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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Additive Manufacturing at LANL: Advanced Characterization to Explore the Science of a New Manufacturing Method

Additive manufacturing (AM), or three-dimensional (3D) printing as it is more commonly known, is defined as the process of joining materials and creating objects by melting, sintering, or fusing material in a layer-by-layer fashion coordinated via 3D model data.1 Subtractive, or traditional, manufacturing methodologies often consist of machining/removing material—like a sculptor—or forming material through the application of pressure—like a potter. Conversely, in an AM process, material is added in individual volume elements and built up in a way similar to interlocking building blocks, but with volume elements that are typically the size of a grain of sand. The additive process often involves less waste when compared to subtractive techniques because material is only added when and where it is needed. Adjustments to the final structure are relatively straightforward and can be simply achieved by adjusting the 3D computer model. This makes the technology much more flexible than traditional, subtractive techniques where new tooling or forming equipment is usually needed to accommodate design changes. Also, the AM processes are beneficial because they permit the fabrication of unique geometries, such as miniaturized metal lattice structures, that cannot be achieved using traditional techniques. An example of a metal lattice structure is the Eiffel Tower with its Figure 1. Schematic showing required linkages between experimental and modeling thrusts in order to achieve science-based qualification. Arrows showing linkages are color-coded according to the funded projects listed. geometric, interconnecting struts that reduce the overall weight of the tower while maintaining strength. In AM, the size of the struts can be made smaller than the diameter of a human hair, which further reduces weight while maintaining strength—a combination of properties that can benefit many applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Preliminary Characterization and Evaluation on ShAPE Manufactured 316H and ODS Steels

This study provides the first- of- a- kind results of direct tube formation through shear assisted processing and extrusion (ShAPE) for oxide dispersion strengthened (ODS) steel material; previously only bar was successfully made. The Advanced Materials and Manufacturing Technology (AMMT) program develops cross-cutting technologies in support of a broad range of nuclear reactor technologies and maintains U.S. leadership in materials and manufacturing technologies for nuclear energy applications. The overarching vision of AMMT is to accelerate the development, qualification, demonstration, and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy. Solid-state advanced manufacturing techniques can overcome some of the challenges in liquid-based additive manufacturing processes and should therefore be considered in material design and manufacturing as well. The work presented in this report forms part of a study on solid-state additive manufacturing techniques of 316 stainless steels and ODS steel components and supports the vision and goals of the AMMT program relevant to accelerate the development and deployment of advanced manufacturing processes. Achieving this can provide a safety improvement through larger safety margins, economic benefit for higher efficiency during operation, and a cost reduction through more effective manufacturing processes and less waste.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Preliminary Characterization and Evaluation on FFF Manufactured 316H and ODS Steels

The Advanced Materials and Manufacturing Technology (AMMT) program develops cross-cutting technologies in support of a broad range of nuclear reactor technologies and maintains U.S. leadership in materials and manufacturing technologies for nuclear energy applications. The overarching vision of AMMT is to accelerate the development, qualification, demonstration, and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy. Solid-state advanced manufacturing techniques can overcome some of the challenges in liquid-based additive manufacturing (AM) processes and should therefore be considered in material design and manufacturing as well. The work presented in this report forms part of a study on solid-state AM techniques of 316 stainless steels (SS) and oxide dispersion strengthened (ODS) steel components and supports the vision and goals of the AMMT program relevant to accelerate the development and deployment of advanced manufacturing processes. Achieving this can provide a safety improvement through larger safety margins, economic benefit for higher efficiency during operation, and a cost reduction through more effective manufacturing processes and less waste. This study provides preliminary information on development of the fused-filament fabrication (FFF) process using different powder types to demonstrate the sensitivity, and therefore the characterization of these sample components. The full solid-state manufacturing feasibility study will be completed and reported in a final feasibility evaluation during 2025. The study investigation used two 14YWT ODS powder batches, which provided information and the effect of different powder morphologies on manufacturability. The two 316SS powders demonstrated the effect of powder size on the manufacturability using FFF. Two product forms, namely a honeycomb structure and flat samples, were manufactured to demonstrate the flexibility of product form.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Functions and Requirements of the Radioactive Waste Test Platform

