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DOE OSTI2016

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2013

Verification Of the Defense Waste Processing Facility's (DWPF) Process Digestion Methods for the Sludge Batch 8 Qualification Sample

This report contains the results and comparison of data generated from inductively coupled plasma – atomic emission spectroscopy (ICP-AES) analysis of Aqua Regia (AR), Sodium Peroxide/Sodium Hydroxide Fusion Dissolution (PF) and Cold Chem (CC) method digestions and Cold Vapor Atomic Absorption analysis of Hg digestions from the DWPF Hg digestion method of Sludge Batch 8 (SB8) Sludge Receipt and Adjustment Tank (SRAT) Receipt and SB8 SRAT Product samples. The SB8 SRAT Receipt and SB8 SRAT Product samples were prepared in the SRNL Shielded Cells, and the SRAT Receipt material is representative of the sludge that constitutes the SB8 Batch or qualification composition. This is the sludge in Tank 51 that is to be transferred into Tank 40, which will contain the heel of Sludge Batch 7b (SB7b), to form the SB8 Blend composition.

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DOE OSTI2013

DWPF Simulant CPC Studies For SB8

Prior to processing a Sludge Batch (SB) in the Defense Waste Processing Facility (DWPF), flowsheet studies using simulants are performed. Typically, the flowsheet studies are conducted based on projected composition(s). The results from the flowsheet testing are used to 1) guide decisions during sludge batch preparation, 2) serve as a preliminary evaluation of potential processing issues, and 3) provide a basis to support the Shielded Cells qualification runs performed at the Savannah River National Laboratory (SRNL). SB8 was initially projected to be a combination of the Tank 40 heel (Sludge Batch 7b), Tank 13, Tank 12, and the Tank 51 heel. In order to accelerate preparation of SB8, the decision was made to delay the oxalate-rich material from Tank 12 to a future sludge batch. SB8 simulant studies without Tank 12 were reported in a separate report.1 The data presented in this report will be useful when processing future sludge batches containing Tank 12. The wash endpoint target for SB8 was set at a significantly higher sodium concentration to allow acceptable glass compositions at the targeted waste loading. Four non-coupled tests were conducted using simulant representing Tank 40 at 110-146% of the Koopman Minimum Acid requirement. Hydrogen was generated during high acid stoichiometry (146% acid) SRAT testing up to 31% of the DWPF hydrogen limit. SME hydrogen generation reached 48% of of the DWPF limit for the high acid run. Two non-coupled tests were conducted using simulant representing Tank 51 at 110-146% of the Koopman Minimum Acid requirement. Hydrogen was generated during high acid stoichiometry SRAT testing up to 16% of the DWPF limit. SME hydrogen generation reached 49% of the DWPF limit for hydrogen in the SME for the high acid run. Simulant processing was successful using previously established antifoam addition strategy. Foaming during formic acid addition was not observed in any of the runs. Nitrite was destroyed in all runs and no N2O was detected during SME processing. Mercury behavior was consistent with that seen in previous SRAT runs. Mercury was stripped below the DWPF limit on 0.8 wt% for all runs. Rheology yield stress fell within or below the design basis of 1-5 Pa. The low acid Tank 40 run (106% acid stoichiometry) had the highest yield stress at 3.78 Pa.

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DOE OSTI2014

Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3 He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2014

