than the CEF and could provide a tool for more rapid analysis of melter feeds in the future. The SMRF platform has been used previously to evaluate melt rate behavior of DWPF glasses, but was modified to accommodate analysis of the offgas stream. Additionally, the Melt Rate Furnace (MRF) and Quartz Melt Rate Furnace (QMRF) were utilized for evaluations. MRF data was used exclusively for melt behavior observations and REDuction/OXidation (REDOX) prediction comparisons and will be briefly discussed in conjunction with its support of the SMRF testing. The QMRF was operated similarly to the SMRF for the same TTR task, but will be discussed in a separate future report. The overall objectives of the SMRF testing were to; 1) Evaluate the efficacy of the SMRF as a platform for steady state melter testing with continuous feeding and offgas analysis; and 2) Generate supplemental melter offgas flammability data to support the melter offgas flammability modelling effort for DWPF implementation of the NG flowsheet.

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Liquid secondary waste: Waste form formulation and qualification

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity and water characteristic curves) were comparable to the properties measured on the Savannah River Site (SRS) Saltstone waste form. Future testing should include efforts to first; 1) determine the rate and amount of ammonia released during each unit operation of the treatment process to determine if additional ammonia management is required, then; 2) reduce the ammonia content of the ETF concentrated brine prior to solidification, making the waste more amenable to grouting, or 3) manage the release of ammonia during production and ongoing release during storage of the waste form, or 4) develop a lower pH process/waste form thereby precluding ammonia release.

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Evaluation of quartz melt rate furnace with the nitric-glycolic flowsheet

The Savannah River National Laboratory (SRNL) was tasked to support validation of the Defense Waste Processing Facility (DWPF) melter offgas flammability model for the Nitric-Glycolic (NG) flowsheet. The work is supplemental to the Cold Cap Evaluation Furnace (CEF) testing conducted in 20141 and the Slurry-fed Melt Rate Furnace (SMRF) testing conducted in 20162 that supported Deliverable 4 of the DWPF & Saltstone Facility Engineering Technical Task Request (TTR).3 The Quartz Melt Rate Furnace (QMRF) was evaluated as a bench-scale scoping tool to potentially be used in lieu of or simply prior to the use of the larger-scale SMRF or CEF. The QMRF platform has been used previously to evaluate melt rate behavior and offgas compositions of DWPF glasses prepared from the Nitric-Formic (NF) flowsheet but not for the NG flowsheet and not with continuous feeding.4 The overall objective of the 2016-2017 testing was to evaluate the efficacy of the QMRF as a lab-scale platform for steady state, continuously fed melter testing with the NG flowsheet as an alternative to more expensive and complex testing with the SMRF or CEF platforms.

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Groundwater flow simulation of the Savannah River Site general separations area

The most recent groundwater flow model of the General Separations Area, Savannah River Site, is referred to as the “GSA/PORFLOW” model. GSA/PORFLOW was developed in 2004 by porting an existing General Separations Area groundwater flow model from the FACT code to the PORFLOW code. The preceding “GSA/FACT” model was developed in 1997 using characterization and monitoring data through the mid-1990’s. Both models were manually calibrated to field data. Significantly more field data have been acquired since the 1990’s and model calibration using mathematical optimization software has become routine and recommended practice. The current task involved updating the GSA/PORFLOW model using selected field data current through at least 2015, and use of the PEST code to calibrate the model and quantify parameter uncertainty. This new GSA groundwater flow model is named “GSA2016” in reference to the year in which most development occurred. The GSA2016 model update is intended to address issues raised by the DOE Low-Level Waste (LLW) Disposal Facility Federal Review Group (LFRG) in a 2008 review of the E-Area Performance Assessment, and by the Nuclear Regulatory Commission in reviews of tank closure and Saltstone Disposal Facility Performance Assessments.