The purpose of this document is to provide the function and requirements for the Radioactive Waste Test Platform (hereafter called “test platform”). Historically, testing has been performed at the bench scale with actual waste samples (Fiskum et al. 2006a, b, 2008, 2009; Lumetta et al. 2009; Edwards et al. 2009; Smith et al. 2000, 2001). These tests form the basis for waste treatment operations at the Hanford Site. However, since 2008, there has been limited capability to perform bench-scale tests with actual waste samples. Prior work has shown that simulants can provide useful information about the performance of unit operations; however, there have been examples in the past where simulants did not adequately capture the expected performance of the tank waste materials. Therefore, as the Hanford Site moves towards initiating treatment of tank waste, it is important to have a platform available that can evaluate process options utilizing tank waste samples. This platform can be utilized to evaluate processing options and to address troubleshooting for unexpected system performance. The initial operations for the test platform will be to contribute towards maturing the Direct Feed Low-Activity Waste (DFLAW) feed qualification program and to demonstrate an option for disposition of the melter offgas stream. This equipment will be the ideal platform to evaluate additional process options including filtration operating conditions associated with precipitation, methods to extend resin lifetime associated with multi-cycle degradation, and evaluation of other approaches to minimize the impact of recycle streams from the effluent management system (including Tc) on melter performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Experience in Remote Demolition of the Activated Biological Shielding of the Multi Purpose Research Reactor (MZFR) on the German Karlsruhe Site - 12208

In 2009, WAK Decommissioning and Waste Management GmbH (WAK) became owner and operator of the waste treatment facilities of Karlsruhe Institute of Technology (KIT) as well as of the prototype reactors, the Compact Sodium-Cooled Fast Reactor (KNK) and Multi-Purpose Reactor (MZFR), both being in an advanced stage of dismantling. Together with the dismantling and decontamination activities of the former WAK reprocessing facility since 1990, the envisaged demolishing of the R and D reactor FR2 and a hot cell facility, all governmentally funded nuclear decommissioning projects on the Karlsruhe site are concentrated under the WAK management. The small space typical of prototype research reactors represented a challenge also during the last phase of activated dismantling, dismantling of the activated biological shield of the MZFR. Successful demolition of the biological shield required detailed planning and extensive testing in the years before. In view of the limited space and the ambient dose rate that was too high for manual work, it was required to find a tool carrier system to take up and control various demolition and dismantling tools in a remote manner. The strategy formulated in the concept of dismantling the biological shield by means of a modified electro-hydraulic demolition excavator in an adaptable working scaffolding turned out to be feasible. The following boundary conditions were essential: - Remote exchange of the dismantling and removal tools in smallest space. - Positioning of various supply facilities on the working platform. - Avoiding of interfering edges. - Optimization of mass flow (removal of the dismantled mass from the working area). - Maintenance in the surroundings of the dismantling area (in the controlled area). - Testing and qualification of the facilities and training of the staff. Both the dismantling technique chosen and the proceeding selected proved to be successful. Using various designs of universal cutters developed on the basis of wall saws, both the activated steel liner and the inner reinforcing layer were cut remotely in one process. This allowed for the efficient execution of the following remote concrete removal steps using mining techniques. The electro-hydraulic demolition excavator that was purchased and then modified turned out to be an ideal tool carrier system with rapid-exchange coupling. Due to the high availability, no major delays occurred. This also was a result of the consistently implemented maintenance and repair concept. With the excavator installed in a modifiable scaffolding suspended from a rotating carrier ring, all dismantling areas could be reached and treated in spite of the small space. Thanks to an optimum organization of work-flows, routine change of dismantling work, and maintenance or repair, the iterative radiological measurement campaigns could be integrated in the whole activity without the dismantling work being disturbed significantly. The ventilation system with pressure grading and pre-filtration units ensured a low contamination level in the dismantling area. It was also possible to manage the dust formed by the milling of concrete surfaces. As it was possible to further cut metal parts and crushed concrete later on, residue flows were optimized. The planned overall period for testing, dismantling the bio-shield and removing the equipment was 36 months. The final duration was 39 months. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Advanced Fuels Campaign Execution Plan

