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DOE OSTI2013

Results For The First Quarter 2013 Tank 50 WAC Slurry Sample: Chemical And Radionuclide Contaminants

This report details the chemical and radionuclide contaminant results for the characterization of the 2013 First Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC). Information from this characterization will be used by Waste Solidification Engineering (WSE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Letter Report: LAW Simulant Development for Cast Stone Screening Test

More than 56 million gallons of radioactive and hazardous waste are stored in 177 underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the wastes and immobilize them in a glass waste form. The WTP includes a pretreatment facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the nonradioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at an offsite federal repository. At least a portion (~35%) of the LAW will be converted to glass in the LAW vitrification facility and will be disposed of onsite at the Integrated Disposal Facility (IDF). The pretreatment and HLW vitrification facilities will have the capacity to treat and immobilize the wastes destined for each facility. However, a second facility will be needed for the expected volume of additional LAW requiring immobilization. A cementitious waste form known as Cast Stone is being considered to provide the required additional LAW immobilization capacity. The Cast Stone waste form must be acceptable for disposal in the IDF. The Cast Stone waste form and immobilization process must be tested to demonstrate that the final Cast Stone waste form can comply with waste acceptance criteria for the IDF disposal facility and that the immobilization processes can be controlled to consistently provide an acceptable waste form product. Further, the waste form must be tested to provide the technical basis for understanding the long term performance of the waste form in the IDF disposal environment. These waste form performance data are needed to support risk assessment and performance assessment (PA) analyses of the long-term environmental impact of the waste disposal in the IDF. A testing program was developed in fiscal year (FY) 2012 describing in some detail the work needed to develop and qualify Cast Stone as a waste form for the solidification of Hanford LAW (Westsik et al. 2012). Included within Westsik et al. (2012) is a section on the near-term needs to address Tri-Party Agreement Milestone M-062-40ZZ. The objectives of the testing program to be conducted in FY 2013 and FY 2014 are to: • Determine an acceptable formulation for the LAW Cast Stone waste form. • Evaluate sources of dry materials for preparing the LAW Cast Stone. • Demonstrate the robustness of the Cast Stone waste form for a range of LAW compositions. • Demonstrate the robustness of the formulation for variability in the Cast Stone process. • Provide Cast Stone contaminant release data for PA and risk assessment evaluations. The first step in determining an acceptable formulation for the LAW Cast Stone waste form is to conduct screening tests to examine expected ranges in pretreated LAW composition, waste stream concentrations, dry-materials sources, and mix ratios of waste feed to dry blend. A statistically designed test matrix will be used to evaluate the effects of these key parameters on the properties of the Cast Stone as it is initially prepared and after curing. The second phase of testing will focus on selection of a baseline Cast Stone formulation for LAW and demonstrating that Cast Stone can meet expected waste form requirements for disposal in the IDF. It is expected that this testing will use the results of the screening tests to define a smaller suite of tests to refine the composition of the baseline Cast Stone formulation (e.g. waste concentration, water to dry mix ratio, waste loading).

Russell, Renee L.↗

DOE OSTI2013

Feasibility of preparing patterned molybdenum coatings on bismuth telluride thermoelectric modules.

Molybdenum electrical interconnects for thermoelectric modules were produced by air plasma spraying a 30%CE%BCm size molybdenum powder through a laser-cut Kapton tape mask. Initial feasibility demonstrations showed that the molybdenum coating exhibited excellent feature and spacing retention (~170%CE%BCm), adhered to bismuth-telluride, and exhibited electrical conductivity appropriate for use as a thermoelectric module interconnect. A design of experiments approach was used to optimize air plasma spray process conditions to produce a molybdenum coating with low electrical resistivity. Finally, a molybdenum coating was successfully produced on a fullscale thermoelectric module. After the addition of a final titanium/gold layer deposited on top of the molybdenum coating, the full scale module exhibited an electrical resistivity of 128%CE%A9, approaching the theoretical resistivity value for the 6mm module leg of 112%CE%A9. Importantly, air plasma sprayed molybdenum did not show significant chemical reaction with bismuth-telluride substrate at the coating/substrate interface. The molybdenum coating microstructure consisted of lamellar splats containing columnar grains. Air plasma sprayed molybdenum embedded deeply (several microns) into the bismuth-telluride substrate, leading to good adhesion between the coating and the substrate. Clusters of round pores (and cracks radiating from the pores) were found immediately beneath the molybdenum coating. These pores are believed to result from tellurium vaporization during the spray process where the molten molybdenum droplets (2623%C2%B0C) transferred their heat of solidification to the substrate at the moment of impact. Substrate cooling during the molybdenum deposition process was recommended to mitigate tellurium vaporization in future studies.