Laboratory studies of 2H evaporator scale dissolution in dilute nitric acid

The rate of 2H evaporator scale solids dissolution in dilute nitric acid has been experimentally evaluated under laboratory conditions in the SRNL shielded cells. The 2H scale sample used for the dissolution study came from the bottom of the evaporator cone section and the wall section of the evaporator cone. The accumulation rate of aluminum and silicon, assumed to be the two principal elemental constituents of the 2H evaporator scale aluminosilicate mineral, were monitored in solution. Aluminum and silicon concentration changes, with heating time at a constant oven temperature of 90 deg C, were used to ascertain the extent of dissolution of the 2H evaporator scale mineral. The 2H evaporator scale solids, assumed to be composed of mostly aluminosilicate mineral, readily dissolves in 1.5 and 1.25 M dilute nitric acid solutions yielding principal elemental components of aluminum and silicon in solution. The 2H scale dissolution rate constant, based on aluminum accumulation in 1.5 and 1.25 M dilute nitric acid solution are, respectively, 9.21E-04 ± 6.39E-04 min{sup -1} and 1.07E-03 ± 7.51E-05 min{sup -1}. Silicon accumulation rate in solution does track the aluminum accumulation profile during the first few minutes of scale dissolution. It however diverges towards the end of the scale dissolution. This divergence therefore means the aluminum-to-silicon ratio in the first phase of the scale dissolution (non-steady state conditions) is different from the ratio towards the end of the scale dissolution. Possible causes of this change in silicon accumulation in solution as the scale dissolution progresses may include silicon precipitation from solution or the 2H evaporator scale is a heterogeneous mixture of aluminosilicate minerals with several impurities. The average half-life for the decomposition of the 2H evaporator scale mineral in 1.5 M nitric acid is 12.5 hours, while the half-life for the decomposition of the 2H evaporator scale in 1.25 M nitric acid is 10.8 hours. Based on averaging the two half-lives from the 2H scale acid dissolution in 1.25 and 1.5 M nitric acid solutions, a reasonable half-live for the dissolution of 2H scales in dilute nitric acid is 11.7 ± 1.3 hours. The plant operational time for chemically cleaning (soaking) the 2H evaporator with dilute nitric acid is 32 hours. It therefore may require about 3 half-lives or less to completely dissolve most of the scales in the Evaporator pot which come into contact with the dilute nitric acid solution. On a mass basis, the Al-to-Si ratio for the scale dissolution in 1.5 M nitric acid averaged 1.30 ± 0.20 and averaged 1.18 ± 0.10 for the 2H scale dissolution in 1.25 M nitric acid. These aluminum-to-silicon ratios are in fairly good agreement with ratios from previous studies. Therefore, there is still more aluminum in the 2H evaporator scales than silicon which implies that there are no significant changes in scale properties which will exclude nitric acid as a viable protic solvent for aluminosilicate scale buildup dissolution from the 2H evaporator. Overall, the monitoring of the scale decomposition reaction in 1.25 and 1.5 M nitric acid may be better ascertained through the determination of aluminum concentration in solution than monitoring silicon in solution. Silicon solution chemistry may lead to partial precipitating of silicon with time as the scale and acid solution is heated.

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DOE OSTI2014

SRS Tank 38H and 43H Supernate Foaming Studies and Tank 38H Reel Tape Solids Characterization

Radioactive waste samples retrieved from Savannah River Site (SRS) Tanks 38H and 43H (concentrate receipt and feed tanks, respectively, for the 2H Evaporator system) were evaluated with regard to their tendency to form foams during air sparging. This work was conducted due to recent processing issues and outages in the evaporator that were believed to have resulted from sample foaming. The samples evaluated for foam formation included supernate collected in April of 2014 (near the time of the evaporator outage) as well as historical samples available within the SRNL shielded cells facility. The April samples included one Tank 43H surface sample (HTF-43-14-42), one Tank 43H sub-surface sample (HTF-43-14-43), and one Tank 38H (HTF-38-14-41) surface sample. In addition, two Tank 43H samples (HTF-43-14-8 and HTF-43-14-9) and one Tank 38H sample (HTF-38-14-6) were also evaluated along with a blended sample of various historical Tank 38H and 43H samples. Characterization results for the April samples are also provided. The composition of the samples was similar to historical evaporator system samples received at SRNL.