The Advanced Fuels Campaign (AFC) Execution Plan details the strategy, mission, scope, and goals—both near-term and long-term—along with the structure and organization of nuclear fuels and materials research, development, and demonstration (RD&D) activities within the Fuel Cycle Technologies (FCT) program. The FCT program, tasked by the U.S. Department of Energy (DOE), employs a science-based approach to advance fuel technologies. This approach integrates theory, experiments, and multi-scale modeling and simulation (M&S) to develop a predictive understanding of fuel fabrication processes and fuel/cladding performance under irradiation, moving beyond traditional empirical methods. The long-term goals of the AFC are guided by the AFC Strategic Plan and align with the DOE Office of Nuclear Energy (NE) Roadmap [1], which outlines a multi-decade vision for demonstrating and qualifying advanced fuel forms to support diverse fuel cycle options. Near-term goals focus on enhancing accident tolerant fuels (ATF) for Light Water Reactors (LWR), a significant challenge that demands balancing immediate objectives with ongoing progress toward advanced reactor missions. Accelerating the traditional fuel qualification process to meet ATF objectives is another critical challenge. A detailed set of 5-year goals, summarized below, has been developed in line with the overarching science-based fuel development approach: • Advanced LWR Fuel Technologies: By 2027, support the development of advanced LWR fuel technologies with improved performance and enhanced accident tolerance. This includes high burnup (HBu), low enriched uranium (LEU)+, coated cladding, and doped fuel, aimed at complementing industry-led significant LWR uprates and plant refurbishments. • Tristructural Isotropic (TRISO) Fuel: Achieve qualification by 2028 and develop improved designs for emerging markets. • Metal Fuel: Achieve qualification by 2028 and develop improved designs for emerging markets. • Molten Salt Fuel: By 2027, deploy a robust program that enables fuel salt qualification technologies needed to support fuel salt research and development (R&D), focusing on emergent needs to derisk fuel salt production and utilization in advanced reactors. • Long-Term ATF: Develop fuel technologies that enable significant power uprates (~50%) in refurbished or new LWRs while optimizing fissile material utilization and waste disposal. The 5-year milestones in the AFC Execution Plan are contingent on an assumed budget. This Execution Plan will be updated annually to reflect actual funding profiles as budget guidance becomes available, ensuring milestones are adjusted accordingly. In summary, the AFC Execution Plan presents a comprehensive strategy to advance nuclear fuel technologies through a science-based approach, addressing both near-term and long-term goals while adapting to funding realities.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

DFLAW Glass and Feed Qualifications for AP-107 to Support WTP Start-Up and Flow-Sheet Development (VSL-18R4500-1, Rev 0)

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at The United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility on the Hanford site while the IHLW product is designed for acceptance into the national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. The Low Activity Waste (LAW) at Hanford will be vitrified in a Joule Heated Ceramic Melter (JHCM) to produce a stable product for safe disposal. Although present in relatively small amounts, a few key constituents in the LAW streams play particularly important roles. Sulfur, chlorine, and fluorine can have significant impacts on LAW glass formulation and waste loadings, while technetium and iodine are important for waste form performance, with technetium being the constituent of most concern with respect to waste disposal in the Integrated Disposal Facility (IDF). The IDF is designed to accept ILAW, secondary waste associated with ILAW production, and other on-site Hanford Non-Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) waste. As a result of volatilization of these species during the vitrification process, they are also present in significant amounts in the liquid effluents from the off-gas treatment system. Understanding the fraction of each constituent incorporated in the LAW glass or in WTP liquid effluents is critical for assessing the performance of waste forms to be disposed in the IDF. Liquid effluents from the LAW vitrification process that are not recycled will likely be solidified to produce a secondary waste form that is not as chemically durable as glass. Therefore, understanding partitioning of the key constituents between the glass and non-glass waste forms is a significant factor in the Performance Assessment (PA) for the Hanford IDF. Clearly, it is preferable to incorporate as much of the technetium into the glass as possible because of the superior chemical durability of that waste form. The objective of the present work is to improve the understanding of the partitioning of technetium and other key waste constituents between the glass and non-glass waste forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