Sarobol, Pylin↗

DOE OSTI2013

Energy-Saving Melting and Revert Reduction Technology (E-SMARRT), Task 5.3: Casting Porosity-Free Grain Refined Magnesium Alloys

The objective of this project was to identify the root causes for micro-porosity in magnesium alloy castings and recommend remedies that can be implemented in production. Three casting magnesium alloys were used to determine the causes of porosity that is a common problem in the magnesium casting industry. These alloys were AM50A, AM60A, and AZ91D, and all were grain refined with C 2 Cl 6 . Gas porosity was found to have little effect if proper degassing methods and foundry practices were used. It was found that sound gating practice, directional solidification towards the riser and generous risering should always dominate in order to produce sound castings. In sand cast magnesium, thermal gradients were found to be an important indicator of heat transfer in the mold. A minimum thermal gradient was found for each alloy that would produce sound castings. There were 2.02°C/cm for AM50A, 1.99°C/cm for AM60A, and 1.97°C/cm for AZ91D. If these gradients are met or exceeded when an area of the casting reaches 50°C below the liquidus for the alloy, the casting will be free of shrinkage porosity in that area. The cooling rates measured during casting magnesium alloys in a pre-heated step permanent mold ranged from 12°F/min in the three inches thick step to 200o F/min in the one inch step. The effect of the cooling rate on the grain size and the mechanical properties was noticeable but not substantial. Grain refining with C 2 Cl 6 had a marked effect on the grain size of the permanent mold step casting. It reduced it by 10-15% in AM50A and by a factor of six in AZ91E from 300 µm to around 55 µm. The mechanical properties of both permanent mold cast AZ91E and AM50A were improved by grain refinement. The most prominent increase was in the elongation of AM50A that more than doubled from 8% to 17%. Computer simulation of porosity was conducted for a well gated commercial wheel casting. The simulation did not predict significant porosity in the wheel. The mechanical testing of tensile specimens excised from the wheel indeed confirmed good properties, in line with the specifications for this alloy. Evidence of minimal micro-porosity was detected in sections cut from the wheel and investigated under optical microscopy. This finding confirms the key role played by utilizing optimal gating and risering practices in minimizing porosity in magnesium castings.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2014

Laboratory Scoping Tests Of Decontamination Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task seeks to examine the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Off-Gas Condensate stream is Technetium-99 ( 99 Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are also expected to be in appreciable concentration in the LAW Off-Gas Condensate are 129 I, 90 Sr, 137 Cs, and {sup 241}Am. This report discusses results of preliminary radionuclide decontamination testing of the simulant. Testing examined use of Monosodium Titanate (MST) to remove 90 Sr and actinides, inorganic reducing agents for 99 Tc, and zeolites for 137 Cs. Test results indicate that excellent removal of 99 Tc was achieved using Sn(II)Cl 2 as a reductant, coupled with sorption onto hydroxyapatite, even in the presence of air and at room temperature. This process was very effective at neutral pH, with a Decontamination Factor (DF) >577 in two hours. It was less effective at alkaline pH. Conversely, removal of the cesium was more effective at alkaline pH, with a DF of 17.9. As anticipated, ammonium ion probably interfered with the Ionsiv®a IE-95 zeolite uptake of 137 Cs. Although this DF of 137 Cs was moderate, additional testing is expected to identify more effective conditions. Similarly, Monosodium Titanate (MST) was more effective at alkaline pH at removing Sr, Pu, and U, with a DF of 319, 11.6, and 10.5, respectively, within 24 hours. Actually, the Ionsiv® IE-95, which was targeting removal of Cs, was also moderately effective for Sr, and highly effective for Pu and U at alkaline pH. The only deleterious effect observed was that the chromium co-precipitates with the {sup 99}Tc during the SnCl 2 reduction. This effect was anticipated, and would have to be considered when managing disposition paths of this stream. Results of this separation testing indicate that sorption/precipitation was a viable concept and has the potential to decontaminate the stream. All radionuclides were at least partially removed by one or more of the materials tested. Based on the results, a possible treatment scenario could involve the use of a reductive precipitation agent (SnCl 2 ) and sorbent at neutral pH to remove the Tc, followed by pH adjustment and the addition of zeolite (Ionsiv® IE-95) to remove the Cs, Sr, and actinides. Addition of MST to remove Sr and actinides may not be needed. Since this was an initial phase of testing, additional tasks to improve separation methods were expected to be identified. Primarily, further testing is needed to identify the conditions for the decontamination process. Once these conditions are established, follow-on tasks likely include evaluation and testing of applicable solid-liquid separation technologies, slurry rheology measurements, composition variability testing and evaluations, corrosion and erosion testing, slurry storage and immobilization investigations, and decontaminated LAW Off-Gas Condensate evaporation and solidification.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Supplemental Immobilization of Hanford Low-Activity Waste: Cast Stone Screening Tests