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DOE OSTI2015

Results of Preliminary Hg Speciation Testing on 4Q14 Tank 50, 1Q15 Tank 50, and SRNL 14-Day TCLP Leachate

The Savannah River National Laboratory (SRNL) was tasked with preparing and shipping samples for Hg speciation by Eurofins Frontier Global Sciences, Inc. in Seattle, WA on behalf of the Savannah River Remediation (SRR) Mercury Task Team. The first shipment of samples was designated to include existing 4Q14 Tank 50 WAC material archived in the SRNL Shielded Cells; 1Q15 Tank 50 material pulled on February 25, 2015; and leachate from the 14-day SRNL conducted Toxicity Characteristic Leaching Procedure (TCLP) performed on a sample of saltstone prepared from the official 1Q15 Tank 50 WAC sample. Eurofins supplied deionized water, 250 mL PETG bottles, 125 mL amber glass bottles, and preservative (1.2 mL concentrated HCl). Tank 50 samples were prepared in triplicate plus a blank, while duplicate samples of the leachate plus a blank were prepared. Each sample was analyzed for five Hg species: total Hg, elemental Hg [Hg(0)], ionic Hg [Hg(I) and Hg(II)], methyl Hg [CH 3 Hg-X, where X is a counter anion], and dimethyl Hg [(CH 3 ) 2 Hg]. The analytes were determined from samples in three separate bottles: 1) methyl Hg, 2) dimethyl Hg and elemental Hg, and 3) total Hg and ionic Hg. Initially it was communicated that Eurofins would determine Hg(0) from bottle 3), but it was reported from bottle 2) for each set of samples. In total, 33 samples were prepared on March 4, 2015 and shipped by next-day air to Eurofins where they were received on March 6, 2015. Details of the sample preparation activities are recorded in the SRNL E-Notebook system.i SRNL deionized water was employed as the blank for the Tank 50 samples, and uncontacted extraction fluid was submitted as the blank for the leachate.

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DOE OSTI2015

Results of Hg Speciation Testing on 2Q15 Tank 50 and Tank 21 TCLP Extraction Fluid Samples and Tank 49 Material

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DOE OSTI2015

Thermal Properties Capability Development Workshop Summary to Support the Implementation Plan for PIE Thermal Conductivity Measurements

The Department of Energy (DOE)-Office of Nuclear Energy (NE), Idaho National Laboratory (INL), and associated nuclear fuels programs have invested heavily over the years in infrastructure and capability development. With the current domestic and international need to develop Accident Tolerant Fuels (ATF), increasing importance is being placed on understanding fuel performance in irradiated conditions and on the need to model and validate that performance to reduce uncertainty and licensing timeframes. INL’s Thermal Properties Capability Development Workshop was organized to identify the capability needed by the various nuclear programs and list the opportunities to meet those needs. In addition, by the end of fiscal year 2015, the decision will be made on the initial thermal properties instruments to populate the shielded cell in the Irradiated Materials Characterization Laboratory (IMCL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Sludge batch 9 (SB9) acceptance evaluation. Radionuclide concentrations in tank 51 SB9 qualification sample prepared at SRNL

Presented in this report are radionuclide concentrations required as part of the program of qualifying Sludge Batch 9 (SB9) for processing in the Defense Waste Processing Facility (DWPF). The SB9 material is currently in Tank 51 and has been washed and prepared for transfer to Tank 40. The acceptance evaluation needs to be completed prior to the transfer of the material in Tank 51 to Tank 40. The sludge slurry in Tank 40 has already been qualified for DWPF processing and is currently being processed as Sludge Batch 8 (SB8). The radionuclide concentrations were measured or estimated in the Tank 51 SB9 Washed Qualification Sample prepared at Savannah River National Laboratory (SRNL). This sample was prepared from a three liter sample of Tank 51 sludge slurry (HTF-51-15-81) taken on July 23, 2015. The sample was delivered to SRNL where it was initially characterized in the Shielded Cells. Under the direction of Savannah River Remediation (SRR) it was then adjusted per the Tank Farm washing strategy as of October 20, 2015. This final slurry now has a composition expected to be similar to that of the slurry in Tank 51 after final preparations have been made for transfer of that slurry to Tank 40.