MODELING THE IMPACT OF ELEVATED MERCURY IN DEFENSE WASTE PROCESSING FACILITY MELTER FEED ON THE MELTER OFF-GAS SYSTEM-PRELIMINARY REPORT

The Defense Waste Processing Facility (DWPF) is currently evaluating an alternative Chemical Process Cell (CPC) flowsheet to increase throughput. It includes removal of the steam-stripping step, which would significantly reduce the CPC processing time and lessen the sampling needs. However, its downside would be to send 100% of the mercury that comes in with the sludge straight to the melter. For example, the new mercury content in the Sludge Batch 5 (SB5) melter feed is projected to be 25 times higher than that in the SB4 with nominal steam stripping of mercury. This task was initiated to study the impact of the worst-case scenario of zero-mercury-removal in the CPC on the DWPF melter offgas system. It is stressed that this study is intended to be scoping in nature, so the results presented in this report are preliminary. In order to study the impact of elevated mercury levels in the feed, it is necessary to be able to predict how mercury would speciate in the melter exhaust under varying melter operating conditions. A homogeneous gas-phase oxidation model of mercury by chloride was developed to do just that. The model contains two critical parameters pertaining to the partitioning of chloride among HCl, Cl, Cl{sub 2}, and chloride salts in the melter vapor space. The values for these parameters were determined at two different melter vapor space temperatures by matching the calculated molar ratio of HgCl (or Hg{sub 2}Cl{sub 2}) to HgCl{sub 2} with those measured during the Experimental-Scale Ceramic Melter (ESCM) tests run at the Pacific Northwest National Laboratory (PNNL). The calibrated model was then applied to the SB5 simulant used in the earlier flowsheet study with an assumed mercury stripping efficiency of zero; the molar ratio of Cl-to-Hg in the resulting melter feed was only 0.4, compared to 12 for the ESCM feeds. The results of the model run at the indicated melter vapor space temperature of 650 C (TI4085D) showed that due to excessive shortage of chloride, only 6% of the mercury fed is expected to get oxidized, mostly as HgCl, while the remaining mercury would exist either as elemental mercury vapor (90%) or HgO (4%). Noting that the measured chloride level in the SB5 qualification sample was an order of magnitude lower than that used in the SB5 simulant, the degree of chloride shortage will be even greater. As a result, the projected level of HgCl in the actual SB5 melter exhaust will be even lower than 6% of the total mercury fed, while that of elemental mercury is likely to be greater than 90%. The homogeneous oxidation of mercury in the off-gas was deemed to be of primary importance based on the postulation that mercury and other volatile salts form submicron sized aerosols upon condensation and thus remain largely in the gas stream downstream of the quencher where they can deposit in the off-gas lines, Steam-Atomized Scrubbers (SAS), and High-Efficiency Mist Eliminator (HEME). Formation of these submicron semi-volatile salts in the condensate liquid is considered to be unlikely, so the liquid phase reactions were considered to be less important. However, subsequent oxidation of mercury in the liquid phase in the off-gas system was examined in a simplified model of the off-gas condensate. It was found that the condensate chemistry was consistent with further oxidation of elemental mercury to Hg{sub 2}Cl{sub 2} and conversion of HgO to chlorides. The results were consistent with the available experimental data. It should also be noted that the model predictions presented in this report do not include any physically entrained solids, which typically account for much of the off-gas carryover on a mass basis. The high elemental mercury vapor content predicted at the DWPF Quencher inlet means that physically entrained solids could provide the necessary surface onto which elemental mercury vapor could condense, thereby coating the solids as well as the internal surfaces of the off-gas system with mercury. Clearly, there are many process benefits to be gained by removing the steam-stripping step from the CPC cycle. The goal of this task was to study what adverse impact the zero-mercury-removal scenario would have on the DWPF melter off-gas system operation. It is stressed again that this study was intended to be scoping in nature, so the results presented in this report are preliminary. Any further substantiation of these results for actual implementation into the DWPF flowsheet would require an in-depth modeling study of all three reaction zones, including the aqueous-phase reactions in the quencher, OGCT, Steam Atomized Scrubber (SAS), and off-gas condenser with recirculated condensate, and the proof-of-principle experiments.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Knowledge Management Initiatives Used to Maintain Regulatory Expertise in Transportation and Storage of Radioactive Materials - 12177