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Final Scientific/Technical Report for DOE/EERE project Advanced Magnetic Refrigerant Materials

A team led by GE Global Research developed new magnetic refrigerant materials needed to enhance the commercialization potential of residential appliances such as refrigerators and air conditioners based on the magnetocaloric effect (a nonvapor compression cooling cycle). The new magnetic refrigerant materials have potentially better performance at lower cost than existing materials, increasing technology readiness level. The performance target of the new magnetocaloric material was to reduce the magnetic field needed to achieve 4 °C adiabatic temperature change (ΔT adiabatic ) from 1.5 Tesla to 0.75 Tesla. Such a reduction in field minimizes the cost of the magnet assembly needed for a magnetic refrigerator. Such a reduction in magnet assembly cost is crucial to achieving commercialization of magnetic refrigerator technology. This project was organized as an iterative alloy development effort with a parallel material modeling task being performed at George Washington University. Four families of novel magnetocaloric alloys were identified, screened, and assessed for their performance potential in a magnetic refrigeration cycle. Compositions from three of the alloy families were manufactured into regenerator components. At the beginning of the project a previously studied magnetocaloric alloy was selected for manufacturing into the first regenerator component. Each of the regenerators was tested in magnetic refrigerator prototypes at a subcontractor at at GE Appliances. The property targets for operating temperature range, operating temperature control, magnetic field sensitivity, and corrosion resistance were met. The targets for adiabatic temperature change (ΔT adiabatic ) and thermal hysteresis were not met. The high thermal hysteresis also prevented the regenerator components from displaying measurable cooling power when tested in prototype magnetic refrigerators. Magnetic refrigerant alloy compositions that were predicted to have low hysteresis were not attainable with conventional alloy processing methods. Preliminary experiments with rapid solidification methods showed a path towards attaining low hysteresis compositions should this alloy development effort be continued.