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DOE OSTI2016

Sludge batch 9 (SB9) accepance evaluation: Radionuclide concentrations in tank 51 SB9 qualification sample prepared at SRNL

Presented in this report are radionuclide concentrations required as part of the program of qualifying Sludge Batch 9 (SB9) for processing in the Defense Waste Processing Facility (DWPF). The SB9 material is currently in Tank 51 and has been washed and prepared for transfer to Tank 40. The acceptance evaluation needs to be completed prior to the transfer of the material in Tank 51 to Tank 40. The sludge slurry in Tank 40 has already been qualified for DWPF processing and is currently being processed as Sludge Batch 8 (SB8). The radionuclide concentrations were measured or estimated in the Tank 51 SB9 Washed Qualification Sample prepared at Savannah River National Laboratory (SRNL). This sample was prepared from a three liter sample of Tank 51 sludge slurry (HTF-51-15-81) taken on July 23, 2015. The sample was delivered to SRNL where it was initially characterized in the Shielded Cells. Under the direction of Savannah River Remediation (SRR) it was then adjusted per the Tank Farm washing strategy as of October 20, 2015. This final slurry now has a compositioniv expected to be similar to that of the slurry in Tank 51 after final preparations have been made for transfer of that slurry to Tank 40.

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DOE OSTI2016

WTP Waste Feed Qualification: Glass Fabrication Unit Operation Testing Report

The waste feed qualification program is being developed to protect the Hanford Tank Waste Treatment and Immobilization Plant (WTP) design, safety basis, and technical basis by assuring waste acceptance requirements are met for each staged waste feed campaign prior to transfer from the Tank Operations Contractor to the feed receipt vessels inside the Pretreatment Facility. The Waste Feed Qualification Program Plan describes the three components of waste feed qualification: 1. Demonstrate compliance with the waste acceptance criteria 2. Determine waste processability 3. Test unit operations at laboratory scale. The glass fabrication unit operation is the final step in the process demonstration portion of the waste feed qualification process. This unit operation generally consists of combining each of the waste feed streams (high-level waste (HLW) and low-activity waste (LAW)) with Glass Forming Chemicals (GFCs), fabricating glass coupons, performing chemical composition analysis before and after glass fabrication, measuring hydrogen generation rate either before or after glass former addition, measuring rheological properties before and after glass former addition, and visual observation of the resulting glass coupons. Critical aspects of this unit operation are mixing and sampling of the waste and melter feeds to ensure representative samples are obtained as well as ensuring the fabrication process for the glass coupon is adequate. Testing was performed using a range of simulants (LAW and HLW simulants), and these simulants were mixed with high and low bounding amounts of GFCs to evaluate the mixing, sampling, and glass preparation steps in shielded cells using laboratory techniques. The tests were performed with off-the-shelf equipment at the Savannah River National Laboratory (SRNL) that is similar to equipment used in the SRNL work during qualification of waste feed for the Defense Waste Processing Facility (DWPF) and other waste treatment facilities at the Savannah River Site. It is not expected that the exact equipment used during this testing will be used during the waste feed qualification testing for WTP, but functionally similar equipment will be used such that the techniques demonstrated would be applicable. For example, the mixing apparatus could use any suitable mixer capable of being remoted and achieving similar mixing speeds to those tested.

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DOE OSTI2016

WTP Waste Feed Qualification: Hydrogen Generation Rate Measurement Apparatus Testing Report