The U.S. Nuclear Regulatory Commission (NRC) was established in 1974 with the mission to license and regulate the civilian use of nuclear materials for commercial, industrial, academic, and medical uses in order to protect public health and safety, and the environment, and promote the common defense and security. Currently, approximately half (∼49%) of the workforce at the NRC has been with the Agency for less than six years. As part of the Agency's mission, the NRC has partial responsibility for the oversight of the transportation and storage of radioactive materials. The NRC has experienced a significant level of expertise leaving the Agency due to staff attrition. Factors that contribute to this attrition include retirement of the experienced nuclear workforce and mobility of staff within or outside the Agency. Several knowledge management (KM) initiatives have been implemented within the Agency, with one of them including the formation of a Division of Spent Fuel Storage and Transportation (SFST) KM team. The team, which was formed in the fall of 2008, facilitates capturing, transferring, and documenting regulatory knowledge for staff to effectively perform their safety oversight of transportation and storage of radioactive materials, regulated under Title 10 of the Code of Federal Regulations (10 CFR) Part 71 and Part 72. In terms of KM, the SFST goal is to share critical information among the staff to reduce the impact from staff's mobility and attrition. KM strategies in place to achieve this goal are: (1) development of communities of practice (CoP) (SFST Qualification Journal and the Packaging and Storing Radioactive Material) in the on-line NRC Knowledge Center (NKC); (2) implementation of a SFST seminar program where the seminars are recorded and placed in the Agency's repository, Agency-wide Documents Access and Management System (ADAMS); (3) meeting of technical discipline group programs to share knowledge within specialty areas; (4) development of written guidance to capture 'administrative and technical' knowledge (e.g., office instructions (OIs), generic communications (e.g., bulletins, generic letters, regulatory issue summary), standard review plans (SRPs), interim staff guidance (ISGs)); (5) use of mentoring strategies for experienced staff to train new staff members; (6) use of Microsoft SharePoint portals in capturing, transferring, and documenting knowledge for staff across the Division from Division management and administrative assistants to the project managers, inspectors, and technical reviewers; and (7) development and implementation of a Division KM Plan. A discussion and description of the successes and challenges of implementing these KM strategies at the NRC/SFST will be provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