36 MATERIALS SCIENCE↗

DOE OSTI2014

Technetium and Iodine Getters to Improve Cast Stone Performance

To determine the effectiveness of the various getter materials prior to their solidification in Cast Stone, a series of batch sorption experiments was performed at Pacific Northwest National Laboratory. To quantify the effectiveness of the removal of Tc(VII) and I(I) from solution by getters, the distribution coefficient, Kd (mL/g), was calculated. Testing involved placing getter material in contact with spiked waste solutions at a 1:100 solid-to-solution ratio for periods up to 45 days with periodic solution sampling. One Tc getter was also tested at a 1:10 solid-to-solution ratio. Two different solution media, 18.2 MΩ deionized water (DI H2O) and a 7.8 M Na LAW simulant, were used in the batch sorption tests. Each test was conducted at room temperature in an anoxic chamber containing N2 with a small amount of H2 (0.7%) to maintain anoxic conditions. Each getter-solution combination was run in duplicate. Three Tc- and I-doping concentrations were used separately in aliquots of both the 18.2 MΩ DI H2O and a 7.8 M Na LAW waste simulant. The 1× concentration was developed based on Hanford Tank Waste Operations Simulator (HTWOS) model runs to support the River Protection Project System Plan Revision 6. The other two concentrations were 5× and 10× of the HTWOS values. The Tc and I tests were run separately (i.e., the solutions did not contain both solutes). Sampling of the solid-solution mixtures occurred nominally after 0.2, 1, 3, 6, 9, 12, 15 days and ~35 to 45 days. Seven getter materials were tested for Tc and five materials were tested for I. The seven Tc getters were blast furnace slag 1 (BFS1) (northwest source), BFS2 (southeast source), Sn(II)-treated apatite, Sn(II) chloride, nano tin phosphate, KMS (a potassium-metal-sulfide), and tin hydroxapatite. The five iodine getters were layered bismuth hydroxide (LBH), argentite mineral, synthetic argentite, silver-treated carbon, and silver-treated zeolite. The Tc Kd values measured from experiments conducted using the 7.8 M Na LAW simulant (the simulant selected to represent LAW) for the first 15 days for four Tc getters (BFS1, BFS2, Sn(II)-treated apatite, and Sn(II) chloride) show no, to a very small, capacity to remove Tc from the LAW simulant. For the Tc-getter experiments in the 7.8 M LAW simulant, the majority of the effluent samples show very small drops in Tc concentrations for the 35-day compared to the 15-day samplings. However, the Tc concentration in the simulant blanks also dropped slightly during this period, so the effect of the getter contacting LAW simulant at 35 days compared to 15 days is minimal; except that the BFS1 1:10 test shows a slow but steady decrease in Tc concentration in the LAW simulant supernatant from the beginning to the 35 day contact at which point about 20% of the original Tc has been removed from solution. Lastly, the KMS getter gives the highest Kd value for Tc at 35 days where Kd values have increased to 104 mL/g. When considering the different I getters reacting with the 7.8 M LAW simulant, two getters are much more effective than the others: Ag zeolite and Syn Arg. The other getters have calculated iodide distribution coefficients that show very limited effectiveness in the caustic conditions created by the LAW simulant. These are preliminary results that will need more detailed analyses including both pre- and post-batch sorption getter solid-phase characterization using state-of-the-art instrumentation such as synchrotron X ray absorption spectroscopy, which can delineate the oxidation state of the Tc and likely iodine species as well as some of the getters key major components, sulfur and iron in the BFS, and tin and sulfur in the tin-bearing and sulfur-bearing getters. This report also describes future experimental studies to be performed to better elucidate the mechanisms controlling the Tc and I sequestration processes in the various getters and leach tests of getter-bearing Cast Stone monoliths.

Tc getters, I getters, Cast Stone, Waste Forms↗

DOE OSTI2014

Alternate Reductant Cold Cap Evaluation Furnace Phase II Testing

Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further research and development of this flowsheet eliminated the formic acid, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric-glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12 th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the CEF cold cap and vapor space data to the benchmark melter flammability models; Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: Define new cold cap reactions and global kinetic parameters in support of the melter flammability model development; Quantify off-gas surging potential of the feed; Characterize off-gas condensate for complete organic and inorganic carbon species. After charging the CEF with cullet from Phase I CEF testing, the melter was slurry-fed with glycolic flowsheet based SB6-Frit 418 melter feed at 36% waste loading and was operated continuously for 25 days. Process data was collected throughout testing and included melter operation parameters and off-gas chemistry. In order to generate off-gas data in support of the flammability model development for the nitric-glycolic flowsheet, vapor space steady state testing in the range of ~300-750°C was conducted under the following conditions, (i) 100% (nominal and excess antifoam levels) and 125% stoichiometry feed and (ii) with and without argon bubbling. Adjustments to feed rate, heater outputs and purge air flow were necessary in order to achieve vapor space temperatures in this range. Surge testing was also completed under nominal conditions for four days with argon bubbling and one day without argon bubbling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Creep of A508/533 Pressure Vessel Steel

ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are allowed by Code Case N-499-2 (now incorporated as an appendix to Section III Division 5 of the Code). This Code Case was developed with a rather sparse data set and focused primarily on rolled plate material (A533 specification). Confirmatory tests of creep behavior of both A508 and A533 are described here that are designed to extend the database in order to build higher confidence in ensuring the structural integrity of the VHTR RPV during off-normal conditions. A number of creep-rupture tests were carried out at temperatures above the 371°C (700°F) Code limit; longer term tests designed to evaluate minimum creep behavior are ongoing. A limited amount of rupture testing was also carried out on welded material. All of the rupture data from the current experiments is compared to historical values from the testing carried out to develop Code Case N-499-2. It is shown that the A508/533 basemetal tested here fits well with the rupture behavior reported from the historical testing. The presence of weldments significantly reduces the time to rupture. The primary purpose of this report is to summarize and record the experimental results in a single document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Results for the Fourth Quarter 2012 Tank 50 WAC Slurry Sample: Chemical and Radionuclide Contaminants