The generation rate of hydrogen gas in the Hanford tank waste will be measured during the qualification of the staged tank waste for processing in the Hanford Tank Waste Treatment and Immobilization Plant. Based on a review of past practices in measurement of the hydrogen generation, an apparatus to perform this measurement has been designed and tested for use during waste feed qualification. The hydrogen generation rate measurement apparatus (HGRMA) described in this document utilized a 100 milliliter sample in a continuously-purged, continuously-stirred vessel, with measurement of hydrogen concentration in the vent gas. The vessel and lid had a combined 220 milliliters of headspace. The vent gas system included a small condenser to prevent excessive evaporative losses from the sample during the test, as well as a demister and filter to prevent particle migration from the sample to the gas chromatography system. The gas chromatograph was an on line automated instrument with a large-volume sample-injection system to allow measurement of very low hydrogen concentrations. This instrument automatically sampled the vent gas from the hydrogen generation rate measurement apparatus every five minutes and performed data regression in real time. The fabrication of the hydrogen generation rate measurement apparatus was in accordance with twenty three (23) design requirements documented in the conceptual design package, as well as seven (7) required developmental activities documented in the task plan associated with this work scope. The HGRMA was initially tested for proof of concept with physical simulants, and a remote demonstration of the system was performed in the Savannah River National Laboratory Shielded Cells Mockup Facility. Final verification testing was performed using non-radioactive simulants of the Hanford tank waste. Three different simulants were tested to bound the expected rheological properties expected during waste feed qualification testing. These simulants were tested at different temperatures using purge gas spiked with varying amounts of hydrogen to provide verification that the system could accurately measure the hydrogen in the vent gas at steady state.

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DOE OSTI2017

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↗

DOE OSTI2018

Gamma-ray raster imaging with robotic data collection. FY18 report

This project has developed an alternative imaging capability for measuring radioactive materials in situ. This approach uses a robotic-mounted gamma-ray detector which can move around an area of interest, sampling the space at a high frequency. By rastering across the area, an image can be created with no collimation and a high efficiency. A regression model was developed to map out material distribution as the robot moves through the environment. An informative path planning algorithm was also developed to maximize the information collected throughout the movement path. The use of a robotic mount allows data collection for long periods of time unattended, and it will also eliminate uncertainties in positioning typically introduced by personnel. This approach will be particularly relevant for gloveboxes, shielded cells, or process piping which may have complex, non-uniform distributions of material.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2019

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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DOE OSTI2019

CST Batch Contact Equilibrium Testing of TCCR Tank 10H Batch 1 and 1A Waste Supernate Samples

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate media to process radioactive waste solutions for the removal of Cs-137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium saltcake solids) within the tank followed by at-tank ion exchange column treatment. Equilibrium batch contact tests conducted in the SRNL Shielded Cells laboratory and associated analyses indicated that both the Tank 10H Batch 1 (pre-production) and 1A (production) waste supernates are unique solutions for which CST cesium removal performance is lower than predicted by ZAM modeling. Cesium distribution coefficients (K d ), percent removal, and CST loading data for the TCCR production supernate batch are provided in Table ES-1. ZAM-based cesium loading predictions are compared to the observed loading for the production batch in Figure ES-1, where significantly lower performance was observed than predicted by the model. Even lower cesium loading results (versus ZAM) were observed previously for the in-tank “teabag” batch contact samples.

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DOE OSTI2019

Determination of Reportable Radionuclides for Defense Waste Processing Facility (DWPF) Sludge Batch 9 (Macrobatch 11)

This report documents analyses and calculations performed by the Savannah River National Laboratory (SRNL) of Tank 40 sludge as part of the Waste Acceptance Product Specifications (WAPS) requirements for Sludge Batch 9 (SB9) for the Defense Waste Processing Facility (DWPF). Tank Farm sample HTF-40-17-5 was received by SRNL Shielded Cells Operations (SCO) and was prepared for radionuclide analyses in addition to various other physical and chemical characterizations. Radionuclide analyses were performed by SRNL Analytical Development (AD). Based on criteria set forth in the Department of Energy (DOE) WAPS and DWPF’s Waste Form Compliance Plan (WCP), as well as, historical lessons from previous sludge batch analyses, a list of radionuclides was developed for which the sample was analyzed. The results of these analyses were then used as input into commercially available software, RadDecay version (v) 5.01. The software was used to calculate the activity of the measured analytes in index years 2017 through 3115 in 100-year increments. Radionuclides with half-lives greater than 10 years and activities that comprised more than 0.01% of the total sludge batch activity are reportable.

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