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Speciation of methylmercury and ethylmercury by gas chromatography cold vapor atomic fluresence spectroscopy

Existing models and simulants of tank disposition media at SRS have presumed the presence of high concentrations of inorganic mercury. However, recent quarterly tank analyses show that mercury is present as organomercurial species at concentrations that may present challenges to remediation and disposition and may exceed the Saltstone Waste Acceptance Criteria (WAC). To-date, methylmercury analysis for Savannah River Remediation (SRR) has been performed off-site by Eurofins Scientific (Lancaster, PA). A series of optimization and validation experiments has been performed at SRNL, which has resulted in the development of on-site organomercury speciation capabilities using purge and trap gas chromatography coupled with thermal desorption cold vapor atomic fluorescence spectroscopy (P&T GC/CVAFS). Speciation has been achieved for methylmercury, with a method reporting limit (MRL) values of 1.42 pg for methylmercury. Results obtained by SRNL from the analysis of past quarterly samples from tanks 21, 40, and 50 have demonstrated statistically indistinguishable concentration values compared with the concentration data obtained from Eurofins, while the data from SRNL has demonstrated significantly improved precision and processing time.

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Impacts of Updated GSA Groundwater Flow Models on the FTF, HTF and SDF PAs

Savannah River Remediation (SRR) requested that the Savannah River National Laboratory (SRNL) evaluate the impacts of an updated General Separations Area (GSA) groundwater flow model on the F-Area Tank Farm (FTF), H-Area Tank Farm (HTF) and Saltstone Disposal Facility (SDF) Performance Assessments (PAs)1. For each facility, PORFLOW simulations were performed using both the 2004 GSA/PORFLOW (i.e., GSA_2004) and GSA2016 (i.e., GSA_2016) flow models. Overall, GSA_2004 and “_reference” (i.e., the archived data) results are practically identical, indicating GSA_2004 simulations reproduce the archived data very well. Compared with GSA_2004 data, the PEST.47 GSA_2016 results (with PORFLOW harmonic averaging at cell faces and diagonal terms in the dispersion tensor) support the following findings.

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Hyperspectral Raman imaging using a spatial heterodyne spectrometer (SHS)

Savannah River National Laboratory, and the larger community, has limited chemical-specific imaging capabilities, which would be beneficial for remote detection of threat chemicals or use in inaccessible and harsh environments, as well as in-line process monitoring. Chemical imaging technology would support multiple directorates through stand-off detection of chemicals (National Security), and imaging of chemical distributions in a process (Nuclear Materials Management, Clean Energy) and in waste tanks (Environmental Stewardship). Chemical-specific imaging is underdeveloped due to issues of sensitivity, stable alignment and calibration, and ease of operation associated with most optical instruments. The project objective is to develop a novel, rugged, highly sensitive spectrometer to support real-time, chemically specific imaging using hyperspectral Raman spectroscopy. Raman spectroscopic analysis will provide chemical specificity and using a spatial heterodyne spectrometer (SHS) will increase the sensitivity due to the high light throughput design. The SHS design does not require moving parts allowing for a very stable system, reducing alignment and calibration issues. The instrument will be assembled and initially applied to Raman gas detection of the Saltstone disposal unit headspace gas concentrations. Raman gas analysis is the most impacted by sensitivity issues and will be used to demonstrate the increased light collection capability of the SHS over a conventional dispersive spectrometer. The system will be later modified to image spectral information in a spatial domain to provide information on the spatial distribution of a sample scene.

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Investigation Of Thermolysis Hydrogen Generation Rate In Tank 38 And Tank 50 Waste Samples With Sodium Glycolate

This report contains hydrogen generation rate (HGR) measurements of Tank 38 and Tank 50 radioactive samples with and without added glycolate as a function of temperature. The objective of the Tank 38 and Tank 50 HGR measurements is to extend the knowledge from previous sample thermolysis HGR measurements and augment simulant testing for developing an equation for the Savannah River Site (SRS) Concentration, Storage, and Transfer Facility (CSTF) thermolysis HGR with and without added glycolate. Tank 38 testing is applicable to the 2H-Evaporator System and associated CSTF waste tanks. Tank 50 HGR measurement is applicable to CSTF tanks and the Saltstone Production Facility (SPF). The glycolate concentration used is based on what would be expected to be near the high-end concentration of glycolate (1570 mg/L in Tank 38 and 350 mg/L in Tank 50). The radioactive dose rate of the Tank 38 material was high enough to require use of the flow system in the Shielded Cells. Tank 50 material was less radioactive, so the higher temperatures were investigated in the flow system while the lower temperatures were investigated in the sealed system contained within a radiological hood.

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