This report details the chemical and radionuclide contaminant results for the characterization of the 2012 Fourth Quarter sampling of Tank 50 for the Saltstone Waste Acceptance Criteria (WAC).1 Information from this characterization will be used by Waste Solidification Engineering (WSE) to support the transfer of low-level aqueous waste from Tank 50 to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50 Waste Characterization System. The following conclusions are drawn from the analytical results provided in this report: The concentration of the reported chemical and radioactive contaminants were less than their respective WAC Limits and Targets, unless noted in this section. Norpar 13 and Isopar L have higher detection limits 5 compared with the Saltstone WAC 1 . The data provided in this report is based upon the concentrations in the sub-sample, and due to the limited solubility in aqueous solution, may not represent the concentrations of the analytes in Tank 50. Diisooctyl adipate (or diisooctyl hexanedioate) was measured at 1.30E+00 mg/L in one of two replicate measurements conducted on an at-depth sample. a The organic analysis of the at-depth sample was conducted at the request of SRR. 4 This analyte was below the detection limit in the surface sample. The low insoluble solids content increases the measurement uncertainty for insoluble species.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Radionuclide Migration through Sediment and Concrete: 16 Years of Investigations

The Waste Management Project provides safe, compliant, and cost-effective waste management services for the Hanford Site and the U.S. Department of Energy (DOE) complex. Part of these services includes safe disposal of low-level waste and mixed low-level waste at the Hanford Low-Level Waste Burial Grounds in accordance with the requirements of DOE Order 435.1, Radioactive Waste Management. To partially satisfy these requirements, performance assessment analyses were completed and approved. DOE Order 435.1 also requires continuing data collection to increase confidence in the critical assumptions used in these analyses to characterize the operational features of the disposal facility that are relied on to satisfy the performance objectives identified in the order. Cement-based solidification and stabilization is considered for hazardous waste disposal because it is easily done and cost-efficient. One critical assumption is that concrete will be used as a waste form or container material at the Hanford Site to control and minimize the release of radionuclide constituents in waste into the surrounding environment. Concrete encasement would contain and isolate the waste packages from the hydrologic environment and act as an intrusion barrier. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. The radionuclides iodine-129, selenium-75, technetium-99, and uranium-238 have been identified as long-term dose contributors (Mann et al. 2001; Wood et al. 1995). Because of their anionic nature in aqueous solutions, these constituents of potential concern may be released from the encased concrete by mass flow and/or diffusion and migrate into the surrounding subsurface environment (Serne et al. 1989; 1992; 1993a, b; 1995). Therefore, it is necessary to assess the performance of the concrete encasement structure and the ability of the surrounding soil to retard radionuclide migration. Each of the test methods performed throughout the lifetime of the project has focused on different aspects of the concrete waste form weathering process. Diffusion of different analytes [technetium-99 (Tc-99), iodine-125 (I-125), stable iodine (I), uranium (U), and rhenium (Re)] has been quantified from experiments under both saturated and unsaturated conditions. The water-saturated conditions provide a conservative estimate of the concrete’s performance in situ, and the unsaturated conditions provide a more accurate estimate of the diffusion of contaminants from the concrete.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Technetium and Iodine Getters to Improve Cast Stone Performance