SLUDGE BATCH VARIABILITY STUDY WITH FRIT 418

The Defense Waste Processing Facility (DWPF) initiated processing Sludge Batch 6 (SB6) in the summer of 2010. In support of processing, the Savannah River National Laboratory (SRNL) provided a recommendation to utilize Frit 418 to process SB6. This recommendation was based on assessments of the compositional projections for SB6 available at the time from the Liquid Waste Organization (LWO) and SRNL (using a model-based approach). To support qualification of SB6, SRNL executed a variability study to assess the applicability of the current durability models for SB6. The durability models were assessed over the expected Frit 418-SB6 composition range. Seventeen glasses were selected for the variability study based on the sludge projections used in the frit recommendation. Five of the glasses are based on the centroid of the compositional region, spanning a waste loading (WL) range of 32 to 40%. The remaining twelve glasses are extreme vertices (EVs) of the sludge region of interest for SB6 combined with Frit 418 and are all at 36% WL. These glasses were fabricated and characterized using chemical composition analysis, X-ray diffraction (XRD) and the Product Consistency Test (PCT). After initiating the SB6 variability study, the measured composition of the SB6 Tank 51 qualification glass produced at the SRNL Shielded Cells Facility indicated that thorium was present in the glass at an appreciable concentration (1.03 wt%), which made it a reportable element for SB6. This concentration of ThO{sub 2} resulted in a second phase of experimental studies. Five glasses were formulated that were based on the centroid of the new sludge compositional region combined with Frit 418, spanning a WL range of 32 to 40%. These glasses were fabricated and characterized using chemical composition analysis and the PCT. Based on the measured PCT response, all of the glasses (with and without thorium) were acceptable with respect to the Environmental Assessment (EA) reference glass regardless of thermal history. All of the normalized boron releases were less than 1 g/L. While all of the targeted glass compositions were predictable with respect to the Product Composition Control System (PCCS) models for durability, a small number of the measured glass compositions were located outside of the lower prediction limit indicating poorer durability than what was actually measured. These unpredictable glasses were in the same lithium metaborate (LM) preparation block during the chemical analyses, which resulted in measured compositions that were not representative of the target compositions. A review of the data did not indicate a clear cause for the problem. Re-digestion and re-measurement of three glasses from this preparation block yielded glass compositions closer to the target values and predicted PCT responses within the PCCS model uncertainty. Therefore, it is believed that the glasses were correctly fabricated and the targeted compositions are closer representations of the true compositions. Per the requirements of the DWPF Glass Product Control Program, the PCCS durability models have been shown to be applicable for the SB6/Frit 418 glass system. PCT results from the glasses fabricated as part of the variability study were shown to be predictable and/or acceptable with respect to the DWPF PCCS models. In addition, the inclusion of ThO{sub 2} was shown to have minimal impact on the acceptability and predictability of the variability study glasses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

THE SLUDGE BATCH 7A GLASS VARIABILITY STUDY WITH FRIT 418 AND FRIT 702

The Defense Waste Processing Facility (DWPF) is preparing to initiate processing of Sludge Batch 7a (SB7a) in May 2011. To support qualification of SB7a, the Savannah River National Laboratory (SRNL) was requested to execute a variability study (VS) to assess the applicability of the current Product Composition Control System (PCCS) durability models for the Frit 418-SB7a compositional region of interest. The objective of this study was to demonstrate applicability of the current durability models to the SB7a compositional region of interest and acceptability of the SB7a glasses with respect to the Environmental Assessment (EA) glass in terms of durability as defined by the Product Consistency Test (PCT). To support programmatic objectives, twenty-eight SB7a glasses were selected based on the nominal sludge projections used to support the frit recommendation. Twenty-three of the SB7a VS glasses were based on the use of Frit 418, while 5 glasses were based on the use of Frit 702. Frit 702 was also identified as a viable candidate for SB7a, especially if SO{sub 4} concentrations are found to be higher than anticipated. Frit 702 has shown a higher SO{sub 4} retention capability as compared to Frit 418. With respect to acceptability, the PCT results of the SB7a-VS glasses are acceptable relative to the EA glass regardless of thermal history (quenched or canister centerline cooled) or compositional view (target or measured). More specifically, all of the SB7a glasses have normalized boron release values (NL [B]) less than 0.9 g/L as compared to the benchmark NL [B] value for EA of 16.695 g/L. With respect to the applicability of the current durability models to the SB7a VS compositional region of interest, all of the study glasses (based on target compositions) lie within the 95% confidence intervals of the model predictions. When model applicability is based on the measured compositions, all of the SB7a VS glasses are predictable with the exception of SB7aVS-02 and SB7aVS-06. Although the NL [B] values of these two glasses range from 0.66 to 0.73 g/L (considered very acceptable), the PCT responses are not considered predictable by the current durability models. The current durability models are conservative for these glasses since they are more durable than predicted by the models. These two glasses are extreme vertices (EV) based compositions coupled with Frit 418 at 36% WL and target the maximum Na{sub 2}O content (15.01 wt% Na{sub 2}O) of the SB7a VS glasses. Higher alkali glasses for which the model overpredicts the PCT response have been observed previously in the Sludge Batch 3 (SB3) Phase 1 VS and the Sludge Batch 6 (SB6) VS.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 6 QUALIFICATION