To determine the effectiveness of the various getter materials prior to their solidification in Cast Stone, a series of batch sorption experiments was performed at Pacific Northwest National Laboratory. To quantify the effectiveness of the removal of Tc(VII) and I(I) from solution by getters, the distribution coefficient, K d (mL/g), was calculated. Testing involved placing getter material in contact with spiked waste solutions at a 1:100 solid-to-solution ratio for periods up to 45 days with periodic solution sampling. One Tc getter was also tested at a 1:10 solid-to-solution ratio. Two different solution media, 18.2 MΩ deionized water (DI H 2 O) and a 7.8 M Na LAW simulant, were used in the batch sorption tests. Each test was conducted at room temperature in an anoxic chamber containing N2 with a small amount of H 2 (0.7%) to maintain anoxic conditions. Each getter-solution combination was run in duplicate. Three Tc- and I-doping concentrations were used separately in aliquots of both the 18.2 MΩ DI H2O and a 7.8 M Na LAW waste simulant. The 1× concentration was developed based on Hanford Tank Waste Operations Simulator (HTWOS) model runs to support the River Protection Project System Plan Revision 6. The other two concentrations were 5× and 10× of the HTWOS values. The Tc and I tests were run separately (i.e., the solutions did not contain both solutes). Sampling of the solid-solution mixtures occurred nominally after 0.2, 1, 3, 6, 9, 12, 15 days and ~35 to 45 days. Seven getter materials were tested for Tc and five materials were tested for I. The seven Tc getters were blast furnace slag 1 (BFS1) (northwest source), BFS2 (southeast source), Sn(II)-treated apatite, Sn(II) chloride, nano tin phosphate, KMS (a potassium-metal-sulfide), and tin hydroxapatite. The five iodine getters were layered bismuth hydroxide (LBH), argentite mineral, synthetic argentite, silver-treated carbon, and silver-treated zeolite. The Tc Kd values measured from experiments conducted using the 7.8 M Na LAW simulant (the simulant selected to represent LAW) for the first 15 days for four Tc getters (BFS1, BFS2, Sn(II)-treated apatite, and Sn(II) chloride) show no, to a very small, capacity to remove Tc from the LAW simulant. For the Tc-getter experiments in the 7.8 M LAW simulant, the majority of the effluent samples show very small drops in Tc concentrations for the 35-day compared to the 15-day samplings. However, the Tc concentration in the simulant blanks also dropped slightly during this period, so the effect of the getter contacting LAW simulant at 35 days compared to 15 days is minimal; except that the BFS1 1:10 test shows a slow but steady decrease in Tc concentration in the LAW simulant supernatant from the beginning to the 35 day contact at which point about 20% of the original Tc has been removed from solution. Lastly, the KMS getter gives the highest K d value for Tc at 35 days where K d values have increased to 104 mL/g. When considering the different I getters reacting with the 7.8 M LAW simulant, two getters are much more effective than the others: Ag zeolite and Syn Arg. The other getters have calculated iodide distribution coefficients that show very limited effectiveness in the caustic conditions created by the LAW simulant. These are preliminary results that will need more detailed analyses including both pre- and post-batch sorption getter solid-phase characterization using state-of-the-art instrumentation such as synchrotron X-ray absorption spectroscopy, which can delineate the oxidation state of the Tc and likely iodine species as well as some of the getters key major components, sulfur and iron in the BFS, and tin and sulfur in the tin-bearing and sulfur-bearing getters. This report also describes future experimental studies to be performed to better elucidate the mechanisms controlling the Tc and I sequestration processes in the various getters and leach tests of getter-bearing Cast Stone monoliths.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Fully-Implicit Orthogonal Reconstructed Discontinuous Petrov-Galerkin Method for Multiphysics Problems

This document is focussed on the development of reconstructed discontinuous Galerkin (RDG P n P m ) method, based on orthogonal basis functions. Orthogonality of basis functions is essential for enabling robust and efficient fully-implicit Newton-Krylov based time discretization. The method is designed for generic partial differential equations, including transient hyperbolic, parabolic and/or elliptic operators, which are attributed to many multiphysics problems. Our driving application of interest is additive manufacturing, which requires simulations of fluid-solid-gas systems, with compressibility effects and phase change (melting/ solidification). Thus, we demonstrate the method capabilities for solving nonlinear heat conduction equation coupled with multi-material Navier-Stokes based compressible fluid (in low Mach number limit) systems. We focus our attention to the method accuracy (in both time and space), as well as solvability of linear algebra involved in linear steps of the Newton-based solver. We demonstrate the accuracy and efficacy of the method, emphasizing the advantages from the orthogonality of basis functions, which makes better conditioning of underlying (approximate) Jacobian matrices, and rapid convergence of Krylov methods. iii