Prior to initiating a new sludge batch in the Defense Waste Processing Facility (DWPF), Savannah River National Laboratory (SRNL) is required to simulate this processing, including Chemical Process Cell (CPC) simulation, waste glass fabrication, and chemical durability testing. This report documents this simulation for the next sludge batch, Sludge Batch 6 (SB6). SB6 consists of Tank 12 material that has been transferred to Tank 51 and subjected to Low Temperature Aluminum Dissolution (LTAD), Tank 4 sludge, and H-Canyon Pu solutions. Following LTAD and the Tank 4 addition, Liquid Waste Operations (LWO) provided SRNL a 3 L sample of Tank 51 sludge for SB6 qualification. Pu solution from H Canyon was also received. SB6 qualification included washing the sample per LWO plans/projections (including the addition of Pu from H Canyon), DWPF CPC simulations, waste glass fabrication (vitrification), and waste glass characterization and chemical durability evaluation. The following are significant observations from this demonstration. Sludge settling improved slightly as the sludge was washed. SRNL recommended (and the Tank Farm implemented) one less wash based on evaluations of Tank 40 heel projections and projections of the glass composition following transfer of Tank 51 to Tank 40. Thorium was detected in significant quantities (>0.1 wt % of total solids) in the sludge. In past sludge batches, thorium has been determined by Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS), seen in small quantities, and reported with the radionuclides. As a result of the high thorium, SRNL-AD has added thorium to their suite of Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES) elements. The acid stoichiometry for the DWPF Sludge Receipt and Adjustment Tank (SRAT) processing of 115%, or 1.3 mol acid per liter of SRAT receipt slurry, was adequate to accomplish some of the goals of SRAT processing: nitrite was destroyed to below 1,000 mg/kg and mercury was removed to below the DWPF target with 750 g of steam per g of mercury. However, rheological properties did not improve and were above the design basis. Hydrogen generation rates did not exceed DWPF limits during the SRAT and Slurry Mix Evaporator (SME) cycles. However, hydrogen generation during the SRAT cycle approached the DWPF limit. The glass fabricated with the Tank 51 SB6 SME product and Frit 418 was acceptable with respect to chemical durability as measured by the Product Consistency Test (PCT). The PCT response was also predictable by the current durability models of the DWPF Product Composition Control System (PCCS). It should be noted, however, that in the first attempt to make glass from the SME product, the contents of the fabrication crucible foamed over. This may be a result of the SME product's REDOX (Reduction/Oxidation - Fe{sup 2+}/{Sigma}Fe) of 0.08 (calculated from SME product analytical results). The following are recommendations drawn from this demonstration. In this demonstration, at the request of DWPF, SRNL caustic boiled the SRAT contents prior to acid addition to remove water (to increase solids concentration). During the nearly five hours of caustic boiling, 700 ppm of antifoam was required to control foaming. SRNL recommends that DWPF not caustic boil/concentrate SRAT receipt prior to acid addition until further studies can be performed to provide a better foaming control strategy or a new antifoam is developed for caustic boiling. Based on this set of runs and a recently completed demonstration with the SB6 Waste Acceptance Product Specifications (WAPS) sample, it is recommended that DWPF not add formic acid at the design addition rate of two gallons per minute for this sludge batch. A longer acid addition time appears to be helpful in allowing slower reaction of formic acid with the sludge and possibly decreases the chance of a foam over during acid addition.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