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2015

Projection Microstereolithography for Architected Materials

Stereolithography (SL) is a prominent paradigm in the realm of additive manufacturing (AM) technologies, which refers to fabrication approaches that build “bottom up” by adding material. Pioneered in 1984 by Chuck Hull, SL relies on light-activated solidification of a liquid photosensitive resin in a layer-by-layer fashion, with the geometry of each layer defined by sectioning a three-dimensional (3D) computer model by closely-spaced parallel planes. The maturation of the technology over nearly three subsequent decades has resulted in the wide availability of commercial SL products, such as the ProJet series from 3D Systems, Inc. (which uses the proprietary acronym SLA to refer to SL technology), which routinely produces parts with 50-100 µm resolution in all three axes, and overall dimensions of 0.25-0.5 m. In the most general terms, advancements in SL technology have generally been aimed at achieving four goals: decreasing (1) system cost and (2) minimum feature size (i.e. improving resolution), while (3) increasing overall part size and (4) build speed. SL systems can be broadly categorized into two classes, by the way in which they deliver illumination to produce the shape of each object layer to be fabricated: laser-scanning and mask projection approaches. Commercially-available SL systems incorporate the laser-scanning approach, following the original scheme described by Hull, since digital image projection technology has only emerged relatively recently as a viable alternative. A significant limitation of laser scanning is the relatively slow speed of rastering the laser to form each layer’s pattern, highlighting the need for a dynamic mask-based approach to polymerize larger areas in a single exposure. The recent proliferation of digital display technologies (referred to as spatial light modulators or SLMs) has enabled the implementation of such “dynamic masking,” increasing system throughput by allowing a complete image layer to be cross-linked simultaneously. These mask projection SL systems are gaining greater prominence, and the landscape of some of the parameters available for system design, as implemented in recent work by a variety of investigators, has been reviewed by Lambert et al.

36 MATERIALS SCIENCE↗

DOE OSTI2015

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Heat of Hydration of Low Activity Cementitious Waste Forms

During the curing of secondary waste grout, the hydraulic materials in the dry mix react exothermally with the water in the secondary low-activity waste (LAW). The heat released, called the heat of hydration, can be measured using a TAM Air Isothermal Calorimeter. By holding temperature constant in the instrument, the heat of hydration during the curing process can be determined. This will provide information that can be used in the design of a waste solidification facility. At the Savannah River National Laboratory (SRNL), the heat of hydration and other physical properties are being collected on grout prepared using three simulants of liquid secondary waste generated at the Hanford Site. From this study it was found that both the simulant and dry mix each had an effect on the heat of hydration. It was also concluded that the higher the cement content in the dry materials mix, the greater the heat of hydration during the curing of grout.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Evaluation of Molybdenum as a Surrogate for Iridium in the GPHS Weld Development

The welding equipment used for welding iridium containers (clads) at Los Alamos National Laboratory is twenty five years old and is undergoing an upgrade. With the upgrade, there is a requirement for requalification of the welding process, and the opportunity for process improvement. Testing of the new system and requalification will require several welds on iridium test parts and clads, and any efforts to improve the process will add to the need for iridium parts. The extreme high cost of iridium imposes a severe limitation on the extent of test welding that can be done. The 2 inch diameter, 0.027 inch thick, iridium blank disc that the clad cup is formed from, is useful for initial weld trials, but it costs $5000. The development clad sets needed for final tests and requalification cost $15,000 per set. A solution to iridium cost issue would be to do the majority of the weld development on a less expensive surrogate metal with similar weld characteristics. One such metal is molybdenum. Since its melting index (melting temperature x thermal conductivity) is closest to iridium, welds on molybdenum should be similar in size for a given weld power level. Molybdenum is inexpensive; a single 2 inch molybdenum disc costs only $9. In order to evaluate molybdenum as a surrogate for iridium, GTA welds were first developed to provide full penetration on 0.030 inch thick molybdenum discs at speeds of 20, 25, and 30 inches per minute (ipm). These weld parameters were then repeated on the standard 0.027 inch thick iridium blanks. The top surface and bottom surface (root) width and grain structure of the molybdenum and iridium welds were compared, and similarities were evident between the two metals. Due to material and thickness differences, the iridium welds were approximately 35% wider than the molybdenum welds. A reduction in iridium weld current of 35% produce welds slightly smaller than the molybdenum welds yet showed that current could be scaled according to molybdenum/iridium weld width ratio to achieve similar welds. Further weld trials using various thicknesses of molybdenum determined that 0.024 inch thick molybdenum material would best match the 0.027 inch thick iridium in achieving comparable welds when using the same welding parameters. Across the range of welding speeds, the characteristic weld pool shape and solidification grain structure in the two materials was also similar. With the similarity of welding characteristics confirmed, and the appropriate thickness of molybdenum determined, it has been concluded that the use of molybdenum discs and tube sections will greatly expand the weld testing opportunities prior to iridium weld qualification

Engineering(42)↗